WorldWideScience

Sample records for fuel lattices benchmark

  1. Benchmarking computer platforms for lattice QCD applications

    International Nuclear Information System (INIS)

    Hasenbusch, M.; Jansen, K.; Pleiter, D.; Wegner, P.; Wettig, T.

    2003-09-01

    We define a benchmark suite for lattice QCD and report on benchmark results from several computer platforms. The platforms considered are apeNEXT, CRAY T3E, Hitachi SR8000, IBM p690, PC-Clusters, and QCDOC. (orig.)

  2. Benchmarking computer platforms for lattice QCD applications

    International Nuclear Information System (INIS)

    Hasenbusch, M.; Jansen, K.; Pleiter, D.; Stueben, H.; Wegner, P.; Wettig, T.; Wittig, H.

    2004-01-01

    We define a benchmark suite for lattice QCD and report on benchmark results from several computer platforms. The platforms considered are apeNEXT, CRAY T3E; Hitachi SR8000, IBM p690, PC-Clusters, and QCDOC

  3. A benchmark comparison of the Canadian Supercritical Water-Cooled Reactor (SCWR) 64-element fuel lattice cell parameters using various computer codes

    International Nuclear Information System (INIS)

    Sharpe, J.; Salaun, F.; Hummel, D.; Moghrabi, A.; Nowak, M.; Pencer, J.; Novog, D.; Buijs, A.

    2015-01-01

    Discrepancies in key lattice physics parameters have been observed between various deterministic (e.g. DRAGON and WIMS-AECL) and stochastic (MCNP, KENO) neutron transport codes in modeling previous versions of the Canadian SCWR lattice cell. Further, inconsistencies in these parameters have also been observed when using different nuclear data libraries. In this work, the predictions of k∞, various reactivity coefficients, and relative ring-averaged pin powers have been re-evaluated using these codes and libraries with the most recent 64-element fuel assembly geometry. A benchmark problem has been defined to quantify the dissimilarities between code results for a number of responses along the fuel channel under prescribed hot full power (HFP), hot zero power (HZP) and cold zero power (CZP) conditions and at several fuel burnups (0, 25 and 50 MW·d·kg"-"1 [HM]). Results from deterministic (TRITON, DRAGON) and stochastic codes (MCNP6, KENO V.a and KENO-VI) are presented. (author)

  4. A benchmark comparison of the Canadian Supercritical Water-Cooled Reactor (SCWR) 64-element fuel lattice cell parameters using various computer codes

    Energy Technology Data Exchange (ETDEWEB)

    Sharpe, J.; Salaun, F.; Hummel, D.; Moghrabi, A., E-mail: sharpejr@mcmaster.ca [McMaster University, Hamilton, ON (Canada); Nowak, M. [McMaster University, Hamilton, ON (Canada); Institut National Polytechnique de Grenoble, Phelma, Grenoble (France); Pencer, J. [McMaster University, Hamilton, ON (Canada); Canadian Nuclear Laboratories, Chalk River, ON, (Canada); Novog, D.; Buijs, A. [McMaster University, Hamilton, ON (Canada)

    2015-07-01

    Discrepancies in key lattice physics parameters have been observed between various deterministic (e.g. DRAGON and WIMS-AECL) and stochastic (MCNP, KENO) neutron transport codes in modeling previous versions of the Canadian SCWR lattice cell. Further, inconsistencies in these parameters have also been observed when using different nuclear data libraries. In this work, the predictions of k∞, various reactivity coefficients, and relative ring-averaged pin powers have been re-evaluated using these codes and libraries with the most recent 64-element fuel assembly geometry. A benchmark problem has been defined to quantify the dissimilarities between code results for a number of responses along the fuel channel under prescribed hot full power (HFP), hot zero power (HZP) and cold zero power (CZP) conditions and at several fuel burnups (0, 25 and 50 MW·d·kg{sup -1} [HM]). Results from deterministic (TRITON, DRAGON) and stochastic codes (MCNP6, KENO V.a and KENO-VI) are presented. (author)

  5. HELIOS calculations for UO2 lattice benchmarks

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1998-01-01

    Calculations for the ANS UO 2 lattice benchmark have been performed with the HELIOS lattice-physics code and six of its cross-section libraries. The results obtained from the different libraries permit conclusions to be drawn regarding the adequacy of the energy group structures and of the ENDF/B-VI evaluation for 238 U. Scandpower A/S, the developer of HELIOS, provided Los Alamos National Laboratory with six different cross section libraries. Three of the libraries were derived directly from Release 3 of ENDF/B-VI (ENDF/B-VI.3) and differ only in the number of groups (34, 89 or 190). The other three libraries are identical to the first three except for a modification to the cross sections for 238 U in the resonance range

  6. Reactor fuel depletion benchmark of TINDER

    International Nuclear Information System (INIS)

    Martin, W.J.; Oliveira, C.R.E. de; Hecht, A.A.

    2014-01-01

    Highlights: • A reactor burnup benchmark of TINDER, coupling MCNP6 to CINDER2008, was performed. • TINDER is a poor candidate for fuel depletion calculations using its current libraries. • Data library modification is necessary if fuel depletion is desired from TINDER. - Abstract: Accurate burnup calculations are key to proper nuclear reactor design, fuel cycle modeling, and disposal estimations. The TINDER code, originally designed for activation analyses, has been modified to handle full burnup calculations, including the widely used predictor–corrector feature. In order to properly characterize the performance of TINDER for this application, a benchmark calculation was performed. Although the results followed the trends of past benchmarked codes for a UO 2 PWR fuel sample from the Takahama-3 reactor, there were obvious deficiencies in the final result, likely in the nuclear data library that was used. Isotopic comparisons versus experiment and past code benchmarks are given, as well as hypothesized areas of deficiency and future work

  7. HELIOS2: Benchmarking against experiments for hexagonal and square lattices

    International Nuclear Information System (INIS)

    Simeonov, T.

    2009-01-01

    HELIOS2, is a 2D transport theory program for fuel burnup and gamma-flux calculation. It solves the neutron and gamma transport equations in a general, two-dimensional geometry bounded by a polygon of straight lines. The applied transport solver may be chosen between: The Method of Collision Probabilities (CP) and The Method of Characteristics(MoC). The former is well known for its successful application for preparation of cross section data banks for 3D simulators for all types lattices for WWERs, PWRs, BWRs, AGRs, RBMK and CANDU reactors. The later, MoC, helps in the areas where the requirements of CP for computational power become too large of practical application. The application of HELIOS2 and The Method of Characteristics for some large from calculation point of view benchmarks is presented in this paper. The analysis combines comparisons to measured data from the Hungarian ZR-6 reactor and JAERI facility of Tank type Critical Assembly (TCA) to verify and validate HELIOS2 and MOC for WWER assembly imitators; configurations with different absorber types- ZrB 2 , B 4 C, Eu 2 O 3 and Gd 2 O 3 ; and critical configurations with stainless steel in the reflector. Core eigenvalues and reaction rates are compared. With the account for the uncertainties the results are generally excellent. Special place in this paper is given to the effect of Iron-made radial reflector. Comparisons to measurements from TIC and TCA for stainless steel and Iron reflected cores are presented. The calculated by HELIOS-2 reactivity effect is in very good agreement with the measurements. (author)

  8. HELIOS2: Benchmarking Against Experiments for Hexagonal and Square Lattices

    International Nuclear Information System (INIS)

    Simeonov, T.

    2009-01-01

    HELIOS2, is a 2D transport theory program for fuel burnup and gamma-flux calculation. It solves the neutron and gamma transport equations in a general, two-dimensional geometry bounded by a polygon of straight lines. The applied transport solver may be chosen between: The Method of Collision Probabilities and The Method of Characteristics. The former is well known for its successful application for preparation of cross section data banks for 3D simulators for all types lattices for WWER's, PWR's, BWR's, AGR's, RBMK and CANDU reactors. The later, method of characteristics, helps in the areas where the requirements of collision probability for computational power become too large of practical application. The application of HELIOS2 and The method of characteristics for some large from calculation point of view benchmarks is presented in this paper. The analysis combines comparisons to measured data from the Hungarian ZR-6 reactor and JAERI's facility of tanktype critical assembly to verify and validate HELIOS2 and method of characteristics for WWER assembly imitators; configurations with different absorber types-ZrB2, B4C, Eu2O3 and Gd2O3; and critical configurations with stainless steel in the reflector. Core eigenvalues and reaction rates are compared. With the account for the uncertainties the results are generally excellent. Special place in this paper is given to the effect of Iron-made radial reflector. Comparisons to measurements from The Temporary International Collective and tanktype critical assembly for stainless steel and Iron reflected cores are presented. The calculated by HELIOS-2 reactivity effect is in very good agreement with the measurements. (Authors)

  9. Uranium-fuel thermal reactor benchmark testing of CENDL-3

    International Nuclear Information System (INIS)

    Liu Ping

    2001-01-01

    CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

  10. RIA Fuel Codes Benchmark - Volume 1

    International Nuclear Information System (INIS)

    Marchand, Olivier; Georgenthum, Vincent; Petit, Marc; Udagawa, Yutaka; Nagase, Fumihisa; Sugiyama, Tomoyuki; Arffman, Asko; Cherubini, Marco; Dostal, Martin; Klouzal, Jan; Geelhood, Kenneth; Gorzel, Andreas; Holt, Lars; Jernkvist, Lars Olof; Khvostov, Grigori; Maertens, Dietmar; Spykman, Gerold; Nakajima, Tetsuo; Nechaeva, Olga; Panka, Istvan; Rey Gayo, Jose M.; Sagrado Garcia, Inmaculada C.; Shin, An-Dong; Sonnenburg, Heinz Guenther; Umidova, Zeynab; Zhang, Jinzhao; Voglewede, John

    2013-01-01

    Reactivity-initiated accident (RIA) fuel rod codes have been developed for a significant period of time and they all have shown their ability to reproduce some experimental results with a certain degree of adequacy. However, they sometimes rely on different specific modelling assumptions the influence of which on the final results of the calculations is difficult to evaluate. The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burnup and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including the review of experimental approaches as well as the interpretation and use of experimental data relevant for safety. As a contribution to this task, WGFS conducted a RIA code benchmark based on RIA tests performed in the Nuclear Safety Research Reactor in Tokai, Japan and tests performed or planned in CABRI reactor in Cadarache, France. Emphasis was on assessment of different modelling options for RIA fuel rod codes in terms of reproducing experimental results as well as extrapolating to typical reactor conditions. This report provides a summary of the results of this task. (authors)

  11. Computational benchmark on the void reactivity effect in MOX lattices. Contribution to a NEA-NSC benchmark study organized by the Working Party on Plutonium Recycling

    International Nuclear Information System (INIS)

    Freudenreich, W.E.; Aaldijk, J.K.

    1994-08-01

    The Working Party on Plutonium Recycling of the Nuclear Science Committee of the OECD Nuclear Energy Agency has initiated a benchmark study on the calculation of the void reactivity effect in MOX lattices. The results presented here were obtained with the continuous energy, generalized geometry Monte Carlo transport code MCNP. The cross-section libraries used were processed from the JEF-2.2 evaluation taking into account selfshielding in the unresolved resonance ranges (selfshielding in the resolved resonance ranges is treated by MCNP). For an infinite lattice of unit cells a positive void reactivity effect was found only for the MOX fuel with the largest Pu content. For an infinite lattice of macro cells (voidable inner zone with different fuel mixtures surrounded by an outer zone of UO 2 fuel with moderator) a positive void reactivity effect was obtained for the three MOX fuel types considered. The results are not representative for MOX-loaded power reactor lattices, but serve only to intercompare reactor physics codes and libraries. (orig.)

  12. Fuel lattice design using heuristics and new strategies

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J. J.; Castillo M, J. A.; Torres V, M.; Perusquia del Cueto, R. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico); Pelta, D. A. [ETS Ingenieria Informatica y Telecomunicaciones, Universidad de Granada, Daniel Saucedo Aranda s/n, 18071 Granada (Spain); Campos S, Y., E-mail: juanjose.ortiz@inin.gob.m [IPN, Escuela Superior de Fisica y Matematicas, Unidad Profesional Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2010-10-15

    This work show some results of the fuel lattice design in BWRs when some allocation pin rod rules are not taking into account. Heuristics techniques like Path Re linking and Greedy to design fuel lattices were used. The scope of this work is to search about how do classical rules in design fuel lattices affect the heuristics techniques results and the fuel lattice quality. The fuel lattices quality is measured by Power Peaking Factor and Infinite Multiplication Factor at the beginning of the fuel lattice life. CASMO-4 code to calculate these parameters was used. The analyzed rules are the following: pin rods with lowest uranium enrichment are only allocated in the fuel lattice corner, and pin rods with gadolinium cannot allocated in the fuel lattice edge. Fuel lattices with and without gadolinium in the main diagonal were studied. Some fuel lattices were simulated in an equilibrium cycle fuel reload, using Simulate-3 to verify their performance. So, the effective multiplication factor and thermal limits can be verified. The obtained results show a good performance in some fuel lattices designed, even thought, the knowing rules were not implemented. A fuel lattice performance and fuel lattice design characteristics analysis was made. To the realized tests, a dell workstation was used, under Li nux platform. (Author)

  13. Fuel lattice design using heuristics and new strategies

    International Nuclear Information System (INIS)

    Ortiz S, J. J.; Castillo M, J. A.; Torres V, M.; Perusquia del Cueto, R.; Pelta, D. A.; Campos S, Y.

    2010-10-01

    This work show some results of the fuel lattice design in BWRs when some allocation pin rod rules are not taking into account. Heuristics techniques like Path Re linking and Greedy to design fuel lattices were used. The scope of this work is to search about how do classical rules in design fuel lattices affect the heuristics techniques results and the fuel lattice quality. The fuel lattices quality is measured by Power Peaking Factor and Infinite Multiplication Factor at the beginning of the fuel lattice life. CASMO-4 code to calculate these parameters was used. The analyzed rules are the following: pin rods with lowest uranium enrichment are only allocated in the fuel lattice corner, and pin rods with gadolinium cannot allocated in the fuel lattice edge. Fuel lattices with and without gadolinium in the main diagonal were studied. Some fuel lattices were simulated in an equilibrium cycle fuel reload, using Simulate-3 to verify their performance. So, the effective multiplication factor and thermal limits can be verified. The obtained results show a good performance in some fuel lattices designed, even thought, the knowing rules were not implemented. A fuel lattice performance and fuel lattice design characteristics analysis was made. To the realized tests, a dell workstation was used, under Li nux platform. (Author)

  14. Uncertainty Analysis of Light Water Reactor Fuel Lattices

    Directory of Open Access Journals (Sweden)

    C. Arenas

    2013-01-01

    Full Text Available The study explored the calculation of uncertainty based on available cross-section covariance data and computational tool on fuel lattice levels, which included pin cell and the fuel assembly models. Uncertainty variations due to temperatures changes and different fuel compositions are the main focus of this analysis. Selected assemblies and unit pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analysis were performed using TSUNAMI-2D sequence in SCALE 6.1. It was found that uncertainties increase with increasing temperature, while kinf decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributing reaction of uncertainty, namely, the neutron capture reaction 238U(n, γ due to the Doppler broadening. In addition, three types (UOX, MOX, and UOX-Gd2O3 of fuel material compositions were analyzed. A remarkable increase in uncertainty in kinf was observed for the case of MOX fuel. The increase in uncertainty of kinf in MOX fuel was nearly twice the corresponding value in UOX fuel. The neutron-nuclide reaction of 238U, mainly inelastic scattering (n, n′, contributed the most to the uncertainties in the MOX fuel, shifting the neutron spectrum to higher energy compared to the UOX fuel.

  15. Integral benchmarks with reference to thorium fuel cycle

    International Nuclear Information System (INIS)

    Ganesan, S.

    2003-01-01

    This is a power point presentation about the Indian participation in the CRP 'Evaluated Data for the Thorium-Uranium fuel cycle'. The plans and scope of the Indian participation are to provide selected integral experimental benchmarks for nuclear data validation, including Indian Thorium burn up benchmarks, post-irradiation examination studies, comparison of basic evaluated data files and analysis of selected benchmarks for Th-U fuel cycle

  16. Assessment of neutron transport codes for application to CANDU fuel lattices analysis

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    1999-08-01

    In order to assess the applicability of WIMS-AECL and HELIOS code to the CANDU fuel lattice analysis, the physics calculations has been carried out for the standard CANDU fuel and DUPIC fuel lattices, and the results were compared with those of Monte Carlo code MCNP-4B. In this study, in order to consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The TRX-1,2,BAPL-1,2,3 pin -cell lattices and KENO criticality safety benchmark calculations have been performed for the new MCNP libraries, and the results have shown that the new MCNP library has sufficient accuracy to be used for physics calculation. Then, the lattice codes have been benchmarked by the MCNP code for the major physics parameters such as the burnup reactivity, void reactivity, relative pin power and Doppler coefficient, etc. for the standard CANDU fuel and DUPIC fuel lattices. For the standard CANDU fuel lattice, it was found that the results of WIMS-AECL calculations are consistent with those of MCNP. For the DUPIC fuel lattice, however, the results of WIMS-AECL calculations with ENDF/B-V library have shown that the discrepancy from the results of MCNP calculations increases when the fuel burnup is relatively high. The burnup reactivities of WIMS-ACEL calculations with ENDF/B-VI library have shown excellent agreements with those of MCNP calculation for both the standard CANDU and DUPIC fuel lattices. However, the Doppler coefficient have relatively large discrepancies compared with MCNP calculations, and the difference increases as the fuel burns. On the other hand, the results of HELIOS calculation are consistent with those of MCNP even though the discrepancy is slightly larger compared with the case of the standard CANDU fuel lattice. this study has shown that the WIMS-AECL products reliable results for the natural uranium fuel. However, it is recommended that the WIMS

  17. Benchmarking lattice physics data and methods for boiling water reactor analysis

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Edenius, M.; Harris, D.R.; Hebert, M.J.; Kapitz, D.M.; Pilat, E.E.; VerPlanck, D.M.

    1983-01-01

    The objective of the work reported was to verify the adequacy of lattice physics modeling for the analysis of the Vermont Yankee BWR using a multigroup, two-dimensional transport theory code. The BWR lattice physics methods have been benchmarked against reactor physics experiments, higher order calculations, and actual operating data

  18. An optimized BWR fuel lattice for improved fuel utilization

    International Nuclear Information System (INIS)

    Bernander, O.; Helmersson, S.; Schoen, C.G.

    1984-01-01

    Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

  19. Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

    International Nuclear Information System (INIS)

    D'Hondt, P.; Gehin, J.; Na, B.C.; Sartori, E.; Wiesenack, W.

    2001-01-01

    One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

  20. Criticality benchmark guide for light-water-reactor fuel in transportation and storage packages

    International Nuclear Information System (INIS)

    Lichtenwalter, J.J.; Bowman, S.M.; DeHart, M.D.; Hopper, C.M.

    1997-03-01

    This report is designed as a guide for performing criticality benchmark calculations for light-water-reactor (LWR) fuel applications. The guide provides documentation of 180 criticality experiments with geometries, materials, and neutron interaction characteristics representative of transportation packages containing LWR fuel or uranium oxide pellets or powder. These experiments should benefit the U.S. Nuclear Regulatory Commission (NRC) staff and licensees in validation of computational methods used in LWR fuel storage and transportation concerns. The experiments are classified by key parameters such as enrichment, water/fuel volume, hydrogen-to-fissile ratio (H/X), and lattice pitch. Groups of experiments with common features such as separator plates, shielding walls, and soluble boron are also identified. In addition, a sample validation using these experiments and a statistical analysis of the results are provided. Recommendations for selecting suitable experiments and determination of calculational bias and uncertainty are presented as part of this benchmark guide

  1. In-core fuel management benchmarks for PHWRs

    International Nuclear Information System (INIS)

    1996-06-01

    Under its in-core fuel management activities, the IAEA set up two co-ordinated research programmes (CRPs) on complete in-core fuel management code packages. At a consultant meeting in November 1988 the outline of the CRP on in-core fuel management benchmars for PHWRs was prepared, three benchmarks were specified and the corresponding parameters were defined. At the first research co-ordination meeting in December 1990, seven more benchmarks were specified. The objective of this TECDOC is to provide reference cases for the verification of code packages used for reactor physics and fuel management of PHWRs. 91 refs, figs, tabs

  2. Benchmark test of CP-PACS for lattice QCD

    International Nuclear Information System (INIS)

    Yoshie, Tomoteru

    1996-01-01

    The CP-PACS is a massively parallel computer dedicated for calculations in computational physics and will be in operation in the spring of 1996 at Center for Computational Physics, University of Tsukuba. In this paper, we describe the architecture of the CP-PACS and report the results of the estimate of the performance of the CP-PACS for typical lattice QCD calculations. (author)

  3. Advanced fuel cycles options for LWRs and IMF benchmark definition

    International Nuclear Information System (INIS)

    Breza, J.; Darilek, P.; Necas, V.

    2008-01-01

    In the paper, different advanced nuclear fuel cycles including thorium-based fuel and inert-matrix fuel are examined under light water reactor conditions, especially VVER-440, and compared. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The inert-matrix fuel consist of plutonium and minor actinides separated from spent UOX fuel fixed in Yttria-stabilised zirconia matrix. The article shows analysed fuel cycles and their short description. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulating in the spent advanced thorium fuel and its comparison to UOX open fuel cycle. Definition of IMF benchmark based on presented scenario is given. (authors)

  4. Review of international solutions to NEACRP benchmark BWR lattice cell problems

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1977-12-01

    This paper summarises international solutions to a set of BWR benchmark problems. The problems, posed as an activity sponsored by the Nuclear Energy Agency Committee on Reactor Physics, were as follows: 9-pin supercell with central burnable poison pin, mini-BWR with 4 pin-cells and water gaps and control rod cruciform, full 7 x 7 pin BWR lattice cell with differential U 235 enrichment, and full 8 x 8 pin BWR lattice cell with water-hole, Pu-loading, burnable poison, and homogenised cruciform control rod. Solutions have been contributed by Denmark, Japan, Sweden, Switzerland and the UK. (author)

  5. Benchmarking

    OpenAIRE

    Meylianti S., Brigita

    1999-01-01

    Benchmarking has different meaning to different people. There are five types of benchmarking, namely internal benchmarking, competitive benchmarking, industry / functional benchmarking, process / generic benchmarking and collaborative benchmarking. Each type of benchmarking has its own advantages as well as disadvantages. Therefore it is important to know what kind of benchmarking is suitable to a specific application. This paper will discuss those five types of benchmarking in detail, includ...

  6. S/sub N/ computational benchmark solutions for slab geometry models of a gas-cooled fast reactor (GCFR) lattice cell

    International Nuclear Information System (INIS)

    McCoy, D.R.

    1981-01-01

    S/sub N/ computational benchmark solutions are generated for a onegroup and multigroup fuel-void slab lattice cell which is a rough model of a gas-cooled fast reactor (GCFR) lattice cell. The reactivity induced by the extrusion of the fuel material into the voided region is determined for a series of partially extruded lattice cell configurations. A special modified Gauss S/sub N/ ordinate array design is developed in order to obtain eigenvalues with errors less than 0.03% in all of the configurations that are considered. The modified Gauss S/sub N/ ordinate array design has a substantially improved eigenvalue angular convergence behavior when compared to existing S/sub N/ ordinate array designs used in neutron streaming applications. The angular refinement computations are performed in some cases by using a perturbation theory method which enables one to obtain high order S/sub N/ eigenvalue estimates for greatly reduced computational costs

  7. Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

    International Nuclear Information System (INIS)

    Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

    2004-01-01

    As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

  8. Spent reactor fuel benchmark composition data for code validation

    International Nuclear Information System (INIS)

    Bierman, S.R.

    1991-09-01

    To establish criticality safety margins utilizing burnup credit in the storage and transport of spent reactor fuels requires a knowledge of the uncertainty in the calculated fuel composition used in making the reactivity assessment. To provide data for validating such calculated burnup fuel compositions, radiochemical assays are being obtained as part of the United States Department of Energy From-Reactor Cask Development Program. Destructive assay data are being obtained from representative reactor fuels having experienced irradiation exposures up to about 55 GWD/MTM. Assay results and associated operating histories on the initial three samples analyzed in this effort are presented. The three samples were taken from different axial regions of the same fuel rod and represent radiation exposures of about 27, 37, and 44 GWD/MTM. The data are presented in a benchmark type format to facilitate identification/referencing and computer code input

  9. Analysis of benchmark lattices with endf/b-vi, jef-2.2 and jendl-3 data

    International Nuclear Information System (INIS)

    Saglam, M.

    1995-01-01

    The NJOY Nuclear Data Processing System has been used to process the ENDF/B-VI , JEF-2.2 and JENDL-3 Nuclear Cross Section Data Bases into multigroup form. A brief description of the data bases is given and the assumptions made in processing the data from evaluated nuclear data file format to multigroup format are presented. The differences and similarities of the Evaluated Nuclear Data Files have been investigated by producing four group cross sections by using the GROUPIE code and calculating thermal, fission spectrum averaged and 2200 m/s cross sections and resonance integrals using the INTER cale. It has been shown that the evaluated data for U238 in JEF and ENDF/B-VI are principally the same while in case of U235 the same is true for JENDL and ENDF/B-VI. The evaluations for U233 and Th232 are different for all three ENDF files. Several utility codes have been written to convert the multigroup library into a WIMS-D4 compatible binary library. The performance and suitability of the generated libraries have been tested with the use of metal tueled TRX lattices, uranium oxide fueled BAPL lattices and Th232-U233 fueled BNL lattices. The use ot a new thermal scattering matrix for Hydrogen from ENDF/B-VI increased keff for 0.5 o/ while the use of ENDF/B-VI U238 decreased it for 2.5 %. Although the original WIMS library performed well for Ihe effective multiplication factor of the lattices there is an improvement for the epithermal to thermal capture rate of U238 while using new data in the TRX and BAPL lattices. The effect of the fission spectrum is investigated for the BNL lattices and it is shown that using U233 fission spectrum instead of the original U235 spectrum gives a keff which agrees better with the experimental value. The results obtained by using new multigroup data are generally acceptable and in the experimental error range. They especially improve the prediction of the reaction rate dependent benchmark parameters

  10. Melcor benchmarking against integral severe fuel damage tests

    Energy Technology Data Exchange (ETDEWEB)

    Madni, I.K. [Brookhaven National Lab., Upton, NY (United States)

    1995-09-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the U.S. Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC to provide independent assessment of MELCOR, and a very important part of this program is to benchmark MELCOR against experimental data from integral severe fuel damage tests and predictions of that data from more mechanistic codes such as SCDAP or SCDAP/RELAP5. Benchmarking analyses with MELCOR have been carried out at BNL for five integral severe fuel damage tests, namely, PBF SFD 1-1, SFD 14, and NRU FLHT-2, analyses, and their role in identifying areas of modeling strengths and weaknesses in MELCOR.

  11. Analysis of the international criticality benchmark no 19 of a realistic fuel dissolver

    International Nuclear Information System (INIS)

    Smith, H.J.; Santamarina, A.

    1991-01-01

    The dispersion of the order of 12000 pcm in the results of the international criticality fuel dissolver benchmark calculation, exercise OECD/19, showed the necessity of analysing the calculational methods used in this case. The APOLLO/PIC method developed to treat this type of problem permits us to propose international reference values. The problem studied here, led us to investigate two supplementary parameters in addition to the double heterogeneity of the fuel: the reactivity variation as a function of moderation and the effects of the size of the fuel pellets during dissolution. The following conclusions were obtained: The fast cross-section sets used by the international SCALE package introduces a bias of - 3000 pcm in undermoderated lattices. More generally, the fast and resonance nuclear data in criticality codes are not sufficiently reliable. Geometries with micro-pellets led to an underestimation of reactivity at the end of dissolution of 3000 pcm in certain 1988 Sn calculations; this bias was avoided in the up-dated 1990 computation because of a correct use of calculation tools. The reactivity introduced by the dissolved fuel is underestimated by 3000 pcm in contributions based on the standard NITAWL module in the SCALE code. More generally, the neutron balance analysis pointed out that standard ND self shielding formalism cannot account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The combination of these three types of bias explain the underestimation of all of the international contributions of the reactivity of dissolver lattices by -2000 to -6000 pcm. The improved 1990 calculations confirm the need to use rigorous methods in the calculation of systems which involve the fuel double heterogeneity. This study points out the importance of periodic benchmarking exercises for probing the efficacity of criticality codes, data libraries and the users

  12. A simplified approach to WWER-440 fuel assembly head benchmark

    International Nuclear Information System (INIS)

    Muehlbauer, P.

    2010-01-01

    The WWER-440 fuel assembly head benchmark was simulated with FLUENT 12 code as a first step of validation of the code for nuclear reactor safety analyses. Results of the benchmark together with comparison of results provided by other participants and results of measurement will be presented in another paper by benchmark organisers. This presentation is therefore focused on our approach to this simulation as illustrated on the case 323-34, which represents a peripheral assembly with five neighbours. All steps of the simulation and some lessons learned are described. Geometry of the computational region supplied as STEP file by organizers of the benchmark was first separated into two parts (inlet part with spacer grid, and the rest of assembly head) in order to keep the size of the computational mesh manageable with regard to the hardware available (HP Z800 workstation with Intel Zeon four-core CPU 3.2 GHz, 32 GB of RAM) and then further modified at places where shape of the geometry would probably lead to highly distorted cells. Both parts of the geometry were connected via boundary profile file generated at cross section, where effect of grid spacers is still felt but the effect of out flow boundary condition used in the computations of the inlet part of geometry is negligible. Computation proceeded in several steps: start with basic mesh, standard k-ε model of turbulence with standard wall functions and first order upwind numerical schemes; after convergence (scaled residuals lower than 10-3) and near-wall meshes local adaptation when needed, realizable k-ε of turbulence was used with second order upwind numerical schemes for momentum and energy equations. During iterations, area-average temperature of thermocouples and area-averaged outlet temperature which are the main figures of merit of the benchmark were also monitored. In this 'blind' phase of the benchmark, effect of spacers was neglected. After results of measurements are available, standard validation

  13. Calculation of Single Cell and Fuel Assembly IRIS Benchmarks Using WIMSD5B and GNOMER Codes

    International Nuclear Information System (INIS)

    Pevec, D.; Grgic, D.; Jecmenica, R.

    2002-01-01

    IRIS reactor (an acronym for International Reactor Innovative and Secure) is a modular, integral, light water cooled, small to medium power (100-335 MWe/module) reactor, which addresses the requirements defined by the United States Department of Energy for Generation IV nuclear energy systems, i.e., proliferation resistance, enhanced safety, improved economics, and waste reduction. An international consortium led by Westinghouse/BNFL was created for development of IRIS reactor; it includes universities, institutes, commercial companies, and utilities. Faculty of Electrical Engineering and Computing, University of Zagreb joined the consortium in year 2001, with the aim to take part in IRIS neutronics design and safety analyses of IRIS transients. A set of neutronic benchmarks for IRIS reactor was defined with the objective to compare results of all participants with exactly the same assumptions. In this paper a calculation of Benchmark 44 for IRIS reactor is described. Benchmark 44 is defined as a core depletion benchmark problem for specified IRIS reactor operating conditions (e.g., temperatures, moderator density) without feedback. Enriched boron, inhomogeneously distributed in axial direction, is used as an integral fuel burnable absorber (IFBA). The aim of this benchmark was to enable a more direct comparison of results of different code systems. Calculations of Benchmark 44 were performed using the modified CORD-2 code package. The CORD-2 code package consists of WIMSD and GNOMER codes. WIMSD is a well-known lattice spectrum calculation code. GNOMER solves the neutron diffusion equation in three-dimensional Cartesian geometry by the Green's function nodal method. The following parameters were obtained in Benchmark 44 analysis: effective multiplication factor as a function of burnup, nuclear peaking factor as a function of burnup, axial offset as a function of burnup, core-average axial power profile, core radial power profile, axial power profile for selected

  14. Benchmarking criticality analysis of TRIGA fuel storage racks.

    Science.gov (United States)

    Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

    2017-01-01

    A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel. Copyright © 2016. Published by Elsevier Ltd.

  15. Benchmarking Computational Fluid Dynamics for Application to PWR Fuel

    International Nuclear Information System (INIS)

    Smith, L.D. III; Conner, M.E.; Liu, B.; Dzodzo, B.; Paramonov, D.V.; Beasley, D.E.; Langford, H.M.; Holloway, M.V.

    2002-01-01

    The present study demonstrates a process used to develop confidence in Computational Fluid Dynamics (CFD) as a tool to investigate flow and temperature distributions in a PWR fuel bundle. The velocity and temperature fields produced by a mixing spacer grid of a PWR fuel assembly are quite complex. Before using CFD to evaluate these flow fields, a rigorous benchmarking effort should be performed to ensure that reasonable results are obtained. Westinghouse has developed a method to quantitatively benchmark CFD tools against data at conditions representative of the PWR. Several measurements in a 5 x 5 rod bundle were performed. Lateral flow-field testing employed visualization techniques and Particle Image Velocimetry (PIV). Heat transfer testing involved measurements of the single-phase heat transfer coefficient downstream of the spacer grid. These test results were used to compare with CFD predictions. Among the parameters optimized in the CFD models based on this comparison with data include computational mesh, turbulence model, and boundary conditions. As an outcome of this effort, a methodology was developed for CFD modeling that provides confidence in the numerical results. (authors)

  16. Benchmark solution of contemporary PWR integral fuel burnable absorbers

    International Nuclear Information System (INIS)

    Stucker, D.L.; Hone, M.J.; Holland, R.A.

    1993-01-01

    This paper presents a closely controlled benchmark solution of the two major contemporary pressurized water reactor integral burnable absorber designs: zirconium diboride (ZrB 2 ) and gadolinia (Gd 2 O 3 ). The comparison is accomplished using self-generating equilibrium cycles with equal energy, equal discharge burnup, and equal safety constraints. The reference plant for this evaluation is a 3411-MW(thermal) Westinghouse four-loop nuclear steam supply system operating with an inlet temperature of 285.9 degrees C, a core coolant mass now rate of 16877.3 kg/s, and coolant pressure of 15.5 MPa. The reactor consists of 193 VANTAGE 5H fuel assemblies that are discharged at a region average burnup of 48.4 GWd/tonne U. Each fuel assembly contains a natural uranium axial blanket 15.24 cm long at the top and the bottom of the fuel rod. The burnable absorber rods are symmetrically radially dispersed within the fuel assembly such that intrabundle power peaking is minimized. The burnable absorber material for both ZrB 2 and Gd 2 O 3 is axially zoned to the central 304.8 cm of the absorber-bearing fuel rods. The fuel management was constrained such that the thermal and safety limitations of F δH q -5 /degrees C were simultaneously achieved. The maximum long-term operating soluble boron concentration was also limited to 446 effective full-power days (EFPDs) including 14 EFPDs of power coastdown were assumed

  17. Benchmarking of epithermal methods in the lattice-physics code EPRI-CELL

    International Nuclear Information System (INIS)

    Williams, M.L.; Wright, R.Q.; Barhen, J.; Rothenstein, W.; Toney, B.

    1982-01-01

    The epithermal cross section shielding methods used in the lattice physics code EPRI-CELL (E-C) have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology. These include: treatment of the external moderator source with intermediate resonance (IR) theory, development of a new Dancoff factor expression to account for clad interactions, development of a new method for treating resonance interference, and application of a generalized least squares method to compute best-estimate values for the Bell factor and group-dependent IR parameters. The modified E-C code with its new ENDF/B-V cross section library is tested for several numerical benchmark problems. Integral parameters computed by EC are compared with those obtained with point-cross section Monte Carlo calculations, and E-C fine group cross sections are benchmarked against point-cross section descrete ordinates calculations. It is found that the code modifications improve agreement between E-C and the more sophisticated methods. E-C shows excellent agreement on the integral parameters and usually agrees within a few percent on fine-group, shielded cross sections

  18. Nuclear analysis of the experimental VHTR fuel lattice

    International Nuclear Information System (INIS)

    Doi, Takeshi; Shindo, Ryuiti; Hirano, Mitsumasa; Takano, Makoto

    1984-11-01

    Nuclear properties of a fuel lattice in the experimental VHTR core were analyzed with DELIGHT-6 and SRAC codes. Analytical results by both codes were compared by using various calculational model. The nuclear parameters were analyzed, such as a multiplication factor of a fuel lattice and it's variation with burnup, a temperature effect on reactivity, an effect of double-heterogeniety in a resonance absorption calculation, a resonance integral of 238 U and a reactivity worth of burnable poison. From these analyses, following results were obtained. Firstly, on calculational models, 1) Effect of double-heterogeniety in the resonance absorption calculation for Mark-III fuel element, causing a decrease of about 5.5 barns in the resonance integral and an increase of about 2.6 %ΔK in the infinite multiplication factor, 2) The heterogeneous calculation with the collision probability method resulted in about 0.6 %ΔK higher the multiplication factor of fuel lattice than that with the point model, 3) The reactivity worth of burnable poison rod by a multi-region model is about 20 % less than that by a 2-region model at an initial state of burnup and it's variation with burnup are fairly different, Secondly, on comparison between the results by DELIGHT-6 and SRAC, 4) The nuclear parameters obtained with both codes agreed well, Lastly, on the improvement in DELIGHT-6, 5) Consideration of the neutron spectrum shielding effect in the resonance effective cross section calculation caused a decrease of about 2.4 %ΔK in the multiplication factor of fuel lattice, 6) The lattice multiplication factor increased about 0.5 %ΔK by introducing lambda-parameters for the non-resonant nuclie. (J.P.N.)

  19. Benchmark problem suite for reactor physics study of LWR next generation fuels

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Ikehara, Tadashi; Ito, Takuya; Saji, Etsuro

    2002-01-01

    This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70 GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO 2 and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management. (author)

  20. DRAGON analysis of MOX fueled VVER cell benchmarks

    International Nuclear Information System (INIS)

    Marleau, G.; Foissac, F.

    2002-01-01

    The computational unit-cell benchmarks problems for LEU and MOX fueled VVER-1000 ('water-water energetic reactor') have been analyzed using the code DRAGON with ENDF/B-V and ENDF/B-VI based WIMS-AECL cross section libraries. The results obtained were compared with those generated using the SAS2H module of the SCALE-4.3 computational code system and with the code HELIOS. Good agreements between DRAGON and HELIOS were obtained when the ENDF/B-VI based library was considered while the ENDF/B-V DRAGON results were generally closer to those obtained using SAS2H. This study was useful for the verification of the DRAGON code and confirms that HELIOS and DRAGON have a similar behavior when compatible cross sections library are used. (author)

  1. Benchmark calculation of APOLLO-2 and SLAROM-UF in a fast reactor lattice

    International Nuclear Information System (INIS)

    Hazama, T.

    2009-07-01

    A lattice cell benchmark calculation is carried out for APOLLO2 and SLAROM-UF on the infinite lattice of a simple pin cell featuring a fast reactor. The accuracy in k-infinity and reaction rates is investigated in their reference and standard level calculations. In the 1. reference level calculation, APOLLO2 and SLAROM-UF agree with the reference value of k-infinity obtained by a continuous energy Monte Carlo calculation within 50 pcm. However, larger errors are observed in a particular reaction rate and energy range. The major problem common to both codes is in the cross section library of 239 Pu in the unresolved energy range. In the 2. reference level calculation, which is based on the ECCO 1968 group structure, both results of k-infinity agree with the reference value within 100 pcm. The resonance overlap effect is observed by several percents in cross sections of heavy nuclides. In the standard level calculation based on the APOLLO2 library creation methodology, a discrepancy appears by more than 300 pcm. A restriction is revealed in APOLLO2. Its standard cross section library does not have a sufficiently small background cross section to evaluate the self shielding effect on 56 Fe cross sections. The restriction can be removed by introducing the mixture self-shielding treatment recently introduced to APOLLO2. SLAROM-UF original standard level calculation based on the JFS-3 library creation methodology is the best among the standard level calculations. Improvement from the SLAROM-UF standard level calculation is achieved mainly by use of a proper weight function for light or intermediate nuclides. (author)

  2. Plutonium fuel lattice neutron behavior in inert matrix

    International Nuclear Information System (INIS)

    Hernandez L, H.; Lucatero, M. A.

    2010-10-01

    In several countries is had been researching the possibility of using plutonium, as weapon degree as reactor degree, as fuel material in commercial reactors to generate electricity. In special a great development has been in Pressure Water Reactors. However, in Mexico the reactors are Boiling Water Reactors type, reason for which the necessity to considers feasibility to use this fuel type in the reactors of nuclear power plant of Laguna Verde. For this propose a comparison of fuel lattice that compose a fuel assembly is made. The fuel assembly will propose to be used whit in the reactor present different inert matrix, as well as burnable poison. The material that compose the inert matrices used are cerium and zirconia (CeO 2 and ZrO 2 ) and as burnable poisons have gadolinium and erbium (Gd 2 O 4 and ErO 2 ). As far as the hydraulic design used is a cell 10 X 10 with two water channels. The lattice calculations are made with the Helios code a library with 35 energy groups, having determined the pin power factors, the infinite multiplication factor and the neutron flux profiles. (Author)

  3. WWER in-core fuel management benchmark definition

    International Nuclear Information System (INIS)

    Apostolov, T.; Alekova, G.; Prodanova, R.; Petrova, T.; Ivanov, K.

    1994-01-01

    Two benchmark problems for WWER-440, including design parameters, operating conditions and measured quantities are discussed in this paper. Some benchmark results for infinitive multiplication factor -K eff , natural boron concentration - C β and relative power distribution - K q obtained by use of the code package are represented. (authors). 5 refs., 3 tabs

  4. Gamma ray benchmark on the spent fuel shipping cask TN 12

    International Nuclear Information System (INIS)

    Blum, P.; Cagnon, R.; Cladel, C.; Ermont, G.; Nimal, J.C.

    1983-05-01

    The purpose of this benchmark is to compare measurements and calculation of gamma-ray dose rates around a shipping cask loaded with 12 spent fuel elements of FESSENHEIM PWR type. The benchmark provides a means to verify gamma-ray sources and gamma-ray transport calculation methods in shipping cask configurations. The comparison between measurements and calculations shows a good agreement except near the fuel element top where the discrepancy reaches a factor 2

  5. Benchmarking

    OpenAIRE

    Beretta Sergio; Dossi Andrea; Grove Hugh

    2000-01-01

    Due to their particular nature, the benchmarking methodologies tend to exceed the boundaries of management techniques, and to enter the territories of managerial culture. A culture that is also destined to break into the accounting area not only strongly supporting the possibility of fixing targets, and measuring and comparing the performance (an aspect that is already innovative and that is worthy of attention), but also questioning one of the principles (or taboos) of the accounting or...

  6. X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable.

  7. X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

    International Nuclear Information System (INIS)

    Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk

    2016-01-01

    In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable

  8. Calculations of IAEA-CRP-6 Benchmark Case 1 through 7 for a TRISO-Coated Fuel Particle

    International Nuclear Information System (INIS)

    Kim, Young Min; Lee, Y. W.; Chang, J. H.

    2005-01-01

    IAEA-CRP-6 is a coordinated research program of IAEA on Advances in HTGR fuel technology. The CRP examines aspects of HTGR fuel technology, ranging from design and fabrication to characterization, irradiation testing, performance modeling, as well as licensing and quality control issues. The benchmark section of the program treats simple analytical cases, pyrocarbon layer behavior, single TRISO-coated fuel particle behavior, and benchmark calculations of some irradiation experiments performed and planned. There are totally seventeen benchmark cases in the program. Member countries are participating in the benchmark calculations of the CRP with their own developed fuel performance analysis computer codes. Korea is also taking part in the benchmark calculations using a fuel performance analysis code, COPA (COated PArticle), which is being developed in Korea Atomic Energy Research Institute. The study shows the calculational results of IAEACRP- 6 benchmark cases 1 through 7 which describe the structural behaviors for a single fuel particle

  9. Fuel lattice design in a boiling water reactor using a knowledge-based automation system

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Wu-Hsiung, E-mail: wstong@iner.gov.tw; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

    2015-11-15

    Highlights: • An automation system was developed for the fuel lattice radial design of BWRs. • An enrichment group peaking equalizing method is applied to optimize the design. • Several heuristic rules and restrictions are incorporated to facilitate the design. • The CPU time for the system to design a 10x10 lattice was less than 1.2 h. • The beginning-of-life LPF was improved from 1.319 to 1.272 for one of the cases. - Abstract: A knowledge-based fuel lattice design automation system for BWRs is developed and applied to the design of 10 × 10 fuel lattices. The knowledge implemented in this fuel lattice design automation system includes the determination of gadolinium fuel pin location, the determination of fuel pin enrichment and enrichment distribution. The optimization process starts by determining the gadolinium distribution based on the pin power distribution of a flat enrichment lattice and some heuristic rules. Next, a pin power distribution flattening and an enrichment grouping process are introduced to determine the enrichment of each fuel pin enrichment type and the initial enrichment distribution of a fuel lattice design. Finally, enrichment group peaking equalizing processes are performed to achieve lower lattice peaking. Several fuel lattice design constraints are also incorporated in the automation system such that the system can accomplish a design which meets the requirements of practical use. Depending on the axial position of the lattice, a different method is applied in the design of the fuel lattice. Two typical fuel lattices with U{sup 235} enrichment of 4.471% and 4.386% were taken as references. Application of the method demonstrates that improved lattice designs can be achieved through the enrichment grouping and the enrichment group peaking equalizing method. It takes about 11 min and 1 h 11 min of CPU time for the automation system to accomplish two design cases on an HP-8000 workstation, including the execution of CASMO-4

  10. Fuel lattice design in a boiling water reactor using a knowledge-based automation system

    International Nuclear Information System (INIS)

    Tung, Wu-Hsiung; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

    2015-01-01

    Highlights: • An automation system was developed for the fuel lattice radial design of BWRs. • An enrichment group peaking equalizing method is applied to optimize the design. • Several heuristic rules and restrictions are incorporated to facilitate the design. • The CPU time for the system to design a 10x10 lattice was less than 1.2 h. • The beginning-of-life LPF was improved from 1.319 to 1.272 for one of the cases. - Abstract: A knowledge-based fuel lattice design automation system for BWRs is developed and applied to the design of 10 × 10 fuel lattices. The knowledge implemented in this fuel lattice design automation system includes the determination of gadolinium fuel pin location, the determination of fuel pin enrichment and enrichment distribution. The optimization process starts by determining the gadolinium distribution based on the pin power distribution of a flat enrichment lattice and some heuristic rules. Next, a pin power distribution flattening and an enrichment grouping process are introduced to determine the enrichment of each fuel pin enrichment type and the initial enrichment distribution of a fuel lattice design. Finally, enrichment group peaking equalizing processes are performed to achieve lower lattice peaking. Several fuel lattice design constraints are also incorporated in the automation system such that the system can accomplish a design which meets the requirements of practical use. Depending on the axial position of the lattice, a different method is applied in the design of the fuel lattice. Two typical fuel lattices with U"2"3"5 enrichment of 4.471% and 4.386% were taken as references. Application of the method demonstrates that improved lattice designs can be achieved through the enrichment grouping and the enrichment group peaking equalizing method. It takes about 11 min and 1 h 11 min of CPU time for the automation system to accomplish two design cases on an HP-8000 workstation, including the execution of CASMO-4 lattice

  11. Benchmarking Benchmarks

    NARCIS (Netherlands)

    D.C. Blitz (David)

    2011-01-01

    textabstractBenchmarking benchmarks is a bundle of six studies that are inspired by the prevalence of benchmarking in academic finance research as well as in investment practice. Three studies examine if current benchmark asset pricing models adequately describe the cross-section of stock returns.

  12. Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark

    International Nuclear Information System (INIS)

    Jecmenica, R.; Pevec, D.; Grgic, D.; Konjarek, D.

    2008-01-01

    FA2D is 2D transport collision probability code developed at Faculty of Electrical Engineering and Computing, University Zagreb. It is used for calculation of cross section data at fuel assembly level. Main objective of its development was capability to generate cross section data to be used for fuel management and safety analyses of PWR reactors. Till now formal verification of code predictions capability is not performed at fuel assembly level, but results of fuel management calculations obtained using FA2D generated cross sections for NPP Krsko and IRIS reactor are compared against Westinghouse calculations. Cross section data were used within NRC's PARCS code and satisfactory preliminary results were obtained. This paper presents results of calculations performed for Nuclear Fuel Industries, Ltd., benchmark using FA2D, and SCALE5 TRITON calculation sequence (based on discrete ordinates code NEWT). Nuclear Fuel Industries, Ltd., Japan, released LWR Next Generation Fuels Benchmark with the aim to verify prediction capability in nuclear design for extended burnup regions. We performed calculations for two different Benchmark problem geometries - UO 2 pin cell and UO 2 PWR fuel assembly. The results obtained with two mentioned 2D spectral codes are presented for burnup dependency of infinite multiplication factor, isotopic concentration of important materials and for local peaking factor vs. burnup (in case of fuel assembly calculation).(author)

  13. Inspection device for fuel rod restraint by support lattice of fuel assembly

    International Nuclear Information System (INIS)

    Hasegawa, Isao; Senga, Masatoshi; Kada, Mitoshi.

    1991-01-01

    An inspection operation section for disposing fuel assembly vertically at predetermined positions, a control section wired therewith, a moving operation section movable in the direction of X, Y and Z axes by a driving signal sent from the control section are disposed to an inspection section main body. A downward bore scope and a upward bore scope, each of such a size as can be inserted to the gaps between the fuel rods, are disposed while opposing to each other for observing the inside of each of cells from above and below in support lattices of fuel assemblies. High performance television cameras are disposed to each of bore scopes to supply images to monitoring televisions in the control section. Thus, a displacing operation section of the inspection operation section is automatically controlled three-dimensionally, the downward bore scope and the upward bore scope are integrally intruded to the inside of the gaps between the predetermined fuel rods from a required height and stopped at a predetermined position, mounted automatically to a required cell of the support lattice to efficiently observe and inspect the fuel rod restraint. (N.H.)

  14. Verify Super Double-Heterogeneous Spherical Lattice Model for Equilibrium Fuel Cycle Analysis AND HTR Spherical Super Lattice Model for Equilibrium Fuel Cycle Analysis

    International Nuclear Information System (INIS)

    Gray S. Chang

    2005-01-01

    The currently being developed advanced High Temperature gas-cooled Reactors (HTR) is able to achieve a simplification of safety through reliance on innovative features and passive systems. One of the innovative features in these HTRs is reliance on ceramic-coated fuel particles to retain the fission products even under extreme accident conditions. Traditionally, the effect of the random fuel kernel distribution in the fuel pebble/block is addressed through the use of the Dancoff correction factor in the resonance treatment. However, the Dancoff correction factor is a function of burnup and fuel kernel packing factor, which requires that the Dancoff correction factor be updated during Equilibrium Fuel Cycle (EqFC) analysis. An advanced KbK-sph model and whole pebble super lattice model (PSLM), which can address and update the burnup dependent Dancoff effect during the EqFC analysis. The pebble homogeneous lattice model (HLM) is verified by the burnup characteristics with the double-heterogeneous KbK-sph lattice model results. This study summarizes and compares the KbK-sph lattice model and HLM burnup analyzed results. Finally, we discuss the Monte-Carlo coupling with a fuel depletion and buildup code--ORIGEN-2 as a fuel burnup analysis tool and its PSLM calculated results for the HTR EqFC burnup analysis

  15. Benchmark criticality experiments for fast fission configuration with high enriched nuclear fuel

    International Nuclear Information System (INIS)

    Sikorin, S.N.; Mandzik, S.G.; Polazau, S.A.; Hryharovich, T.K.; Damarad, Y.V.; Palahina, Y.A.

    2014-01-01

    Benchmark criticality experiments of fast heterogeneous configuration with high enriched uranium (HEU) nuclear fuel were performed using the 'Giacint' critical assembly of the Joint Institute for Power and Nuclear Research - Sosny (JIPNR-Sosny) of the National Academy of Sciences of Belarus. The critical assembly core comprised fuel assemblies without a casing for the 34.8 mm wrench. Fuel assemblies contain 19 fuel rods of two types. The first type is metal uranium fuel rods with 90% enrichment by U-235; the second one is dioxide uranium fuel rods with 36% enrichment by U-235. The total fuel rods length is 620 mm, and the active fuel length is 500 mm. The outer fuel rods diameter is 7 mm, the wall is 0.2 mm thick, and the fuel material diameter is 6.4 mm. The clad material is stainless steel. The side radial reflector: the inner layer of beryllium, and the outer layer of stainless steel. The top and bottom axial reflectors are of stainless steel. The analysis of the experimental results obtained from these benchmark experiments by developing detailed calculation models and performing simulations for the different experiments is presented. The sensitivity of the obtained results for the material specifications and the modeling details were examined. The analyses used the MCNP and MCU computer programs. This paper presents the experimental and analytical results. (authors)

  16. Development of computer code SIMPSEX for simulation of FBR fuel reprocessing flowsheets: II. additional benchmarking results

    International Nuclear Information System (INIS)

    Shekhar Kumar; Koganti, S.B.

    2003-07-01

    Benchmarking and application of a computer code SIMPSEX for high plutonium FBR flowsheets was reported recently in an earlier report (IGC-234). Improvements and recompilation of the code (Version 4.01, March 2003) required re-validation with the existing benchmarks as well as additional benchmark flowsheets. Improvements in the high Pu region (Pu Aq >30 g/L) resulted in better results in the 75% Pu flowsheet benchmark. Below 30 g/L Pu Aq concentration, results were identical to those from the earlier version (SIMPSEX Version 3, code compiled in 1999). In addition, 13 published flowsheets were taken as additional benchmarks. Eleven of these flowsheets have a wide range of feed concentrations and few of them are β-γ active runs with FBR fuels having a wide distribution of burnup and Pu ratios. A published total partitioning flowsheet using externally generated U(IV) was also simulated using SIMPSEX. SIMPSEX predictions were compared with listed predictions from conventional SEPHIS, PUMA, PUNE and PUBG. SIMPSEX results were found to be comparable and better than the result from above listed codes. In addition, recently reported UREX demo results along with AMUSE simulations are also compared with SIMPSEX predictions. Results of the benchmarking SIMPSEX with these 14 benchmark flowsheets are discussed in this report. (author)

  17. Thermal lattice benchmarks for testing basic evaluated data files, developed with MCNP4B

    International Nuclear Information System (INIS)

    Maucec, M.; Glumac, B.

    1996-01-01

    The development of unit cell and full reactor core models of DIMPLE S01A and TRX-1 and TRX-2 benchmark experiments, using Monte Carlo computer code MCNP4B is presented. Nuclear data from ENDF/B-V and VI version of cross-section library were used in the calculations. In addition, a comparison to results obtained with the similar models and cross-section data from the EJ2-MCNPlib library (which is based upon the JEF-2.2 evaluation) developed in IRC Petten, Netherlands is presented. The results of the criticality calculation with ENDF/B-VI data library, and a comparison to results obtained using JEF-2.2 evaluation, confirm the MCNP4B full core model of a DIMPLE reactor as a good benchmark for testing basic evaluated data files. On the other hand, the criticality calculations results obtained using the TRX full core models show less agreement with experiment. It is obvious that without additional data about the TRX geometry, our TRX models are not suitable as Monte Carlo benchmarks. (author)

  18. Benchmarking Pt-based electrocatalysts for low temperature fuel cell reactions with the rotating disk electrode

    DEFF Research Database (Denmark)

    Pedersen, Christoffer Mølleskov; Escribano, Maria Escudero; Velazquez-Palenzuela, Amado Andres

    2015-01-01

    We present up-to-date benchmarking methods for testing electrocatalysts for polymer exchange membrane fuel cells (PEMFC), using the rotating disk electrode (RDE) method. We focus on the oxygen reduction reaction (ORR) and the hydrogen oxidation reaction (HOR) in the presence of CO. We have chosen...

  19. Optimization of BWR fuel lattice enrichment and gadolinia distribution using genetic algorithms and knowledge

    International Nuclear Information System (INIS)

    Martin-del-Campo, Cecilia; Francois, Juan Luis; Carmona, Roberto; Oropeza, Ivonne P.

    2007-01-01

    An optimization methodology based on the Genetic Algorithms (GA) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The optimization algorithm was linked to the HELIOS code to evaluate the neutronic parameters included in the objective function. The goal is to search for a fuel lattice with the lowest average enrichment, which satisfy a reactivity target, a local power peaking factor (PPF), lower than a limit value, and an average gadolinia concentration target. The methodology was applied to the design of a 10 x 10 fuel lattice, which can be used in fuel assemblies currently used in the two BWRs operating at Mexico. The optimization process showed an excellent performance because it found forty lattice designs in which the worst one has a better neutronic performance than the reference lattice design. The main contribution of this study is the development of an efficient procedure for BWR fuel lattice design, using GA with an objective function (OF) which saves computing time because it does not require lattice burnup calculations

  20. Benchmark physics tests in the metallic-fueled assembly ZPPR-15

    International Nuclear Information System (INIS)

    McFarlane, H.F.; Brumbach, S.B.; Carpenter, S.G.; Collins, P.J.

    1989-01-01

    Results of the first benchmark physics tests of a metallic-fueled, demonstration-size liquid-metal reactor (LMR) are reported. A simple, two-zone, cylindrical conventional assembly was built with three distinctly different compositions to represent the stages of the Integral Fast Reactor fuel cycle. Experiments included criticality, control, power distribution, reaction rate ratios, reactivity coefficients, shielding, kinetics, and spectrum. Analysis was done with three-dimensional nodal diffusion calculations and ENDF/B-V.2 cross sections. Predictions of the ZPPR-15 reactor physics parameters agreed sufficiently well with the measured values to justify confidence in design analyses for metallic-fueled LMRs

  1. Mixed-oxide (MOX) fuel performance benchmark. Summary of the results for the PRIMO MOX rod BD8

    International Nuclear Information System (INIS)

    Ott, L.J.; Sartori, E.; Costa, A.; ); Sobolev, V.; Lee, B-H.; Alekseev, P.N.; Shestopalov, A.A.; Mikityuk, K.O.; Fomichenko, P.A.; Shatrova, L.P.; Medvedev, A.V.; Bogatyr, S.M.; Khvostov, G.A.; Kuznetsov, V.I.; Stoenescu, R.; Chatwin, C.P.

    2009-01-01

    The OECD/NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, nuclear fuel performance, and fuel cycle issues related to the disposition of weapons-grade plutonium as MOX fuel. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close cooperation with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A major part of these activities includes benchmark studies. This report describes the results of the PRIMO rod BD8 benchmark exercise, the second benchmark by the TFRPD relative to MOX fuel behaviour. The corresponding PRIMO experimental data have been released, compiled and reviewed for the International Fuel Performance Experiments (IFPE) database. The observed ranges (as noted in the text) in the predicted thermal and FGR responses are reasonable given the variety and combination of thermal conductivity and FGR models employed by the benchmark participants with their respective fuel performance codes

  2. OECD/NEA Sandia Fuel Project phase I: Benchmark of the ignition testing

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina, E-mail: martina_adorni@hotmail.it [UNIPI (Italy); Herranz, Luis E. [CIEMAT (Spain); Hollands, Thorsten [GRS (Germany); Ahn, Kwang-II [KAERI (Korea, Republic of); Bals, Christine [GRS (Germany); D' Auria, Francesco [UNIPI (Italy); Horvath, Gabor L. [NUBIKI (Hungary); Jaeckel, Bernd S. [PSI (Switzerland); Kim, Han-Chul; Lee, Jung-Jae [KINS (Korea, Republic of); Ogino, Masao [JNES (Japan); Techy, Zsolt [NUBIKI (Hungary); Velazquez-Lozad, Alexander; Zigh, Abdelghani [USNRC (United States); Rehacek, Radomir [OECD/NEA (France)

    2016-10-15

    Highlights: • A unique PWR spent fuel pool experimental project is analytically investigated. • Predictability of fuel clad ignition in case of a complete loss of coolant in SFPs is assessed. • Computer codes reasonably estimate peak cladding temperature and time of ignition. - Abstract: The OECD/NEA Sandia Fuel Project provided unique thermal-hydraulic experimental data associated with Spent Fuel Pool (SFP) complete drain down. The study conducted at Sandia National Laboratories (SNL) was successfully completed (July 2009 to February 2013). The accident conditions of interest for the SFP were simulated in a full scale prototypic fashion (electrically heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate severe accident code validation and to reduce modeling uncertainties within the codes. Phase I focused on axial heating and burn propagation in a single PWR 17 × 17 assembly (i.e. “hot neighbors” configuration). Phase II addressed axial and radial heating and zirconium fire propagation including effects of fuel rod ballooning in a 1 × 4 assembly configuration (i.e. single, hot center assembly and four, “cooler neighbors”). This paper summarizes the comparative analysis regarding the final destructive ignition test of the phase I of the project. The objective of the benchmark is to evaluate and compare the predictive capabilities of computer codes concerning the ignition testing of PWR fuel assemblies. Nine institutions from eight different countries were involved in the benchmark calculations. The time to ignition and the maximum temperature are adequately captured by the calculations. It is believed that the benchmark constitutes an enlargement of the validation range for the codes to the conditions tested, thus enhancing the code applicability to other fuel assembly designs and configurations. The comparison of

  3. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  4. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  5. Burnable absorber-integrated Guide Thimble (BigT) - 1. Design concepts and neutronic characterization on the fuel assembly benchmarks

    International Nuclear Information System (INIS)

    Yahya, Mohd-Syukri; Yu, Hwanyeal; Kim, Yonghee

    2016-01-01

    This paper presents the conceptual designs of a new burnable absorber (BA) for the pressurized water reactor (PWR), which is named 'Burnable absorber-integrated Guide Thimble' (BigT). The BigT integrates BA materials into standard guide thimble in a PWR fuel assembly. Neutronic sensitivities and practical design considerations of the BigT concept are points of highlight in the first half of the paper. Specifically, the BigT concepts are characterized in view of its BA material and spatial self-shielding variations. In addition, the BigT replaceability requirement, bottom-end design specifications and thermal-hydraulic considerations are also deliberated. Meanwhile, much of the second half of the paper is devoted to demonstrate practical viability of the BigT absorbers via comparative evaluations against the conventional BA technologies in representative 17x17 and 16x16 fuel assembly lattices. For the 17x17 lattice evaluations, all three BigT variants are benchmarked against Westinghouse's existing BA technologies, while in the 16x16 assembly analyses, the BigT designs are compared against traditional integral gadolinia-urania rod design. All analyses clearly show that the BigT absorbers perform as well as the commercial BA technologies in terms of reactivity and power peaking management. In addition, it has been shown that sufficiently high control rod worth can be obtained with the BigT absorbers in place. All neutronic simulations were completed using the Monte Carlo Serpent code with ENDF/B-VII.0 library. (author)

  6. Minimizing the power peaking factor of fuel lattices using factors of group for boiling water reactors

    International Nuclear Information System (INIS)

    Guzman, J. R.; Longoria, L. C.; De la Cruz, E.; Arredondo, C.

    2010-10-01

    A method to design the distribution and composition of nuclear fuel for the array of fuel rods in a lattice for BWRs is presented in this work. The aim of the method is to minimize the power peaking factor until an adequate value is reached. Also, this method uses a few calculations of lattice. The method is based on the classification of the fuel rods in two groups: the group of fuel rods with the higher power level (group pow ), and the other group of fuel rods (no-group pow ). The enrichment of 235 U of each fuel rod of the group pow is multiplied by a factor called group fissile factor (f group ), and the enrichment of 235 U of each fuel rod of the no-group pow is multiplied by a factor called no-group fissile factor (f no-group ). These factors are fitted so that the power peaking factor is minimized. The importance of the method with the use of these two factors is applied to the design of a fuel lattice for BWRs as the Laguna Verde nuclear power plant. The calculations of lattice are made by means of the Helios code. (Author)

  7. Calculation Of A Lattice Physics Parameter For SBWR Fuel Bundle Design

    International Nuclear Information System (INIS)

    Sardjono, Y.

    1996-01-01

    The maximum power peaking factor for Nuclear Power Plant SBWR type is 1.5. The precision for that calculation is related with the result of unit cell analysis each rod in the fuel bundles. This analysis consist of lattice eigenvalue, lattice average diffusion cross section as well as relative power peaking factor in the fuel rod for each fuel bundles. The calculation by using TGBLA computer code which is based on the transport and 168 group diffusion theory. From this calculation can be concluded that the maximum relative power peaking factor is 1.304 and lower than design limit

  8. Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

    International Nuclear Information System (INIS)

    Bowman, S.M.

    1991-04-01

    To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as ''burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs

  9. Comparison of square and hexagonal fuel lattices for high conversion PWRs

    International Nuclear Information System (INIS)

    Kotlyar, D.; Shwageraus, E.

    2011-01-01

    This paper reports on an investigation into fuel design choices of a PWR operating in a self sustainable Th- 233 U fuel cycle. Achieving such self-sustainable with respect to fissile material fuel cycle would practically eliminate concerns over nuclear fuel supply hundreds of years into the future. Moreover, utilization of light water reactor technology and its associated vast experience would allow faster deployment of such fuel cycle without immediate need for development of fast reactor technology, which tends to be more complex and costly. In order to evaluate feasibility of this concept, two types of fuel assembly lattices were considered: square and hexagonal. The hexagonal lattice may offer some advantages over the square one. For example, the fertile blanket fuel can be packed more tightly reducing the blanket volume fraction in the core and potentially allowing to achieve higher core average power density. Furthermore, hexagonal lattice may allow more uniform leakage of neutrons from fissile to fertile regions and therefore more uniform neutron captures in thorium blanket. The calculations were carried out with Monte-Carlo based BGCore system, which includes neutronic, fuel depletion and thermo-hydraulic modules. The results were compared to those obtained from Serpent Monte-Carlo code and deterministic fuel assembly transport code BOXER. One of the major design challenges associated with the square seed-blanket concept is high power peaking due to the high concentration of fissile material in the seed region. In order to explore feasibility of the studied designs, the calculations were extended to include 3D fuel assembly analysis with thermal-hydraulic feedback. The coupled neutronic - thermal-hydraulic calculations were performed with BGCore code system. The analysis showed that both hexagonal and square seed-blanket fuel assembly designs have a potential of achieving net breeding. While no major neutronic advantages were observed for either fuel

  10. Validation of CENDL and JEFF evaluated nuclear data files for TRIGA calculations through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors

    International Nuclear Information System (INIS)

    Uddin, M.N.; Sarker, M.M.; Khan, M.J.H.; Islam, S.M.A.

    2009-01-01

    The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69-group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.

  11. OECD benchmark a of MOX fueled PWR unit cells using SAS2H, triton and mocup

    International Nuclear Information System (INIS)

    Ganda, F.; Greenspan, A.

    2005-01-01

    Three code systems are tested by applying them to calculate the OECD PWR MOX unit cell benchmark A. The codes tested are the SAS2H code sequence of the SCALE5 code package using 44 group library, MOCUP (MCNP4C + ORIGEN2), and the new TRITON depletion sequence of SCALE5 using 238 group cross sections generated using CENTRM with continuous energy cross sections. The burnup-dependent k ∞ and actinides concentration calculated by all three code-systems were found to be in good agreement with the OECD benchmark average results. Limited results were calculated also with the WIMS-ANL code package. WIMS-ANL was found to significantly under-predict k ∞ as well as the concentration of Pu 242 , consistently with the predictions of the WIMS-LWR reported by two of the OECD benchmark participants. Additionally, SAS2H is benchmarked against MOCUP for a hydride fuel containing unit cell, giving very satisfactory agreement. (authors)

  12. Effect of Pu-rich agglomerate in MOX fuel on a lattice calculation

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Yamamoto, Toru; Namekawa, Masakazu

    2007-01-01

    The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17x17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small. (author)

  13. Thorium-Based Fuels Preliminary Lattice Cell Studies for Candu Reactors

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Rizoiu, A.C.

    2009-01-01

    The choice of nuclear power as a major contributor to the future global energy needs must take into account acceptable risks of nuclear weapon proliferation, in addition to economic competitiveness, acceptable safety standards, and acceptable waste disposal options. Candu reactors offer a proven technology, safe and reliable reactor technology, with an interesting evolutionary potential for proliferation resistance, their versatility for various fuel cycles creating premises for a better utilization of global fuel resources. Candu reactors impressive degree of fuel cycle flexibility is a consequence of its channel design, excellent neutron economy, on-power refueling, and simple fuel bundle. These features facilitate the introduction and exploitation of various fuel cycles in Candu reactors in an evolutionary fashion. The main reasons for our interest in Thorium-based fuel cycles have been, globally, to extend the energy obtainable from natural Uranium and, locally, to provide a greater degree of energy self-reliance. Applying the once through Thorium (OTT) cycle in existing and advanced Candu reactors might be seen as an evaluative concept for the sustainable development both from the economic and waste management points of view. Two Candu fuel bundles project will be used for the proposed analysis, namely the Candu standard fuel bundle with 37 fuel elements and the CANFLEX fuel bundle with 43 fuel elements. Using the Canadian proposed scheme - loading mixed ThO 2 -SEU CANFLEX bundles in Candu 6 reactors - simulated at lattice cell level led to promising conclusions on operation at higher fuel burnups, reduction of the fissile content to the end of the cycle, minor actinide content reduction in the spent fuel, reduction of the spent fuel radiotoxicity, presence of radionuclides emitting strong gamma radiation for proliferation resistance benefit. The calculations were performed using the lattice codes WIMS and Dragon (together with the corresponding nuclear data

  14. Power Reactor Fuel Reprocessing Plant-2, Tarapur: a benchmark in Indian PHWR spent fuel reprocessing

    International Nuclear Information System (INIS)

    Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

    2017-01-01

    Power Reactor Fuel Reprocessing Plant-2 (PREFRE-2) is latest operating spent nuclear fuel reprocessing plant in India. This plant has improved design based on latest technology and feedback provided by the earlier plants. The design of PREFRE-2 plant is in five cycles of solvent extraction using TBP as extractant. The plant is commissioned in year 2011 after regulatory clearances

  15. Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4

    International Nuclear Information System (INIS)

    Martinez F, M.A.; Valle G, E. del; Alonso V, G.

    2007-01-01

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

  16. Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

    International Nuclear Information System (INIS)

    Kim, Taek Kyum; Downar, Thomas J.

    2002-01-01

    Research on the utilization of thorium-based fuels in the intermediate neutron spectrum of a tight-pitch light water reactor (LWR) lattice is reported. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thorium-based fuels in the intermediate spectrum of tight-pitch LWRs have considerable advantages in terms of conversion ratio, reactivity control, nonproliferation characteristics, and a reduced production of long-lived radiotoxic wastes. Because of the high conversion ratio of thorium-based fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11 to 17% higher than that of 238 U fertile fuels. However, unlike 238 U fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing 238 U with 232 Th in advanced high-conversion intermediate spectrum LWRs, such as the reduced-moderator reactor or the supercritical reactor

  17. Proposal and analysis of the benchmark problem suite for reactor physics study of LWR next generation fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-10-01

    In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI has established the Working Party on Reactor Physics for LWR Next Generation Fuels. The next generation fuels mean the ones aiming for further extended burn-up such as 70 GWd/t over the current design. The Working Party has proposed six benchmark problems, which consists of pin-cell, PWR fuel assembly and BWR fuel assembly geometries loaded with uranium and MOX fuels, respectively. The specifications of the benchmark problem neglect some of the current limitations such as 5 wt% {sup 235}U to achieve the above-mentioned target. Eleven organizations in the Working Party have carried out the analyses of the benchmark problems. As a result, status of accuracy with the current data and method and some problems to be solved in the future were clarified. In this report, details of the benchmark problems, result by each organization, and their comparisons are presented. (author)

  18. Simulations and measurements of adiabatic annular flows in triangular, tight lattice nuclear fuel bundle model

    Energy Technology Data Exchange (ETDEWEB)

    Saxena, Abhishek, E-mail: asaxena@lke.mavt.ethz.ch [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Zboray, Robert [Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Prasser, Horst-Michael [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

    2016-04-01

    High conversion light water reactors (HCLWR) having triangular, tight-lattice fuels bundles could enable improved fuel utilization compared to present day LWRs. However, the efficient cooling of a tight lattice bundle has to be still proven. Major concern is the avoidance of high-quality boiling crisis (film dry-out) by the use of efficient functional spacers. For this reason, we have carried out experiments on adiabatic, air-water annular two-phase flows in a tight-lattice, triangular fuel bundle model using generic spacers. A high-spatial-resolution, non-intrusive measurement technology, cold neutron tomography, has been utilized to resolve the distribution of the liquid film thickness on the virtual fuel pin surfaces. Unsteady CFD simulations have also been performed to replicate and compare with the experiments using the commercial code STAR-CCM+. Large eddies have been resolved on the grid level to capture the dominant unsteady flow features expected to drive the liquid film thickness distribution downstream of a spacer while the subgrid scales have been modeled using the Wall Adapting Local Eddy (WALE) subgrid model. A Volume of Fluid (VOF) method, which directly tracks the interface and does away with closure relationship models for interfacial exchange terms, has also been employed. The present paper shows first comparison of the measurement with the simulation results.

  19. Investigation of fuel lattice pitch changes influence on reactor performance through evaluate the neutronic parameters

    International Nuclear Information System (INIS)

    Zareian Ronizi, F.; Fadaei, A.H.; Setayeshi, S.; Shahidi, A.R.

    2015-01-01

    Highlights: • One of the most complex issues that Nu-engineers deal with is the design of NR core. • Numerous factors in nuclear core design depend on Fuel-to-Moderator volume ratio. • Aim of this research is to investigate RX performance for different lattice pitches. - Abstract: Nuclear reactor core design is one of the most complex issues that nuclear engineers deal with. The number and complexity of effective parameters and their impact on reactor design, which makes the problem difficult to solve, require precise knowledge of these parameters and their influence on the reactor operation. Numerous factors in a nuclear reactor core design depend on the Fuel-to-Moderator volume ratio, V F /V M , in a fuel cell. This ratio can be modified by changing the lattice pitch which is the thickness of water channels between fuels plates while keeping fuel slab dimensions fixed. Cooling and moderating properties of water are affected by such a change in a reactor core, and hence some parameters related to these properties might be changed. The aim of this research is to provide the suitable knowledge for nuclear core designing. To reach this goal, the first operating core of Tehran Research Reactor (TRR) with different lattice pitches is simulated, and the effect of different lattice pitches on some parameters such as effective multiplication factor (K eff ), reactor life time, distribution of neutron flux and power density in the core, as well as moderator temperature and density coefficient of reactivity are evaluated. The nuclear reactor analysis code, MTR-PC package is employed to carry out the considered calculation. Finally, the results are presented in some tables and graphs that provide useful information for nuclear engineers in the nuclear reactor core design

  20. Validation study of SRAC2006 code system based on evaluated nuclear data libraries for TRIGA calculations by benchmarking integral parameters of TRX and BAPL lattices of thermal reactors

    International Nuclear Information System (INIS)

    Khan, M.J.H.; Sarker, M.M.; Islam, S.M.A.

    2013-01-01

    Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations

  1. Numerical analysis of the reactivity for the dry lattices above the water level of the critical fuel cores

    International Nuclear Information System (INIS)

    Nauchi, Yasushi; Kameyama, Takanori

    2003-01-01

    Criticality analysis has been performed for dozens of tank type cores in which fuel lattices are loaded vertically and partially immersed in light water. The reactivity effect of dry part of lattices stuck above the critical water level has been calculated using the continuous energy Monte Carlo method. The reactivity effect exceeds 0.8% both for MOX and UOX fuel lattices of large buckling (B z 2 > 0.0025 cm -2 ). It is evaluated that at least 20 cm length of fuel rods above the critical water level has significant reactivity effect. (author)

  2. Proton Exchange Membrane Fuel Cell Engineering Model Powerplant. Test Report: Benchmark Tests in Three Spatial Orientations

    Science.gov (United States)

    Loyselle, Patricia; Prokopius, Kevin

    2011-01-01

    Proton exchange membrane (PEM) fuel cell technology is the leading candidate to replace the aging alkaline fuel cell technology, currently used on the Shuttle, for future space missions. This test effort marks the final phase of a 5-yr development program that began under the Second Generation Reusable Launch Vehicle (RLV) Program, transitioned into the Next Generation Launch Technologies (NGLT) Program, and continued under Constellation Systems in the Exploration Technology Development Program. Initially, the engineering model (EM) powerplant was evaluated with respect to its performance as compared to acceptance tests carried out at the manufacturer. This was to determine the sensitivity of the powerplant performance to changes in test environment. In addition, a series of tests were performed with the powerplant in the original standard orientation. This report details the continuing EM benchmark test results in three spatial orientations as well as extended duration testing in the mission profile test. The results from these tests verify the applicability of PEM fuel cells for future NASA missions. The specifics of these different tests are described in the following sections.

  3. Reactivity and reaction rate measurements in U--D2O lattices with coaxial fuel

    International Nuclear Information System (INIS)

    Pellarin, D.J.; Morris, B.M.

    1976-12-01

    Integral reaction rate parameters, intracell thermal neutron flux profiles, and material bucklings were measured for D 2 O-moderated uniform lattices in the exponential facility at the Savannah River Laboratory. Two different slightly enriched coaxial uranium fuel assemblies were examined over a wide range of triangular lattice pitches. Integral parameters are reported for inner and outer fuel separately providing data for a more detailed and rigorous comparison with computation than has been previously available. Results are compared with RAHAB calculations using ENDF/B-IV cross sections. Large discrepancies in agreement between calculation and experiment, outside of experimental errors and uncertainties in the input cross sections, probably result from the resonance capture models used by RAHAB

  4. Evaluation of temperature coefficients of reactivity for 233U--thorium fueled HTGR lattices. Final report

    International Nuclear Information System (INIS)

    Newman, D.F.; Leonard, B.R. Jr.; Trapp, T.J.; Gore, B.F.; Kottwitz, D.A.; Thompson, J.K.; Purcell, W.L.; Stewart, K.B.

    1977-05-01

    A comparison of calculated and measured neutron multiplication factors as a function of temperature was made for three graphite-moderated lattices in the High Temperature Lattice Test Reactor (HTLTR) using 233 UO 2 --ThO 2 fuels in varying amounts and configurations. Correlation of neutronic analysis methods and cross section data with the experimental measurements forms the basis for assessing the accuracy of the methods and data and developing confidence in the ability to predict the temperature coefficient of reactivity for various High Temperature Gas-Cooled Reactor (HTGR) conditions in which 233 U and thorium are present in the fuel. The calculated values of k/sub infinity/(T) were correlated with measured values using two least-squares-fitted correlation coefficients: (1) a normalization factor, and (2) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross section data

  5. Benchmark reference data on post irradiation analysis of light water reactor fuel samples

    International Nuclear Information System (INIS)

    Guardini, S.; Guzzi, G.

    1983-01-01

    The structure of the present report is as follows: in section I the benchmark activity (BM) is described in detail; characteristics of the reactors and fuel assemblies examinated are given, and the technical aspects of the chemical and analytical processes are discussed. In section II all the techniques used to certify the analytical data are presented, together with a discussion of evaluated random and systematic uncertainties. A comparison with the calculated values and the interpretation with ICT (Isotopic Correlation Techniques) is also presented in this section. Section III presents the results. In practice the complete sets of results referring to all JRC measurements are given here for the sake of the completeness and consistency of this final report

  6. Study of the lattice parameter evolution of PWR irradiated MOX fuel by X-Ray diffraction

    International Nuclear Information System (INIS)

    Clavier, B.

    1995-01-01

    Fuel irradiation leads to a swelling resulting from the formation of gaseous (Kr, Xe) or solid fission products which are found either in solution or as solid inclusions in the matrix. This phenomena has to be evaluated to be taken into account in fuel cladding Interaction. Fuel swelling was studied as a function of burn up by measuring the corresponding cell constant evolution by X-Ray diffraction. This study was realized on Mixed Oxide Fuels (MOX) irradiated in a Pressurized Water Reactor (PWR) at different burn-up for 3 initial Pu contents. Lattice parameter evolutions were followed as a function of burn-up for the irradiated fuel with and without an annealing thermal treatment. These experimental evolutions are compared to the theoretical evolutions calculated from the hard sphere model, using the fission product concentrations determined by the APPOLO computer code. Contribution of varying parameters influencing the unit cell value is discussed. Thermal treatment effects were checked by metallography, X-Ray diffraction and microprobe analysis. After thermal treatment, no structural change was observed but a decrease of the lattice parameter was measured. This modification results essentially from self-irradiation defect annealing and not from stoichiometry variations. Microprobe analysis showed that about 15% of the formed Molybdenum is in solid solution In the oxide matrix. Micrographs showed the existence of Pu packs in the oxide matrix which induces a broadening of diffraction lines. The RIETVELD method used to analyze the X-Ray patterns did not allow to characterize independently the Pu packs and the oxide matrix lattice parameters. Nevertheless, with this method, the presence of micro-strains in the irradiated nuclear fuel could be confirmed. (author)

  7. A parametric thermohydraulic study an advanced pressurized light water reactor with a tight fuel rod lattice

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Hame, W.

    1982-12-01

    A parametric thermohydraulic study for an Advanced Pressurized Light Water Reactor (APWR) with a tight fuel rod lattice has been performed. The APWR improves the uranium utilisation. The APWR core should be placed in a modern German PWR plant. Within this study about 200 different reactors have been calculated. The tightening of the fuel rod lattice implies a decrease of the net electrical output of the plant, which is greater for the heterogeneous reactor than for the homogeneous reactor. APWR cores mean higher core pressure drops and higher water velocities in the core region. The cores tend to be shorter and the number of fuel rods to be higher than for the PWR. At the higher fuel rod pitch to diameter ratios (p/d) the DNB limitation is more stringent than the limitation on the fuel rod linear rating given by the necessity of reflooding after a reactor accident. The contrary is true for the lower p/d ratios. Subcooled boiling in the highest rated coolant channels occurs for the most of the calculated reactors. (orig.) [de

  8. EVALUATION OF U10MO FUEL PLATE IRRADIATION BEHAVIOR VIA NUMERICAL AND EXPERIMENTAL BENCHMARKING

    Energy Technology Data Exchange (ETDEWEB)

    Samuel J. Miller; Hakan Ozaltun

    2012-11-01

    This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which result in 3-dimensional creep behavior that is a product of these components. Results show that assumptions made in 2D models for the out-of-plane stresses and strains cannot capture the 3-dimensional physics accurately and thus 2D approximations are not computationally accurate. Stress-strain fields are dependent on plate geometry and irradiation conditions, thus, if stress based criteria is used to predict plate behavior (as opposed to material impurities, fine micro-structural defects, or sharp power gradients), unique 3D finite element formulation for each plate is required.

  9. Square lattice honeycomb tri-carbide fuels for 50 to 250 KN variable thrust NTP design

    International Nuclear Information System (INIS)

    Anghaie, Samim; Knight, Travis; Gouw, Reza; Furman, Eric

    2001-01-01

    Ultrahigh temperature solid solution of tri-carbide fuels are used to design an ultracompact nuclear thermal rocket generating 950 seconds of specific impulse with scalable thrust level in range of 50 to 250 kilo Newtons. Solid solutions of tri-carbide nuclear fuels such as uranium-zirconium-niobium carbide. UZrNbC, are processed to contain certain mixing ratio between uranium carbide and two stabilizing carbides. Zirconium or niobium in the tri-carbide could be replaced by tantalum or hafnium to provide higher chemical stability in hot hydrogen environment or to provide different nuclear design characteristics. Recent studies have demonstrated the chemical compatibility of tri-carbide fuels with hydrogen propellant for a few to tens of hours of operation at temperatures ranging from 2800 K to 3300 K, respectively. Fuel elements are fabricated from thin tri-carbide wafers that are grooved and locked into a square-lattice honeycomb (SLHC) shape. The hockey puck shaped SLHC fuel elements are stacked up in a grooved graphite tube to form a SLHC fuel assembly. A total of 18 fuel assemblies are arranged circumferentially to form two concentric rings of fuel assemblies with zirconium hydride filling the space between assemblies. For 50 to 250 kilo Newtons thrust operations, the reactor diameter and length including reflectors are 57 cm and 60 cm, respectively. Results of the nuclear design and thermal fluid analyses of the SLHC nuclear thermal propulsion system are presented

  10. Calculation of the Thermal Radiation Benchmark Problems for a CANDU Fuel Channel Analysis Using the CFX-10 Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

    2006-07-15

    To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to whole fuel channel geometry. With assumptions of a non-participating medium completely enclosed with the diffuse, gray and opaque surfaces, the solutions of the benchmark problems are obtained by the concept of surface resistance to radiation accounting for the view factors and the emissivities. The view factors are calculated by the program MATRIX version 1.0 avoiding the difficulty of hand calculation for the complex geometries. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer.

  11. Calculation of the Thermal Radiation Benchmark Problems for a CANDU Fuel Channel Analysis Using the CFX-10 Code

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

    2006-07-01

    To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to whole fuel channel geometry. With assumptions of a non-participating medium completely enclosed with the diffuse, gray and opaque surfaces, the solutions of the benchmark problems are obtained by the concept of surface resistance to radiation accounting for the view factors and the emissivities. The view factors are calculated by the program MATRIX version 1.0 avoiding the difficulty of hand calculation for the complex geometries. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer

  12. Automatic fuel lattice design in a boiling water reactor using a particle swarm optimization algorithm and local search

    International Nuclear Information System (INIS)

    Lin Chaung; Lin, Tung-Hsien

    2012-01-01

    Highlights: ► The automatic procedure was developed to design the radial enrichment and gadolinia (Gd) distribution of fuel lattice. ► The method is based on a particle swarm optimization algorithm and local search. ► The design goal were to achieve the minimum local peaking factor. ► The number of fuel pins with Gd and Gd concentration are fixed to reduce search complexity. ► In this study, three axial sections are design and lattice performance is calculated using CASMO-4. - Abstract: The axial section of fuel assembly in a boiling water reactor (BWR) consists of five or six different distributions; this requires a radial lattice design. In this study, an automatic procedure based on a particle swarm optimization (PSO) algorithm and local search was developed to design the radial enrichment and gadolinia (Gd) distribution of the fuel lattice. The design goals were to achieve the minimum local peaking factor (LPF), and to come as close as possible to the specified target average enrichment and target infinite multiplication factor (k ∞ ), in which the number of fuel pins with Gd and Gd concentration are fixed. In this study, three axial sections are designed, and lattice performance is calculated using CASMO-4. Finally, the neutron cross section library of the designed lattice is established by CMSLINK; the core status during depletion, such as thermal limits, cold shutdown margin and cycle length, are then calculated using SIMULATE-3 in order to confirm that the lattice design satisfies the design requirements.

  13. Summary report on the international comparison of NEACRP burnup benchmark calculations for high conversion light water reactor lattices

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Ishiguro, Yukio; Takano, Hideki

    1988-10-01

    The results of the NEACRP HCLWR cell burnup benchmark calculations are summarized in this report. Fifteen organizations from eight countries participated in this benchmark and submitted twenty solutions. Large differences are still observed among the calculated values of void reactivities and conversion ratios. These differences are mainly caused from the discrepancies in the reaction rates of U-238, Pu-239 and fission products. The physics problems related to these results are briefly investigated in the report. In the specialists' meeting on this benchmark calculations held in April 1988, it was recommended to perform continuous energy Monte Carlo calculations in order to obtain reference solutions for design codes. The conclusions resulted from the specialists' meeting are also presented. (author)

  14. A 3D stylized half-core CANDU benchmark problem

    International Nuclear Information System (INIS)

    Pounders, Justin M.; Rahnema, Farzad; Serghiuta, Dumitru; Tholammakkil, John

    2011-01-01

    A 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem is presented. The benchmark problem is comprised of a heterogeneous lattice of 37-element natural uranium fuel bundles, heavy water moderated, heavy water cooled, with adjuster rods included as reactivity control devices. Furthermore, a 2-group macroscopic cross section library has been developed for the problem to increase the utility of this benchmark for full-core deterministic transport methods development. Monte Carlo results are presented for the benchmark problem in cooled, checkerboard void, and full coolant void configurations.

  15. Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E., E-mail: emerzari@anl.gov [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Fischer, P. [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Yuan, H. [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL (United States); Van Tichelen, K.; Keijers, S. [SCK-CEN, Boeretang 200, Mol (Belgium); De Ridder, J.; Degroote, J.; Vierendeels, J. [Ghent University, Ghent (Belgium); Doolaard, H.; Gopala, V.R.; Roelofs, F. [NRG, Petten (Netherlands)

    2016-03-15

    Highlights: • A EUROTAM-US INERI consortium has performed a benchmark exercise related to fast reactor assembly simulations. • LES calculations for a wire-wrapped rod bundle are compared with RANS calculations. • Results show good agreement for velocity and cross flows. - Abstract: As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

  16. VERA Pin and Fuel Assembly Depletion Benchmark Calculations by McCARD and DeCART

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin; Cho, Jin Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Monte Carlo (MC) codes have been developed and used to simulate a neutron transport since MC method was devised in the Manhattan project. Solving the neutron transport problem with the MC method is simple and straightforward to understand. Because there are few essential approximations for the 6- dimension phase of a neutron such as the location, energy, and direction in MC calculations, highly accurate solutions can be obtained through such calculations. In this work, the VERA pin and fuel assembly (FA) depletion benchmark calculations are performed to examine the depletion capability of the newly generated DeCART multi-group cross section library. To obtain the reference solutions, MC depletion calculations are conducted using McCARD. Moreover, to scrutinize the effect by stochastic uncertainty propagation, uncertainty propagation analyses are performed using a sensitivity and uncertainty (S/U) analysis method and stochastic sampling (S.S) method. It is still expensive and challenging to perform a depletion analysis by a MC code. Nevertheless, many studies and works for a MC depletion analysis have been conducted to utilize the benefits of the MC method. In this study, McCARD MC and DeCART MOC transport calculations are performed for the VERA pin and FA depletion benchmarks. The DeCART depletion calculations are conducted to examine the depletion capability of the newly generated multi-group cross section library. The DeCART depletion calculations give excellent agreement with the McCARD reference one. From the McCARD results, it is observed that the MC depletion results depend on how to split the burnup interval. First, only to quantify the effect of the stochastic uncertainty propagation at 40 DTS, the uncertainty propagation analyses are performed using the S/U and S.S. method.

  17. Benchmark for the qualification of gamma shielding calculation methods for light-water type reactor spent fuels

    International Nuclear Information System (INIS)

    Blum, P.; Cagnon, R.; Nimal, J.C.

    1982-01-01

    This report gives the results of a campaign of gamma dose rates measurement in the vicinity of a transport package loaded with 12 PWR spent fuel assemblies, so that the characteristics of the package and the fuel. It describes the measuring methods, and gives the accuracy of the data which will be usefull, as benchmarks, to the control of the calculation methods used to verify the gamma shielding of the packages. It shows how to calculate gamma dose rates from the data given on the package and the fuel, and gives the results of a calculation with the Mecure IV code and compares them to the measurements

  18. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  19. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-01-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  20. Simulation of sound waves using the Lattice Boltzmann Method for fluid flow: Benchmark cases for outdoor sound propagation

    NARCIS (Netherlands)

    Salomons, E.M.; Lohman, W.J.A.; Zhou, H.

    2016-01-01

    Propagation of sound waves in air can be considered as a special case of fluid dynamics. Consequently, the lattice Boltzmann method (LBM) for fluid flow can be used for simulating sound propagation. In this article application of the LBM to sound propagation is illustrated for various cases:

  1. Numerical simulation of direct methanol fuel cells using lattice Boltzmann method

    Energy Technology Data Exchange (ETDEWEB)

    Delavar, Mojtaba Aghajani; Farhadi, Mousa; Sedighi, Kurosh [Faculty of Mechanical Engineering, Babol University of Technology, Babol, P.O. Box 484 (Iran)

    2010-09-15

    In this study Lattice Boltzmann Method (LBM) as an alternative of conventional computational fluid dynamics method is used to simulate Direct Methanol Fuel Cell (DMFC). A two dimensional lattice Boltzmann model with 9 velocities, D2Q9, is used to solve the problem. The computational domain includes all seven parts of DMFC: anode channel, catalyst and diffusion layers, membrane and cathode channel, catalyst and diffusion layers. The model has been used to predict the flow pattern and concentration fields of different species in both clear and porous channels to investigate cell performance. The results have been compared well with results in literature for flow in porous and clear channels and cell polarization curves of the DMFC at different flow speeds and feed methanol concentrations. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-31

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

  3. The use of irradiation zones in the fuel element lattice of research piles

    International Nuclear Information System (INIS)

    Delattre, P.; Genthon, J.P.

    1960-01-01

    The first part of this report examines the advantages and disadvantages of the various types of canal which may be found in the fuel element lattice of research piles. Some examples relative to the piles EL2 and EL3 are discussed in detail. From the conclusions drawn in the last part, several norms are extracted which make it possible to define the conditions the various canals must fulfil in order that they can be used to the best possible advantage for each type of irradiation. (author) [fr

  4. Analysis of neutronics benchmarks for the utilization of mixed oxide fuel in light water reactor using DRAGON code

    International Nuclear Information System (INIS)

    Nithyadevi, Rajan; Thilagam, L.; Karthikeyan, R.; Pal, Usha

    2016-01-01

    Highlights: • Use of advanced computational code – DRAGON-5 using advanced self shielding model USS. • Testing the capability of DRAGON-5 code for the analysis of light water reactor system. • Wide variety of fuels LEU, MOX and spent fuel have been analyzed. • Parameters such as k ∞ , one, few and multi-group macroscopic cross-sections and fluxes were calculated. • Suitability of deterministic methodology employed in DRAGON-5 code is demonstrated for LWR. - Abstract: Advances in reactor physics have led to the development of new computational technologies and upgraded cross-section libraries so as to produce an accurate approximation to the true solution for the problem. Thus it is necessary to revisit the benchmark problems with the advanced computational code system and upgraded cross-section libraries to see how far they are in agreement with the earlier reported values. Present study is one such analysis with the DRAGON code employing advanced self shielding models like USS and 172 energy group ‘JEFF3.1’ cross-section library in DRAGLIB format. Although DRAGON code has already demonstrated its capability for heavy water moderator systems, it is now tested for light water reactor (LWR) and fast reactor systems. As a part of validation of DRAGON for LWR, a VVER computational benchmark titled “Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel-Volume 3” submitted by the Russian Federation has been taken up. Presently, pincell and assembly calculations are carried out considering variation in fuel temperature (both fresh and spent), moderator temperatures and boron content in the moderator. Various parameters such as infinite neutron multiplication (k ∞ ) factor, one group integrated flux, few group homogenized cross-sections (absorption, nu-fission) and reaction rates (absorption, nu-fission) of individual isotopic nuclides are calculated for different reactor states. Comparisons of results are made with the reported Monte Carlo

  5. Consultants' Meeting on Review Benchmarking of Nuclear Data for the Th/U Fuel Cycle. Summary Report

    International Nuclear Information System (INIS)

    Capote Noy, R.

    2011-02-01

    A summary is given of the Consultants' Meeting (CM) on Review and Benchmarking of Nuclear Data for the Th/U Fuel Cycle. An IAEA Coordinated Research Project (CRP) on 'Nuclear Data for Th/U Fuel Cycle' was concluded in 2005. The CRP activities resulted in new evaluated nuclear data files for 232 Th, 231 , 233 Pa (later adopted for the ENDF/B-VII.0 library) and improvements to existing evaluations for 232 , 233 , 234 , 236 U. Available nuclear data evaluations for 230 - 232 Th, 231,233 Pa and 232 , 233 , 234 U were reviewed including ROSFOND2010, CENDL-3.1, JENDL-4, JEFF-3.1.1, MINSKACT, and ENDF/B-VII.0 libraries. Benchmark results of available evaluations for 232 Th and 233 U were also discussed. (author)

  6. Lattice Boltzmann modeling of transport phenomena in fuel cells and flow batteries

    Science.gov (United States)

    Xu, Ao; Shyy, Wei; Zhao, Tianshou

    2017-06-01

    Fuel cells and flow batteries are promising technologies to address climate change and air pollution problems. An understanding of the complex multiscale and multiphysics transport phenomena occurring in these electrochemical systems requires powerful numerical tools. Over the past decades, the lattice Boltzmann (LB) method has attracted broad interest in the computational fluid dynamics and the numerical heat transfer communities, primarily due to its kinetic nature making it appropriate for modeling complex multiphase transport phenomena. More importantly, the LB method fits well with parallel computing due to its locality feature, which is required for large-scale engineering applications. In this article, we review the LB method for gas-liquid two-phase flows, coupled fluid flow and mass transport in porous media, and particulate flows. Examples of applications are provided in fuel cells and flow batteries. Further developments of the LB method are also outlined.

  7. Sensitivity and Uncertainty Analysis for coolant void reactivity in a CANDU Fuel Lattice Cell Model

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Seung Yeol; Shim, Hyung Jin [Seoul National University, Seoul (Korea, Republic of)

    2016-10-15

    In this study, the EPBM is implemented in Seoul National university Monte Carlo (MC) code, McCARD which has the k uncertainty evaluation capability by the adjoint-weighted perturbation (AWP) method. The implementation is verified by comparing the sensitivities of the k-eigenvalue difference to the microscopic cross sections computed by the DPBM and the direct subtractions for the TMI-1 pin-cell problem. The uncertainty of the coolant void reactivity (CVR) in a CANDU fuel lattice model due to the ENDF/B-VII.1 covariance data is calculated by its sensitivities estimated by the EPBM. The method based on the eigenvalue perturbation theory (EPBM) utilizes the 1st order adjoint-weighted perturbation (AWP) technique to estimate the sensitivity of the eigenvalue difference. Furthermore this method can be easily applied in a S/U analysis code system equipped with the eigenvalue sensitivity calculation capability. The EPBM is implemented in McCARD code and verified by showing good agreement with reference solution. Then the McCARD S/U analysis have been performed with the EPBM module for the CVR in CANDU fuel lattice problem. It shows that the uncertainty contributions of nu of {sup 235}U and gamma reaction of {sup 238}U are dominant.

  8. The added value of international benchmarks for fuel performance codes: an illustration on the basis of TRANSURANUS

    International Nuclear Information System (INIS)

    Van Uffelen, P.; Schubert, A.; Gyeori, C.; Van De Laar, J.

    2009-01-01

    Safety authorities and fuel designers, as well as nuclear research centers rely heavily on fuel performance codes for predicting the behaviour and life-time of fuel rods. The simulation tools are developed and validated on the basis of experimental results, some of which is in the public domain such as the International Fuel Performance Experiments database of the OECD/NEA and IAEA. Publicly available data constitute an excellent basis for assessing codes themselves, but also to compare codes that are being developed by independent teams. The present report summarises the advantages for the TRANSURANUS code by taking part in previous benchmarks organised by the IAEA, and outlines the preliminary results along with the perspectives of our participation in the current coordinated research project FUMEXIII

  9. Simulation of Sound Waves Using the Lattice Boltzmann Method for Fluid Flow: Benchmark Cases for Outdoor Sound Propagation.

    Science.gov (United States)

    Salomons, Erik M; Lohman, Walter J A; Zhou, Han

    2016-01-01

    Propagation of sound waves in air can be considered as a special case of fluid dynamics. Consequently, the lattice Boltzmann method (LBM) for fluid flow can be used for simulating sound propagation. In this article application of the LBM to sound propagation is illustrated for various cases: free-field propagation, propagation over porous and non-porous ground, propagation over a noise barrier, and propagation in an atmosphere with wind. LBM results are compared with solutions of the equations of acoustics. It is found that the LBM works well for sound waves, but dissipation of sound waves with the LBM is generally much larger than real dissipation of sound waves in air. To circumvent this problem it is proposed here to use the LBM for assessing the excess sound level, i.e. the difference between the sound level and the free-field sound level. The effect of dissipation on the excess sound level is much smaller than the effect on the sound level, so the LBM can be used to estimate the excess sound level for a non-dissipative atmosphere, which is a useful quantity in atmospheric acoustics. To reduce dissipation in an LBM simulation two approaches are considered: i) reduction of the kinematic viscosity and ii) reduction of the lattice spacing.

  10. An Evaluation of Criticality Margin by an Application of Parallelogram Lattice Arrangement in the Nuclear Fuel Storage Rack

    International Nuclear Information System (INIS)

    Kim, Song Hyun; Kim, Hong Chul; Shin, Chang Ho; Kim, Jong Kyung; Kim, Kyo Youn

    2010-01-01

    The criticality evaluation in the nuclear fuel storage rack is essentially required for the prevention of the criticality accident. The square lattice structure of the storage rack is commonly used because it has a simple structure for the storage of the numerous fuel assemblies as well as the good mechanical strength. For the design of the fuel storage rack, the boron plate is commonly used for the criticality reduction. In this study, an arrangement method with the parallelogram lattice structure is proposed for the reduction of the boron concentration or the rack pitch. The criticality margins by the application of the parallelogram lattice were evaluated with MCNP5 code. From the result, the reduction of the boron concentrated in the borated-Al plate was evaluated

  11. A subchannel and CFD analysis of void distribution for the BWR fuel bundle test benchmark

    International Nuclear Information System (INIS)

    In, Wang-Kee; Hwang, Dae-Hyun; Jeong, Jae Jun

    2013-01-01

    Highlights: ► We analyzed subchannel void distributions using subchannel, system and CFD codes. ► The mean error and standard deviation at steady states were compared. ► The deviation of the CFD simulation was greater than those of the others. ► The large deviation of the CFD prediction is due to interface model uncertainties. -- Abstract: The subchannel grade and microscopic void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility have been evaluated with a subchannel analysis code MATRA, a system code MARS and a CFD code CFX-10. Sixteen test series from five different test bundles were selected for the analysis of the steady-state subchannel void distributions. Four test cases for a high burn-up 8 × 8 fuel bundle with a single water rod were simulated using CFX-10 for the microscopic void distribution benchmark. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5 to 25%. The results of the transient calculations were also similar to each other and very reasonable. The CFD simulation reproduced the overall radial void distribution trend which produces less vapor in the central part of the bundle and more vapor in the periphery. However, the predicted variation of the void distribution inside the subchannels is small, while the measured one is large showing a very high concentration in the center of the subchannels. The variations of the void distribution between the center of the subchannels and the subchannel gap are estimated to be about 5–10% for the CFD prediction and more than 20% for the experiment

  12. Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air

    Energy Technology Data Exchange (ETDEWEB)

    Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2004-10-18

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called “channeling” effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute’s Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice.Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins

  13. Pilot-benchmarking of the WENRA safety reference levels for the spent fuel intermediate storage facility Ahaus

    International Nuclear Information System (INIS)

    Lorenz, Bernd; Roeder, Markus; Brandt, Klaus-Dieter

    2008-01-01

    Full text: The Western European Nuclear Regulator's Association (WENRA) has 2007 issued the draft of the 'Waste and Spent Fuel Storage Safety Reference Levels'. The objective of WENRA is to strive for a harmonized safety level of nuclear facilities within the European Community and these Reference Levels are a benchmark method to demonstrate the achieved level for the regulatory system and the implementation as well. Safety Reference Levels exist at the moment for Reactor Safety, Waste Storage and Decommissioning in different stages of development. ENISS, the European Nuclear Installations Safety Standards Initiative, a FORATOM based special organisation of nuclear operators, has discussed these Safety Reference Levels very intensively with WENRA and the agreement was to make a implementation benchmark-exercise for the storage facilities before the authorities finally agree on the Reference Levels. This benchmark was scheduled for the year 2008. Because of the special situation in Germany where a large number of storage facilities is in operation the German authorities felt that it would be useful to initiate a Pilot-Benchmark to get first results on the feasibility of the Reference Levels and the burden imposed to authorities and operators by these benchmark-exercises. GNS, a subsidiary company of the utilities, agreed to step into this process on a voluntary basis with its storage facility for spent fuel in Ahaus. The exercise was done in a very efficient way and in good co-operation between the authorities, local and federal, and the operator. The results in terms of safety assessments have been very satisfactory showing the high degree of safety. Although the facility was for the first time licensed already in 1987 the compliance with nearly all Reference Levels from 2007 could be demonstrated. It became also clear that newer facilities would fulfil the desired safety standard too. Nevertheless, in spite of the good results the exercise revealed some weak

  14. A practical optimization procedure for radial BWR fuel lattice design using tabu search with a multiobjective function

    International Nuclear Information System (INIS)

    Francois, J.L.; Martin-del-Campo, C.; Francois, R.; Morales, L.B.

    2003-01-01

    An optimization procedure based on the tabu search (TS) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The procedure was coded in a computing system in which the optimization code uses the tabu search method to select potential solutions and the HELIOS code to evaluate them. The goal of the procedure is to search for an optimal fuel utilization, looking for a lattice with minimum average enrichment, with minimum deviation of reactivity targets and with a local power peaking factor (PPF) lower than a limit value. Time-dependent-depletion (TDD) effects were considered in the optimization process. The additive utility function method was used to convert the multiobjective optimization problem into a single objective problem. A strategy to reduce the computing time employed by the optimization was developed and is explained in this paper. An example is presented for a 10x10 fuel lattice with 10 different fuel compositions. The main contribution of this study is the development of a practical TDD optimization procedure for BWR fuel lattice design, using TS with a multiobjective function, and a strategy to economize computing time

  15. EPRI depletion benchmark calculations using PARAGON

    International Nuclear Information System (INIS)

    Kucukboyaci, Vefa N.

    2015-01-01

    Highlights: • PARAGON depletion calculations are benchmarked against the EPRI reactivity decrement experiments. • Benchmarks cover a wide range of enrichments, burnups, cooling times, and burnable absorbers, and different depletion and storage conditions. • Results from PARAGON-SCALE scheme are more conservative relative to the benchmark data. • ENDF/B-VII based data reduces the excess conservatism and brings the predictions closer to benchmark reactivity decrement values. - Abstract: In order to conservatively apply burnup credit in spent fuel pool criticality analyses, code validation for both fresh and used fuel is required. Fresh fuel validation is typically done by modeling experiments from the “International Handbook.” A depletion validation can determine a bias and bias uncertainty for the worth of the isotopes not found in the fresh fuel critical experiments. Westinghouse’s burnup credit methodology uses PARAGON™ (Westinghouse 2-D lattice physics code) and its 70-group cross-section library, which have been benchmarked, qualified, and licensed both as a standalone transport code and as a nuclear data source for core design simulations. A bias and bias uncertainty for the worth of depletion isotopes, however, are not available for PARAGON. Instead, the 5% decrement approach for depletion uncertainty is used, as set forth in the Kopp memo. Recently, EPRI developed a set of benchmarks based on a large set of power distribution measurements to ascertain reactivity biases. The depletion reactivity has been used to create 11 benchmark cases for 10, 20, 30, 40, 50, and 60 GWd/MTU and 3 cooling times 100 h, 5 years, and 15 years. These benchmark cases are analyzed with PARAGON and the SCALE package and sensitivity studies are performed using different cross-section libraries based on ENDF/B-VI.3 and ENDF/B-VII data to assess that the 5% decrement approach is conservative for determining depletion uncertainty

  16. The new deterministic 3-D radiation transport code Multitrans: C5G7 MOX fuel assembly benchmark

    International Nuclear Information System (INIS)

    Kotiluoto, P.

    2003-01-01

    The novel deterministic three-dimensional radiation transport code MultiTrans is based on combination of the advanced tree multigrid technique and the simplified P3 (SP3) radiation transport approximation. In the tree multigrid technique, an automatic mesh refinement is performed on material surfaces. The tree multigrid is generated directly from stereo-lithography (STL) files exported by computer-aided design (CAD) systems, thus allowing an easy interface for construction and upgrading of the geometry. The deterministic MultiTrans code allows fast solution of complicated three-dimensional transport problems in detail, offering a new tool for nuclear applications in reactor physics. In order to determine the feasibility of a new code, computational benchmarks need to be carried out. In this work, MultiTrans code is tested for a seven-group three-dimensional MOX fuel assembly transport benchmark without spatial homogenization (NEA C5G7 MOX). (author)

  17. Effect of lattice deformation on temperature fields and heat transfer in the fuel elements of characteristic zones for a model of fast reactor fuel assembly

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Matyukhin, N.M.; Sviridenko, E.Ya.

    1980-01-01

    Given are the experimental results for temperature fields in the model assembly in nonribbed simulators of the BN-600-type reactor fuel elements in the course of deformation of the lattice caused by shifting of the central and peripheral (lateral, angular) fuel elements by the value of the gap between the fuel elements (the limiting case when the fuel elements touch each other along the whole length). An assembly consisting of 37 electroheated pipes arranged in a triangular lattice with a relative step of S/d=1.185 is used as a model. The experiments were carried out on the sodium stand at constant energy release along the length of the fuel element simulators and at the Pe number changing in the 14-700 range. The data obtained show considerable increase of nonuniformities of the fuel element temperatures for characteristic zones of the fuel cassette assembly models of the fast reactor at deviations of the lattice geometric sizes from the nominal ones. For the central nonribbed element the temperature nonuniformity increases approximately 7.5 times and for the lateral element approximately 6 times when the elements touch each other along the whole length. The shift the central nonribbed element by the value of the gap between the fu.el elements leads to the decrease of heat transfer in comparison with heat transfer at the nominal geometry approximately 3-7 times in the 10-450 range for the Pe numbers. It is shown that the coolant temperature distribution along the assembly radius has a complex character (with a peak between the centre and the perifery) caused by redistribution of coolant consumptions due to fuel element lattice deformation

  18. Development of a fuzzy logic method to build objective functions in optimization problems: application to BWR fuel lattice design

    International Nuclear Information System (INIS)

    Martin-del-Campo, C.; Francois, J.L.; Barragan, A.M.; Palomera, M.A.

    2005-01-01

    In this paper we develop a methodology based on the use of the Fuzzy Logic technique to build multi-objective functions to be used in optimization processes applied to in-core nuclear fuel management. As an example, we selected the problem of determining optimal radial fuel enrichment and gadolinia distributions in a typical 'Boiling Water Reactor (BWR)' fuel lattice. The methodology is based on the use of the mathematical capability of Fuzzy Logic to model nonlinear functions of arbitrary complexity. The utility of Fuzzy Logic is to map an input space into an output space, and the primary mechanism for doing this is a list of if-then statements called rules. The rules refer to variables and adjectives that describe those variables and, the Fuzzy Logic technique interprets the values in the input vectors and, based on the set of rules assigns values to the output vector. The methodology was developed for the radial optimization of a BWR lattice where the optimization algorithm employed is Tabu Search. The global objective is to find the optimal distribution of enrichments and burnable poison concentrations in a 10*10 BWR lattice. In order to do that, a fuzzy control inference system was developed using the Fuzzy Logic Toolbox of Matlab and it has been linked to the Tabu Search optimization process. Results show that Tabu Search combined with Fuzzy Logic performs very well, obtaining lattices with optimal fuel utilization. (authors)

  19. Analysis of neutron parameters in light water moderated lattices of ThO2 and UO2 fuel rods

    International Nuclear Information System (INIS)

    Onusic Junior, J.; Oosterkamp, W.J.

    1977-01-01

    A large number of light water moderated lattices of UO 2 and ThO 2 fuel rods were analyzed with the code HAMMER. The purpose of the study was to compare experimental results with computer calculated values. The model employed is described and some modification were introduced in the resonance parameters of Th-232 to increase the agreement with the experimental value [pt

  20. Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

    International Nuclear Information System (INIS)

    Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

    2016-01-01

    One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

  1. 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)

  2. Analytic Methods for Benchmarking Hydrogen and Fuel Cell Technologies; NREL (National Renewable Energy Laboratory)

    Energy Technology Data Exchange (ETDEWEB)

    Melaina, Marc; Saur, Genevieve; Ramsden, Todd; Eichman, Joshua

    2015-05-28

    This presentation summarizes NREL's hydrogen and fuel cell analysis work in three areas: resource potential, greenhouse gas emissions and cost of delivered energy, and influence of auxiliary revenue streams. NREL's hydrogen and fuel cell analysis projects focus on low-­carbon and economic transportation and stationary fuel cell applications. Analysis tools developed by the lab provide insight into the degree to which bridging markets can strengthen the business case for fuel cell applications.

  3. Measurement and CFD calculation of spacer loss coefficient for a tight-lattice fuel bundle

    International Nuclear Information System (INIS)

    In, Wang Kee; Shin, Chang Hwan; Kwack, Young Kyun; Lee, Chi Young

    2015-01-01

    Highlights: • Experiment and CFD analysis evaluated the pressure drop in a spacer grid. • The measurement and CFD errors for the spacer loss coefficient were estimated. • The spacer loss coefficient for the dual-cooled annular fuel bundle was determined. • The CFD prediction agrees with the measured spacer loss coefficient within 8%. - Abstract: An experiment and computational fluid dynamics (CFD) analysis were performed to evaluate the pressure drop in a spacer grid for a dual-cooled annular fuel (DCAF) bundle. The DCAF bundle for the Korean optimum power reactor (OPR1000) is a 12 × 12 tight-lattice rod array with a pitch-to-diameter ratio of 1.08 owing to a larger outer diameter of the annular fuel rod. An experiment was conducted to measure the pressure drop in spacer grid for the DCAF bundle. The test bundle is a full-size 12 × 12 rod bundle with 11 spacer grid. The test condition covers a Reynolds number range of 2 × 10 4 –2 × 10 5 by changing the temperature and flow rate of water. A CFD analysis was also performed to predict the pressure drop through a spacer grid using the full-size and partial bundle models. The pressure drop and loss coefficient of a spacer grid were predicted and compared with the experimental results. The CFD predictions of spacer pressure drop and loss coefficient agree with the measured values within 8%. The spacer loss coefficient for the DCAF bundle is estimated to be approximately 1.50 at a nominal operating condition of OPR1000, i.e., Re = 4 × 10 5

  4. Benchmark testing of CENDL-2 for U-fuel thermal reactors

    International Nuclear Information System (INIS)

    Zhang Baocheng; Liu Guisheng; Liu Ping

    1995-01-01

    Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

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

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

    2002-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-01

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

  7. Positron annihilation spectroscopy study of lattice defects in non-irradiated doped and un-doped fuels

    Directory of Open Access Journals (Sweden)

    Chollet Mélanie

    2017-01-01

    Full Text Available Fission gas behavior within the fuel structure plays a major role for the safety of nuclear fuels during operation in the nuclear power plant. Fission gas distribution and retention is determined by both, micro- and lattice-structure of the fuel matrix. The ADOPT (Advanced Doped Pellet Technology fuel, containing chromium and aluminum additives, shows larger grain sizes than standard (undoped UO2 fuel, enhancing the fission gas retention properties of the matrix. However, the additions of such trivalent cations shall also induce defects in the lattice. In this study, we investigated the microstructure of such doped fuels as well as a reference standard UO2 by positron annihilation spectroscopy (PAS. Although this technique is particularly sensitive to lattice point defects in materials, a wider application in the UO2 research is still missing. The PAS-lifetime components were measured in the hotlab facility of PSI using a 22Na source sandwiched between two 500-μm-thin sample discs. The values of lifetime at the center and the rim of both samples, examined to check at the radial homogeneity of the pellets, are not significantly different. The mean lifetimes were found to be longer in the ADOPT material, 220 ps, than in standard UO2, 190 ps, which indicates a larger presence of additional defects, presumably generated by the dopants. While two-component decomposition (bulk + one defect component could be performed for the standard material, only one lifetime component was found in the doped material. The absence of the bulk component in the ADOPT sample refers to a saturated positron trapping (i.e., all positrons are trapped at defects. In order to associate a type of lattice defect to each PAS component, interpretation of the PAS experimental observations was conducted with respect to existing experimental and modeling studies. This work has shown the efficiency of PAS to detect lattice point defects in UO2 produced by Cr and Al oxides

  8. Validation of new evaluations for the main fuel nuclides using the ICSBEP handbook benchmarks

    International Nuclear Information System (INIS)

    Koscheev, V.; Manturov, G.; Rozhikhin, Y.; Tsibulya, A.

    2008-01-01

    The newest evaluations, adopted for Endf/B-VII.0, JEFF-3.1, JENDL-3.3 and the Russian library RUSFOND nuclear data files, for the most important fissile isotopes 235 U, 238 U, and 239 Pu are compared between each other and tested through a set of integral experiments, among them removal cross section under fission threshold of 238 U, critical infinite media Scherzo-556, and ICSBEP Handbook criticality safety benchmarks. Globally our benchmarking shows that these evaluations are in many cases very close. However, essential differences are observed through the analysis of critical systems with big enough content of 238 U. Large diversity still exists in inelastic scattering cross sections. We have to note that the divergence in the 238 U capture cross-section that existed in previous evaluations, has practically disappeared

  9. Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2

    International Nuclear Information System (INIS)

    Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa; Bando, Masaru.

    1993-10-01

    An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author)

  10. A nodal method of calculating power distributions for LWR-type reactors with square fuel lattices

    International Nuclear Information System (INIS)

    Hoeglund, Randolph.

    1980-06-01

    A nodal model is developed for calculating the power distribution in the core of a light water reactor with a square fuel lattice. The reactor core is divided into a number of more or less cubic nodes and a nodal coupling equation, which gives the thermal power density in one node as a function of the power densities in the neighbour nodes, is derived from the neutron diffusion equations for two energy groups. The three-dimensional power distribution can be computed iteratively using this coupling equation, for example following the point Jacobi, the Gauss-Seidel or the point successive overrelaxation scheme. The method has been included as the neutronic model in a reactor core simulation computer code BOREAS, where it is combined with a thermal-hydraulic model in order to make a simultaneous computation of the interdependent power and void distributions in a boiling water reactor possible. Also described in this report are a method for temporary one-dimensional iteration developed in order to accelerate the iterative solution of the problem and the Haling principle which is widely used in the planning of reloading operations for BWR reactors. (author)

  11. Two-dimensional benchmark calculations for PNL-30 through PNL-35

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1997-01-01

    Interest in critical experiments with lattices of mixed-oxide (MOX) fuel pins has been revived by the possibility that light water reactors will be used for disposition of weapons-grade plutonium. A series of six experiments with MOX lattices, designated PNL-30 through PNL-35, was performed at Pacific Northwest Laboratories in 1975 and 1976, and a set of benchmark specifications for these experiments subsequently was adopted by the Cross Section Evaluation Working Group (CSEWG). Although there appear to be some problems with these experiments, they remain the only CSEWG benchmarks for MOX lattices. The number of fuel pins in these experiments is relatively low, corresponding to fewer than 4 typical pressurized-water-reactor fuel assemblies. Accordingly, they are more appropriate as benchmarks for lattice-physics codes than for reactor-core simulator codes. Unfortunately, the CSEWG specifications retain the full three-dimensional (3D) detail of the experiments, while lattice-physics codes almost universally are limited to two dimensions (2D). This paper proposes an extension of the benchmark specifications to include a 2D model, and it justifies that extension by comparing results from the MCNP Monte Carlo code for the 2D and 3D specifications

  12. Fuel Cell Development for NASA's Human Exploration Program: Benchmarking with "The Hydrogen Economy"

    Science.gov (United States)

    Scott, John H.

    2007-01-01

    The theoretically high efficiency and low temperature operation of hydrogen-oxygen fuel cells has motivated them to be the subject of much study since their invention in the 19th Century, but their relatively high life cycle costs kept them as a "solution in search of a problem" for many years. The first problem for which fuel cells presented a truly cost effective solution was that of providing a power source for NASA's human spaceflight vehicles in the 1960 s. NASA thus invested, and continues to invest, in the development of fuel cell power plants for this application. This development program continues to place its highest priorities on requirements for minimum system mass and maximum durability and reliability. These priorities drive fuel cell power plant design decisions at all levels, even that of catalyst support. However, since the mid-1990's, prospective environmental regulations have driven increased governmental and industrial interest in "green power" and the "Hydrogen Economy." This has in turn stimulated greatly increased investment in fuel cell development for a variety of commercial applications. This investment is bringing about notable advances in fuel cell technology, but, as these development efforts place their highest priority on requirements for minimum life cycle cost and field safety, these advances are yielding design solutions quite different at almost every level from those needed for spacecraft applications. This environment thus presents both opportunities and challenges for NASA's Human Exploration Program

  13. Benchmark calculations of power distribution within fuel assemblies. Phase 2: comparison of data reduction and power reconstruction methods in production codes

    International Nuclear Information System (INIS)

    2000-01-01

    Systems loaded with plutonium in the form of mixed-oxide (MOX) fuel show somewhat different neutronic characteristics compared with those using conventional uranium fuels. In order to maintain adequate safety standards, it is essential to accurately predict the characteristics of MOX-fuelled systems and to further validate both the nuclear data and the computation methods used. A computation benchmark on power distribution within fuel assemblies to compare different techniques used in production codes for fine flux prediction in systems partially loaded with MOX fuel was carried out at an international level. It addressed first the numerical schemes for pin power reconstruction, then investigated the global performance including cross-section data reduction methods. This report provides the detailed results of this second phase of the benchmark. The analysis of the results revealed that basic data still need to be improved, primarily for higher plutonium isotopes and minor actinides. (author)

  14. Montecarlo calculation for a benchmark on interactive effects of Gadolinium poisoned pins in BWRs

    International Nuclear Information System (INIS)

    Borgia, M.G.; Casali, F.; Cepraga, D.

    1985-01-01

    K infinite and burn-up calculations have been done in the frame of a benchmark organized by Physic Reactor Committee of NEA. The calculations, performed by the Montecarlo code KIM, concerned BWR lattices having UO*L2 fuel rodlets with and without gadolinium oxide

  15. WIMSD4 calculations of the Westinghouse 'EDASA' lattices with plutonium dioxide fuel

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1977-03-01

    A series of Westinghouse critical PuO 2 /UO 2 pin-cell assemblies is analysed using the lattice code WIMSD4. The results are presented in terms of computed k-effective values, with comment on the choice of method for calculating high leakage systems and on the adequacy of WIMSD4 for evaluating plutonium enriched lattices. (author)

  16. Solutions obtained to international heat transfer benchmarking problems for nuclear fuel casks using Q/TRAN

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-02-01

    In 1985 Sandia National Laboratories participated in the Nuclear Energy Agency Committee on Reactor Physics (NEACRP) Specialists' Meeting on Heat Transfer Assessment of Transportation Packages. The objective of the meeting was to establish a set of model problems for use in comparing the performance of thermal analysis computer codes that may be used in the design of nuclear fuel shipping casks. The selected problems are to be used to compare code results for the thermal phenomena of conduction, convection, and radiation in cask-like problems. Two model problems were used in this study. The first problem required the determination of the steady-state temperatures of a 16 x 16 array of heated and unheated pins (representing fuel and control rod positions) of a simulated PWR fuel assembly. The second problem required the determination of transient temperatures of a finned surface (representing the external surface of a cask) subjected to an internal heat flux and to an external engulfing fire. Solutions to the problems were obtained with the code ''Q/TRAN.'' Solutions and descriptions of the necessary modeling techniques are given in this report

  17. A pore-scale model for the cathode electrode of a proton exchange membrane fuel cell by lattice Boltzmann method

    Energy Technology Data Exchange (ETDEWEB)

    Molaeimanesh, Gholam Reza; Akbari, Mohammad Hadi [Shiraz University, Shiraz (Iran, Islamic Republic of)

    2015-03-15

    A pore-scale model based on the lattice Boltzmann method (LBM) is proposed for the cathode electrode of a PEM fuel cell with heterogeneous and anisotropic porous gas diffusion layer (GDL) and interdigitated flow field. An active approach is implemented to model multi-component transport in GDL, which leads to enhanced accuracy, especially at higher activation over-potentials. The core of the paper is the implementation of an electrochemical reaction with an active approach in a multi-component lattice Boltzmann model for the first time. After model validation, the capability of the presented model is demonstrated through a parametric study. Effects of activation over-potential, pressure differential between inlet and outlet gas channels, land width to channel width ratio, and channel width are investigated. The results show the significant influence of GDL microstructure on the oxygen distribution and current density profile.

  18. Evaluation of AMPX-KENO benchmark calculations for high-density spent fuel storage racks

    International Nuclear Information System (INIS)

    Turner, S.E.; Gurley, M.K.

    1981-01-01

    The AMPX-KENO computer code package is commonly used to evaluate criticality in high-density spent fuel storage rack designs. Consequently, it is important to know the reliability that can be placed on such calculations and whether or not the results are conservative. This paper evaluates a series of AMPX-KENO calculations which have been made on selected critical experiments. The results are compared with similar analyses reported in the literature by the Oak Ridge National Laboratory and BandW. 8 refs

  19. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Primm III, RT

    2002-05-29

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.

  20. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    International Nuclear Information System (INIS)

    Primm III, RT

    2002-01-01

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors

  1. A simplified 2D HTTR benchmark problem

    International Nuclear Information System (INIS)

    Zhang, Z.; Rahnema, F.; Pounders, J. M.; Zhang, D.; Ougouag, A.

    2009-01-01

    To access the accuracy of diffusion or transport methods for reactor calculations, it is desirable to create heterogeneous benchmark problems that are typical of relevant whole core configurations. In this paper we have created a numerical benchmark problem in 2D configuration typical of a high temperature gas cooled prismatic core. This problem was derived from the HTTR start-up experiment. For code-to-code verification, complex details of geometry and material specification of the physical experiments are not necessary. To this end, the benchmark problem presented here is derived by simplifications that remove the unnecessary details while retaining the heterogeneity and major physics properties from the neutronics viewpoint. Also included here is a six-group material (macroscopic) cross section library for the benchmark problem. This library was generated using the lattice depletion code HELIOS. Using this library, benchmark quality Monte Carlo solutions are provided for three different configurations (all-rods-in, partially-controlled and all-rods-out). The reference solutions include the core eigenvalue, block (assembly) averaged fuel pin fission density distributions, and absorption rate in absorbers (burnable poison and control rods). (authors)

  2. Benchmarking ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Marck, Steven C. van der

    2006-01-01

    The new major release VII.0 of the ENDF/B nuclear data library has been tested extensively using benchmark calculations. These were based upon MCNP-4C3 continuous-energy Monte Carlo neutronics simulations, together with nuclear data processed using the code NJOY. Three types of benchmarks were used, viz., criticality safety benchmarks (fusion) shielding benchmarks, and reference systems for which the effective delayed neutron fraction is reported. For criticality safety, more than 700 benchmarks from the International Handbook of Criticality Safety Benchmark Experiments were used. Benchmarks from all categories were used, ranging from low-enriched uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), to mixed uranium-plutonium, metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments at Lawrence Livermore National Laboratory (for 6 Li, 7 Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D 2 O, H 2 O, concrete, polyethylene and teflon). For testing delayed neutron data more than thirty measurements in widely varying systems were used. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, and two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra. In criticality safety, many benchmarks were chosen from the category with a thermal spectrum, low-enriched uranium, compound fuel (LEU-COMP-THERM), because this is typical of most current-day reactors, and because these benchmarks were previously underpredicted by as much as 0.5% by most nuclear data libraries (such as ENDF/B-VI.8, JEFF-3.0). The calculated results presented here show that this underprediction is no longer there for ENDF/B-VII.0. The average over 257

  3. Nanotechnology convergence and modeling paradigm of sustainable energy system using polymer electrolyte membrane fuel cell as a benchmark example

    International Nuclear Information System (INIS)

    Chung, Pil Seung; So, Dae Sup; Biegler, Lorenz T.; Jhon, Myung S.

    2012-01-01

    Developments in nanotechnology have led to innovative progress and converging technologies in engineering and science. These demand novel methodologies that enable efficient communications from the nanoscale all the way to decision-making criteria for actual production systems. In this paper, we discuss the convergence of nanotechnology and novel multi-scale modeling paradigms by using the fuel cell system as a benchmark example. This approach includes complex multi-phenomena at different time and length scales along with the introduction of an optimization framework for application-driven nanotechnology research trends. The modeling paradigm introduced here covers the novel holistic integration from atomistic/molecular phenomena to meso/continuum scales. System optimization is also discussed with respect to the reduced order parameters for a coarse-graining procedure in multi-scale model integration as well as system design. The development of a hierarchical multi-scale paradigm consolidates the theoretical analysis and enables large-scale decision-making of process level design, based on first-principles, and therefore promotes the convergence of nanotechnology to sustainable energy technologies.

  4. Functionalized single-walled carbon nanotube-based fuel cell benchmarked against US DOE 2017 technical targets.

    Science.gov (United States)

    Jha, Neetu; Ramesh, Palanisamy; Bekyarova, Elena; Tian, Xiaojuan; Wang, Feihu; Itkis, Mikhail E; Haddon, Robert C

    2013-01-01

    Chemically modified single-walled carbon nanotubes (SWNTs) with varying degrees of functionalization were utilized for the fabrication of SWNT thin film catalyst support layers (CSLs) in polymer electrolyte membrane fuel cells (PEMFCs), which were suitable for benchmarking against the US DOE 2017 targets. Use of the optimum level of SWNT -COOH functionality allowed the construction of a prototype SWNT-based PEMFC with total Pt loading of 0.06 mg(Pt)/cm²--well below the value of 0.125 mg(Pt)/cm² set as the US DOE 2017 technical target for total Pt group metals (PGM) loading. This prototype PEMFC also approaches the technical target for the total Pt content per kW of power (<0.125 g(PGM)/kW) at cell potential 0.65 V: a value of 0.15 g(Pt)/kW was achieved at 80°C/22 psig testing conditions, which was further reduced to 0.12 g(Pt)/kW at 35 psig back pressure.

  5. Nanotechnology convergence and modeling paradigm of sustainable energy system using polymer electrolyte membrane fuel cell as a benchmark example

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Pil Seung; So, Dae Sup; Biegler, Lorenz T.; Jhon, Myung S., E-mail: mj3a@andrew.cmu.edu [Carnegie Mellon University, Department of Chemical Engineering (United States)

    2012-08-15

    Developments in nanotechnology have led to innovative progress and converging technologies in engineering and science. These demand novel methodologies that enable efficient communications from the nanoscale all the way to decision-making criteria for actual production systems. In this paper, we discuss the convergence of nanotechnology and novel multi-scale modeling paradigms by using the fuel cell system as a benchmark example. This approach includes complex multi-phenomena at different time and length scales along with the introduction of an optimization framework for application-driven nanotechnology research trends. The modeling paradigm introduced here covers the novel holistic integration from atomistic/molecular phenomena to meso/continuum scales. System optimization is also discussed with respect to the reduced order parameters for a coarse-graining procedure in multi-scale model integration as well as system design. The development of a hierarchical multi-scale paradigm consolidates the theoretical analysis and enables large-scale decision-making of process level design, based on first-principles, and therefore promotes the convergence of nanotechnology to sustainable energy technologies.

  6. Comparison of heuristic optimization techniques for the enrichment and gadolinia distribution in BWR fuel lattices and decision analysis

    International Nuclear Information System (INIS)

    Castillo, Alejandro; Martín-del-Campo, Cecilia; Montes-Tadeo, José-Luis; François, Juan-Luis; Ortiz-Servin, Juan-José; Perusquía-del-Cueto, Raúl

    2014-01-01

    Highlights: • Different metaheuristic optimization techniques were compared. • The optimal enrichment and gadolinia distribution in a BWR fuel lattice was studied. • A decision making tool based on the Position Vector of Minimum Regret was applied. • Similar results were found for the different optimization techniques. - Abstract: In the present study a comparison of the performance of five heuristic techniques for optimization of combinatorial problems is shown. The techniques are: Ant Colony System, Artificial Neural Networks, Genetic Algorithms, Greedy Search and a hybrid of Path Relinking and Scatter Search. They were applied to obtain an “optimal” enrichment and gadolinia distribution in a fuel lattice of a boiling water reactor. All techniques used the same objective function for qualifying the different distributions created during the optimization process as well as the same initial conditions and restrictions. The parameters included in the objective function are the k-infinite multiplication factor, the maximum local power peaking factor, the average enrichment and the average gadolinia concentration of the lattice. The CASMO-4 code was used to obtain the neutronic parameters. The criteria for qualifying the optimization techniques include also the evaluation of the best lattice with burnup and the number of evaluations of the objective function needed to obtain the best solution. In conclusion all techniques obtain similar results, but there are methods that found better solutions faster than others. A decision analysis tool based on the Position Vector of Minimum Regret was applied to aggregate the criteria in order to rank the solutions according to three functions: neutronic grade at 0 burnup, neutronic grade with burnup and global cost which aggregates the computing time in the decision. According to the results Greedy Search found the best lattice in terms of the neutronic grade at 0 burnup and also with burnup. However, Greedy Search is

  7. Overview and benchmark analysis of fuel cell parameters estimation for energy management purposes

    Science.gov (United States)

    Kandidayeni, M.; Macias, A.; Amamou, A. A.; Boulon, L.; Kelouwani, S.; Chaoui, H.

    2018-03-01

    Proton exchange membrane fuel cells (PEMFCs) have become the center of attention for energy conversion in many areas such as automotive industry, where they confront a high dynamic behavior resulting in their characteristics variation. In order to ensure appropriate modeling of PEMFCs, accurate parameters estimation is in demand. However, parameter estimation of PEMFC models is highly challenging due to their multivariate, nonlinear, and complex essence. This paper comprehensively reviews PEMFC models parameters estimation methods with a specific view to online identification algorithms, which are considered as the basis of global energy management strategy design, to estimate the linear and nonlinear parameters of a PEMFC model in real time. In this respect, different PEMFC models with different categories and purposes are discussed first. Subsequently, a thorough investigation of PEMFC parameter estimation methods in the literature is conducted in terms of applicability. Three potential algorithms for online applications, Recursive Least Square (RLS), Kalman filter, and extended Kalman filter (EKF), which has escaped the attention in previous works, have been then utilized to identify the parameters of two well-known semi-empirical models in the literature, Squadrito et al. and Amphlett et al. Ultimately, the achieved results and future challenges are discussed.

  8. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Perret, G. [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland)

    2012-07-01

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  9. Neutron multipilication factors as a function of temperature: a comparison of calculated and measured values for lattices using 233UO2-ThO2 fuel in graphite

    International Nuclear Information System (INIS)

    Newman, D.F.; Gore, B.F.

    1978-01-01

    Neutron multiplication factors calculated as a function of temperature for three graphite-moderated 233 UO 2 -ThO 2 -fueled lattices are correlated with the values measured for these lattices in the high-temperature lattice test reactor (HTLTR). The correlation analysis is accomplished by fitting calculated values of k/sub infinity/(T) to the measured values using two least-squares-fitted correlation coefficients: (a) a normalization factor and (b) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross-section data. Use of an alternate cross-section data set for thorium, which has a smaller resonance integral than ENDF/B-IV data, improved the agreement between calculated and measured temperature coefficients of reactivity for the three experimental lattices. The results of the correlations are used to estimate the bias in the temperature coefficient of reactivity calculated for a lattice typical of fresh 233 U recycle fuel for a high-temperature gas-cooled reactor (HTGR). This extrapolation to a lattice having a heavier fissile loading than the experimental lattices is accomplished using a sensitivity analysis of the estimated bias to alternate thorium cross-section data used in calculations of k/sub infinity/(T). The envelope of uncertainty expected to contain the actual values for the temperature coefficient of the reactivity for the 233 U-fueled HTGR lattice studied remains negative at 1600 K (1327 0 C). Although a broader base of experimental data with improved accuracy is always desirable, the existing data base provided by the HTLTR experiments is judged to be adequate for the verification of neutronic calculations for the HTGR containing 233 U fuel at its current state of development

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

    International Nuclear Information System (INIS)

    Dalle, Hugo M.

    2009-01-01

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

  11. Participation in benchmark MATIS-H of NEA/OCDE: uses CFD codes applied to nuclear safety. Study of the spacer grids in the fuel elements

    International Nuclear Information System (INIS)

    Pena-Monferrer, C.; Chiva, S.; Munoz-cobo, J. L.; Vela, E.

    2012-01-01

    This paper develops participation in benchmark MATIS-H, promoted by the NEA / OECD-KAERI, involving the study of turbulent flow in a rod beam with spacers in an experimental installation. Its aim is the analysis of hydraulic behavior of turbulent flow in the subchannels of the fuel elements, essential for the improvement of safety margins in normal and transient operations and to maximize the use of nuclear energy through an optimal design of grids.

  12. Quantification of TRISO fuel heterogeneity effects in HTGR lattice physics calculations

    International Nuclear Information System (INIS)

    Perfetti, C. M.; Anghaie, S.; Dugan, E.; Marcille, T.

    2010-01-01

    A large number of LEU-MHR fuel compact models were generated with randomly distributed TRISO particle fuel and were simulated using MCNP5, and it was determined how several neutronic parameters, including k-infinite, the thermal and fast diffusion coefficients, and the four factors, varied across the randomly-generated cases. A sensitivity study was also performed to determine how the four factors depend on the definition of the thermal energy group. Values of k-infinite for the cases had a sample standard deviation of 248 pcm and were found to follow an approximately normal distribution about the mean value of k-infinite. Although all of the four factors were found to have similar sample standard deviations, the resonance escape probability was found to be the most variable parameter with a sample relative standard deviation between 0.07% and 0.08%. HTGR fuel compact homogenization methods typically examine only one reference fuel compact that contains a uniform distribution of TRISO particles, but in reality the TRISO particles are randomly distributed throughout the fuel compact. Thus, the neutronic parameters for actual fuel compacts differ randomly from those in the reference model. To license next-generation High-Temperature Gas Reactors engineers must quantify all uncertainties of the design and this random variation in neutron parameters is a previously unmeasured quantity; this study measures this uncertainty by examining the variation in k-infinite for HTGR fuel compact models with randomly distributed TRISO fuel. (authors)

  13. Experimental determination of the Koo fuel temperature coefficient for an HTGR lattice

    Energy Technology Data Exchange (ETDEWEB)

    Agostini, P.; Benedetti, F.; Brighenti, G.; Chiodi, P. L.; Dell' Oro, P.; Giuliani, C.; Tassan, S.

    1974-10-15

    This paper describes temperature-dependent k-infinity measurements conducted using an assembly of loose HTGR coated particles in the BR-2 reactor by means of null reactivity oscillating method comparing the effect of poisoned and unpoisoned lattices like tests performed in the Physical Constants Test Reactor (PCTR) at Hanford. The RB-2 reactor was the property of the Italian firm AGIP NUCLEARE and operated at the Montecuccolino Center in Bologna.

  14. Extending the basic function of lattice oxygen in lepidocrocite titanate - The conversion of intercalated fatty acid to liquid hydrocarbon fuels

    Science.gov (United States)

    Maluangnont, Tosapol; Arsa, Pornanan; Sooknoi, Tawan

    2017-12-01

    We report herein the basicity of the external and internal lattice oxygen (OL) in lepidocrocite titanates with respect to CO2 and palmitic acid, respectively. Several compositions have been tested with different types of the metal M aliovalently (co)substituted for Ti, K0.8[MyTi2-y]O4 (M = Li, Mg, Fe, Co, Ni, Cu, Zn, Cu/Ni and Cu/Zn). The low CO2 desorption peak temperature (70-100 °C) suggests that the external OL sites are weakly basic similar to TiO2. However, the internal OL sites are sufficiently basic to deprotonate palmitic acid, forming the intercalated potassium palmitate at the interlayer spaces. The latter serves as a two-dimensional (2D) molecular reactor for the production of liquid hydrocarbon fuels via deoxygenation under atmospheric N2. A relationship has been observed between the yield of the liquid products vs the partial charge of the lattice oxygen (δO). Since the deoxygenation pathway is highly dependent on the metal substitution, the redox-active sites might also play some roles. The co-substituted K0.8[Cu0.2Ni0.2]Ti1.6O4 produced 68.0% yield of the liquid products, with 51% saturated and 15% unsaturated C15 hydrocarbons at 350 °C.

  15. CIEMAT’s contribution to the phase II of the OECD-NEA RIA benchmark on thermo-mechanical fuel codes performance

    Energy Technology Data Exchange (ETDEWEB)

    Sagrado, I.C.; Vallejo, I.; Herranz, L.E.

    2015-07-01

    As a part of the international efforts devoted to validate and/or update the current fuel safety criteria, the OECD-NEA has launched a second phase of the RIA benchmark on thermomechanical fuel codes performance. CIEMAT contributes simulating the ten scenarios proposed with FRAPTRAN and SCANAIR. Both codes lead to similar predictions during the heating-up; however, during the cooling-down significant deviations may appear. They are mainly caused by the estimations of gap closure and re-opening and the clad to water heat exchange approaches. The uncertainty analysis performed for the SCANAIR estimations leads to uncertainty ranges below 15% and 28% for maximum temperatures and deformations, respectively. The corresponding sensitivity analysis shows that, in addition to the injected energy, special attention should be paid to fuel thermal expansion and clad yield stress models. (Author)

  16. Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

    2016-01-01

    Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

  17. Uncertainty and sensitivity analysis in reactivity-initiated accident fuel modeling: synthesis of organisation for economic co-operation and development (OECD/nuclear energy agency (NEA benchmark on reactivity-initiated accident codes phase-II

    Directory of Open Access Journals (Sweden)

    Olivier Marchand

    2018-03-01

    Full Text Available In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010–2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II was thus launched early in 2014, which has been organized in two complementary activities: (1 comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2 assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made. Keywords: RIA, Codes Benchmarking, Fuel Modelling, OECD

  18. Monte Carlo benchmark calculations for 400MWTH PBMR core

    International Nuclear Information System (INIS)

    Kim, H. C.; Kim, J. K.; Kim, S. Y.; Noh, J. M.

    2007-01-01

    A large interest in high-temperature gas-cooled reactors (HTGR) has been initiated in connection with hydrogen production in recent years. In this study, as a part of work for establishing Monte Carlo computation system for HTGR core analysis, some benchmark calculations for pebble-type HTGR were carried out using MCNP5 code. The core of the 400MW t h Pebble-bed Modular Reactor (PBMR) was selected as a benchmark model. Recently, the IAEA CRP5 neutronics and thermal-hydraulics benchmark problem was proposed for the testing of existing methods for HTGRs to analyze the neutronics and thermal-hydraulic behavior for the design and safety evaluations of the PBMR. This study deals with the neutronic benchmark problems, for fresh fuel and cold conditions (Case F-1), and first core loading with given number densities (Case F-2), proposed for PBMR. After the detailed MCNP modeling of the whole facility, benchmark calculations were performed. Spherical fuel region of a fuel pebble is divided into cubic lattice element in order to model a fuel pebble which contains, on average, 15000 CFPs (Coated Fuel Particles). Each element contains one CFP at its center. In this study, the side length of each cubic lattice element to have the same amount of fuel was calculated to be 0.1635 cm. The remaining volume of each lattice element was filled with graphite. All of different 5 concentric shells of CFP were modeled. The PBMR annular core consists of approximately 452000 pebbles in the benchmark problems. In Case F-1 where the core was filled with only fresh fuel pebble, a BCC(body-centered-cubic) lattice model was employed in order to achieve the random packing core with the packing fraction of 0.61. The BCC lattice was also employed with the size of the moderator pebble increased in a manner that reproduces the specified F/M ratio of 1:2 while preserving the packing fraction of 0.61 in Case F-2. The calculations were pursued with ENDF/B-VI cross-section library and used sab2002 S(α,

  19. Investigation of GDL compression effects on the performance of a PEM fuel cell cathode by lattice Boltzmann method

    Science.gov (United States)

    Molaeimanesh, G. R.; Nazemian, M.

    2017-08-01

    Proton exchange membrane (PEM) fuel cells with a great potential for application in vehicle propulsion systems will have a promising future. However, to overcome the exiting challenges against their wider commercialization further fundamental research is inevitable. The effects of gas diffusion layer (GDL) compression on the performance of a PEM fuel cell is not well-recognized; especially, via pore-scale simulation technique capturing the fibrous microstructure of the GDL. In the current investigation, a stochastic microstructure reconstruction method is proposed which can capture GDL microstructure changes by compression. Afterwards, lattice Boltzmann pore-scale simulation technique is adopted to simulate the reactive gas flow through 10 different cathode electrodes with dissimilar carbon paper GDLs produced from five different compression levels and two different carbon fiber diameters. The distributions of oxygen mole fraction, water vapor mole fraction and current density for the simulated cases are presented and analyzed. The results of simulations demonstrate that when the fiber diameter is 9 μm adding compression leads to lower average current density while when the fiber diameter is 7 μm the compression effect is not monotonic.

  20. Analysis of a computational benchmark for a high-temperature reactor using SCALE

    International Nuclear Information System (INIS)

    Goluoglu, S.

    2006-01-01

    Several proposed advanced reactor concepts require methods to address effects of double heterogeneity. In doubly heterogeneous systems, heterogeneous fuel particles in a moderator matrix form the fuel region of the fuel element and thus constitute the first level of heterogeneity. Fuel elements themselves are also heterogeneous with fuel and moderator or reflector regions, forming the second level of heterogeneity. The fuel elements may also form regular or irregular lattices. A five-phase computational benchmark for a high-temperature reactor (HTR) fuelled with uranium or reactor-grade plutonium has been defined by the Organization for Economic Cooperation and Development, Nuclear Energy Agency (OECD NEA), Nuclear Science Committee, Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles. This paper summarizes the analysis results using the latest SCALE code system (to be released in CY 2006 as SCALE 5.1). (authors)

  1. Resonance shielding in thermal reactor lattices

    International Nuclear Information System (INIS)

    Rothenstein, W.; Taviv, E.; Aminpour, M.

    1982-01-01

    The theoretical foundations of a new methodology for the accurate treatment of resonance absorption in thermal reactor lattice analysis are presented. This methodology is based on the solution of the point-energy transport equation in its integral or integro-differential form for a heterogeneous lattice using detailed resonance cross-section profiles. The methodology is applied to LWR benchmark analysis, with emphasis on temperature dependence of resonance absorption during fuel depletion, spatial and mutual self-shielding, integral parameter analysis and treatment of cluster geometry. The capabilities of the OZMA code, which implements the new methodology are discussed. These capabilities provide a means against which simpler and more rapid resonance absorption algorithms can be checked. (author)

  2. Development of quantitative analytical procedures on two-phase flow in tight-lattice fuel bundles for reduced-moderation light-water reactors

    International Nuclear Information System (INIS)

    Ohnuki, A.; Kureta, M.; Takae, K.; Tamai, H.; Akimoto, H.; Yoshida, H.

    2004-01-01

    The research project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR is a light water reactor for which a higher conversion ratio more than one can be expected. In order to attain this higher conversion ratio, triangular tight-lattice fuel bundles whose gap spacing between each fuel rod is around 1 mm are required. As for the thermal design of the RMWR core, conventional analytical methods are no good because the conventional composition equations can not predict the RMWR core with high accuracy. Then, development of new quantitative analytical procedures was carried out. Those analytical procedures are constructed by model experiments and advanced two-phase flow analysis codes. This paper describes the results of the model experiments and analytical results with the developed analysis codes. (authors)

  3. Benchmark of the CASMO-3G/MICROBURN-B codes for Commonwealth Edison boiling water reactors

    International Nuclear Information System (INIS)

    Wheeler, J.K.; Pallotta, A.S.

    1992-01-01

    The Commonwealth Edison Company has performed an extensive benchmark against measured data from three boiling water reactors using the Studsvik lattice physics code CASMO-3G and the Siemens Nuclear Power three-dimensional simulator code MICROBURN-B. The measured data of interest for this benchmark are the hot and cold reactivity, and the core power distributions as measured by the traversing incore probe system and gamma scan data for fuel pins and assemblies. A total of nineteen unit-cycles were evaluated. The database included fuel product lines manufactured by General Electric and Siemens Nuclear Power, wit assemblies containing 7 x 7 to 9 x 9 pin configurations, several water rod designs, various enrichments and gadolina loadings, and axially varying lattice designs throughout the enriched portion of the bundle. The results of the benchmark present evidence that the CASMO-3G/MICROBURN-B code package can adequately model the range of fuel and core types in the benchmark, and the codes are acceptable for performing neutronic analyses of Commonwealth Edison's boiling water reactors

  4. Library Benchmarking

    Directory of Open Access Journals (Sweden)

    Wiji Suwarno

    2017-02-01

    Full Text Available The term benchmarking has been encountered in the implementation of total quality (TQM or in Indonesian termed holistic quality management because benchmarking is a tool to look for ideas or learn from the library. Benchmarking is a processof measuring and comparing for continuous business process of systematic and continuous measurement, the process of measuring and comparing for continuous business process of an organization to get information that can help these organization improve their performance efforts.

  5. Interlaboratory computational comparisons of critical fast test reactor pin lattices

    International Nuclear Information System (INIS)

    Mincey, J.F.; Kerr, H.T.; Durst, B.M.

    1979-01-01

    An objective of the Consolidated Fuel Reprocessing Program's (CFRP) nuclear engineering group at Oak Ridge National Laboratory (ORNL) is to ensure that chemical equipment components designed for the reprocessing of spent LMFBR fuel (among other fuel types) are safe from a criticality standpoint. As existing data are inadequate for the general validation of computational models describing mixed plutonium--uranium oxide systems with isotopic compositions typical of LMFBR fuel, a program of critical experiments has been initiated at the Battelle Pacific Northwest Laboratories (PNL). The first series of benchmark experiments consisted of five square-pitched lattices of unirradiated Fast Test Reactor (FTR) fuel moderated and reflected by light water. Calculations of these five experiments have been conducted by both ORNL/CFRP and PNL personnel with the purpose of exploring how accurately various computational models will predict k/sub eff/ values for such neutronic systems and if differences between k/sub eff/ values obtained with these different models are significant

  6. Proposal of new 235U nuclear data to improve keff biases on 235U enrichment and temperature for low enriched uranium fueled lattices moderated by light water

    International Nuclear Information System (INIS)

    Wu, Haicheng; Okumura, Keisuke; Shibata, Keiichi

    2005-06-01

    The under prediction of k eff depending on 235 U enrichment in low enriched uranium fueled systems, which had been a long-standing puzzle especially for slightly enriched ones, was studied in this report. Benchmark testing was carried out with several evaluated nuclear data files, including the new uranium evaluations from preliminary ENDF/B-VII and CENDL-3.1. Another problem reviewed here was k eff underestimation vs. temperature increase, which was observed in the sightly enriched system with recent JENDL and ENDF/B uranium evaluations. Through the substitute analysis of nuclear data of 235 U and 238 U, we propose a new evaluation of 235 U data to solve both of the problems. The new evaluation was tested for various uranium fueled systems including low or highly enriched metal and solution benchmarks in the ICSBEP handbook. As a result, it was found that the combination of the new evaluation of 235 U and the 238 U data from the preliminary ENDF/B-VII gives quite good results for most of benchmark problems. (author)

  7. Interactive benchmarking

    DEFF Research Database (Denmark)

    Lawson, Lartey; Nielsen, Kurt

    2005-01-01

    We discuss individual learning by interactive benchmarking using stochastic frontier models. The interactions allow the user to tailor the performance evaluation to preferences and explore alternative improvement strategies by selecting and searching the different frontiers using directional...... in the suggested benchmarking tool. The study investigates how different characteristics on dairy farms influences the technical efficiency....

  8. RUNE benchmarks

    DEFF Research Database (Denmark)

    Peña, Alfredo

    This report contains the description of a number of benchmarks with the purpose of evaluating flow models for near-shore wind resource estimation. The benchmarks are designed based on the comprehensive database of observations that the RUNE coastal experiment established from onshore lidar...

  9. Benchmark selection

    DEFF Research Database (Denmark)

    Hougaard, Jens Leth; Tvede, Mich

    2002-01-01

    Within a production theoretic framework, this paper considers an axiomatic approach to benchmark selection. It is shown that two simple and weak axioms; efficiency and comprehensive monotonicity characterize a natural family of benchmarks which typically becomes unique. Further axioms are added...... in order to obtain a unique selection...

  10. Heavy nucleus resonant absorption calculation benchmarks

    International Nuclear Information System (INIS)

    Tellier, H.; Coste, H.; Raepsaet, C.; Van der Gucht, C.

    1993-01-01

    The calculation of the space and energy dependence of the heavy nucleus resonant absorption in a heterogeneous lattice is one of the hardest tasks in reactor physics. Because of the computer time and memory needed, it is impossible to represent finely the cross-section behavior in the resonance energy range for everyday computations. Consequently, reactor physicists use a simplified formalism, the self-shielding formalism. As no clean and detailed experimental results are available to validate the self-shielding calculations, Monte Carlo computations are used as a reference. These results, which were obtained with the TRIPOLI continuous-energy Monte Carlo code, constitute a set of numerical benchmarks than can be used to evaluate the accuracy of the techniques or formalisms that are included in any reactor physics codes. Examples of such evaluations, for the new assembly code APOLLO2 and the slowing-down code SECOL, are given for cases of 238 U and 232 Th fuel elements

  11. Depletion benchmarks calculation of random media using explicit modeling approach of RMC

    International Nuclear Information System (INIS)

    Liu, Shichang; She, Ding; Liang, Jin-gang; Wang, Kan

    2016-01-01

    Highlights: • Explicit modeling of RMC is applied to depletion benchmark for HTGR fuel element. • Explicit modeling can provide detailed burnup distribution and burnup heterogeneity. • The results would serve as a supplement for the HTGR fuel depletion benchmark. • The method of adjacent burnup regions combination is proposed for full-core problems. • The combination method can reduce memory footprint, keeping the computing accuracy. - Abstract: Monte Carlo method plays an important role in accurate simulation of random media, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. Three stochastic geometry modeling methods including Random Lattice Method, Chord Length Sampling and explicit modeling approach with mesh acceleration technique, have been implemented in RMC to simulate the particle transport in the dispersed fuels, in which the explicit modeling method is regarded as the best choice. In this paper, the explicit modeling method is applied to the depletion benchmark for HTGR fuel element, and the method of combination of adjacent burnup regions has been proposed and investigated. The results show that the explicit modeling can provide detailed burnup distribution of individual TRISO particles, and this work would serve as a supplement for the HTGR fuel depletion benchmark calculations. The combination of adjacent burnup regions can effectively reduce the memory footprint while keeping the computational accuracy.

  12. Foreign travel report: Visits to UK, Belgium, Germany, and France to benchmark European spent fuel and waste management technology

    International Nuclear Information System (INIS)

    Ermold, L.F.; Knecht, D.A.

    1993-08-01

    The ICPP WINCO Spent Fuel and Waste Management Development Program recently was funded by DOE-EM to develop new technologies for immobilizing ICPP spent fuels, sodium-bearing liquid waste, and calcine to a form suitable for disposal. European organizations are heavily involved, in some cases on an industrial scale in areas of waste management, including spent fuel disposal and HLW vitrification. The purpose of this trip was to acquire first-hand European efforts in handling of spent reactor fuel and nuclear waste management, including their processing and technical capabilities as well as their future planning. Even though some differences exist in European and U.S. DOE waste compositions and regulations, many aspects of the European technologies may be applicable to the U.S. efforts, and several areas offer potential for technical collaboration

  13. Foreign travel report: Visits to UK, Belgium, Germany, and France to benchmark European spent fuel and waste management technology

    Energy Technology Data Exchange (ETDEWEB)

    Ermold, L.F.; Knecht, D.A.

    1993-08-01

    The ICPP WINCO Spent Fuel and Waste Management Development Program recently was funded by DOE-EM to develop new technologies for immobilizing ICPP spent fuels, sodium-bearing liquid waste, and calcine to a form suitable for disposal. European organizations are heavily involved, in some cases on an industrial scale in areas of waste management, including spent fuel disposal and HLW vitrification. The purpose of this trip was to acquire first-hand European efforts in handling of spent reactor fuel and nuclear waste management, including their processing and technical capabilities as well as their future planning. Even though some differences exist in European and U.S. DOE waste compositions and regulations, many aspects of the European technologies may be applicable to the U.S. efforts, and several areas offer potential for technical collaboration.

  14. Description and exploitation of benchmarks involving {sup 149}Sm, a fission product taking part of the burn up credit in spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Anno, J.; Poullot, G. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Fouillaud, P.; Grivot, P. [CEA Centre d`Etudes de Valduc, 21 - Is-sur-Tille (France)

    1995-12-31

    Up to now, there was no benchmark to validate the Fission Products (FPs) cross sections in criticality safety calculations. The protection and nuclear safety institute (IPSN) has begun an experimental program on 6 FPs ({sup 103}Rh, {sup 133}Cs, {sup 143}Nd, {sup 149}Sm, {sup 152}Sm, and {sup 155}Gd daughter of {sup 155}Eu) giving alone a decrease of reactivity equal to half the whole FPs in spent fuels (except Xe and I). Here are presented the experiments with the {sup 149}Sm and the results obtained with the APOLLO I-MORET III calculations codes. 11 experiments are carried out in a zircaloy tank of 3.5 1 containing slightly nitric acid solutions of Samarium (96,9% in weight of {sup 149S}m) at 0.1048 -0.2148 - 0.6262 g/l concentrations. It was placed in the middle of arrays of UO{sub 2} rods (4.742 % U5 weight %) at square pitch of 13 mm. The underwater height of the rods is the critical parameter. In addition, 7 experiments were performed with the same apparatus with water and boron proving a good experimental representativeness and a good accuracy of the calculations. As the reactivity worth of the Sm tank is between 2000 and 6000 10{sup -5}, the benchmarks are well representative and the cumulative absorption ratios show that {sup 149}Sm is well qualified under 1 eV. (authors). 8 refs., 7 figs., 6 tabs.

  15. Description and exploitation of benchmarks involving 149Sm, a fission product taking part of the burn up credit in spent fuels

    International Nuclear Information System (INIS)

    Anno, J.; Poullot, G.

    1995-01-01

    Up to now, there was no benchmark to validate the Fission Products (FPs) cross sections in criticality safety calculations. The protection and nuclear safety institute (IPSN) has begun an experimental program on 6 FPs ( 103 Rh, 133 Cs, 143 Nd, 149 Sm, 152 Sm, and 155 Gd daughter of 155 Eu) giving alone a decrease of reactivity equal to half the whole FPs in spent fuels (except Xe and I). Here are presented the experiments with the 149 Sm and the results obtained with the APOLLO I-MORET III calculations codes. 11 experiments are carried out in a zircaloy tank of 3.5 1 containing slightly nitric acid solutions of Samarium (96,9% in weight of 149S m) at 0.1048 -0.2148 - 0.6262 g/l concentrations. It was placed in the middle of arrays of UO 2 rods (4.742 % U5 weight %) at square pitch of 13 mm. The underwater height of the rods is the critical parameter. In addition, 7 experiments were performed with the same apparatus with water and boron proving a good experimental representativeness and a good accuracy of the calculations. As the reactivity worth of the Sm tank is between 2000 and 6000 10 -5 , the benchmarks are well representative and the cumulative absorption ratios show that 149 Sm is well qualified under 1 eV. (authors). 8 refs., 7 figs., 6 tabs

  16. Analysis of a molten salt reactor benchmark

    International Nuclear Information System (INIS)

    Ghosh, Biplab; Bajpai, Anil; Degweker, S.B.

    2013-01-01

    This paper discusses results of our studies of an IAEA molten salt reactor (MSR) benchmark. The benchmark, proposed by Japan, involves burnup calculations of a single lattice cell of a MSR for burning plutonium and other minor actinides. We have analyzed this cell with in-house developed burnup codes BURNTRAN and McBURN. This paper also presents a comparison of the results of our codes and those obtained by the proposers of the benchmark. (author)

  17. WWER-1000 Burnup Credit Benchmark (CB5)

    International Nuclear Information System (INIS)

    Manolova, M.A.

    2002-01-01

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

  18. Burn-up credit criticality safety benchmark phase VII - UO2 fuel: study of spent fuel compositions for long-term disposal

    International Nuclear Information System (INIS)

    2012-01-01

    After spent nuclear fuel (SNF) is discharged from a nuclear reactor, fuel composition and reactivity continue to vary as a function of time due to the decay of unstable nuclides. Accurate predictions of the concentrations of long-lived radionuclides in SNF, which represent a significant potential hazard to human beings and to the environment over a very long period, are particularly necessary for radiological dose assessments. This report assesses the ability of existing computer codes and associated nuclear data to predict isotopic compositions and their corresponding neutron multiplication factor (k eff ) values for pressurised-water-reactor (PWR) UO 2 fuel at 50 GWd/MTU burn-up in a generic spent fuel cask configuration. Fuel decay compositions and k eff values have been calculated for 30 post-irradiation time steps out to one million years

  19. WLUP benchmarks

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2002-01-01

    The IAEA-WIMS Library Update Project (WLUP) is on the end stage. The final library will be released on 2002. It is a result of research and development made by more than ten investigators during 10 years. The organization of benchmarks for testing and choosing the best set of data has been coordinated by the author of this paper. It is presented the organization, name conventions, contents and documentation of WLUP benchmarks, and an updated list of the main parameters for all cases. First, the benchmarks objectives and types are given. Then, comparisons of results from different WIMSD libraries are included. Finally it is described the program QVALUE for analysis and plot of results. Some examples are given. The set of benchmarks implemented on this work is a fundamental tool for testing new multigroup libraries. (author)

  20. Solution of the two dimensional diffusion and transport equations in a rectangular lattice with an elliptical fuel element using Fourier transform methods: One and two group cases

    International Nuclear Information System (INIS)

    Williams, M.M.R.; Hall, S.K.; Eaton, M.D.

    2014-01-01

    Highlights: • A rectangular reactor cell with an elliptical fuel element. • Solution of transport and diffusion equations by Fourier expansion. • Numerical examples showing convergence. • Two group cell problems. - Abstract: A method for solving the diffusion and transport equations in a rectangular lattice cell with an elliptical fuel element has been developed using a Fourier expansion of the neutron flux. The method is applied to a one group model with a source in the moderator. The cell flux is obtained and also the associated disadvantage factor. In addition to the one speed case, we also consider the two group equations in the cell which now become an eigenvalue problem for the lattice multiplication factor. The method of solution relies upon an efficient procedure to solve a large set of simultaneous linear equations and for this we use the IMSL library routines. Our method is compared with the results from a finite element code. The main drawback of the problem arises from the very large number of terms required in the Fourier series which taxes the storage and speed of the computer. Nevertheless, useful solutions are obtained in geometries that would normally require the use of finite element or analogous methods, for this reason the Fourier method is useful for comparison with that type of numerical approach. Extension of the method to more intricate fuel shapes, such as stars and cruciforms as well as superpositions of these, is possible

  1. On the observation of a huge lattice contraction and crystal habit modifications in LiMn2O4 prepared by a fuel assisted solution combustion

    International Nuclear Information System (INIS)

    Ragavendran, K.; Sherwood, D.; Vasudevan, D.; Emmanuel, Bosco

    2009-01-01

    Two batches of poly-crystalline lithium manganate were prepared by a fuel assisted solution combustion method. LiMn 2 O 4 (S) was prepared using starch as the fuel and LiMn 2 O 4 (P) was prepared using poly vinyl alcohol (PVA) as the fuel. XRD studies indicated a significant and consistent shift in the 2θ values of all the hkl peaks to higher values in LiMn 2 O 4 (P) compared to LiMn 2 O 4 (S) indicating a lattice contraction in the former. TG/DTA studies indicated a higher formation temperature (∼25 deg. C higher) for LiMn 2 O 4 (P). The higher formation temperature most likely promotes the oxidation of some Mn 3+ to Mn 4+ with a lower ionic radius causing a lattice contraction. This hypothesis is confirmed through XPS studies which indicated the presence of a higher fraction of Mn 4+ in LiMn 2 O 4 (P) than that present in LiMn 2 O 4 (S). A crystal shape algorithm was used to generate the crystal habits of lithium manganate from their XRD data leading to an understanding on the exposed hkl planes in these materials. From the atomic arrangement on the exposed hkl planes it is predicted that LiMn 2 O 4 (P) would be less prone to manganese dissolution and hence would possess a higher cycle life when compared to LiMn 2 O 4 (S).

  2. On the observation of a huge lattice contraction and crystal habit modifications in LiMn 2O 4 prepared by a fuel assisted solution combustion

    Science.gov (United States)

    Ragavendran, K.; Sherwood, D.; Vasudevan, D.; Emmanuel, Bosco

    2009-08-01

    Two batches of poly-crystalline lithium manganate were prepared by a fuel assisted solution combustion method. LiMn 2O 4(S) was prepared using starch as the fuel and LiMn 2O 4(P) was prepared using poly vinyl alcohol (PVA) as the fuel. XRD studies indicated a significant and consistent shift in the 2 θ values of all the hkl peaks to higher values in LiMn 2O 4(P) compared to LiMn 2O 4(S) indicating a lattice contraction in the former. TG/DTA studies indicated a higher formation temperature (∼25 °C higher) for LiMn 2O 4(P). The higher formation temperature most likely promotes the oxidation of some Mn 3+ to Mn 4+ with a lower ionic radius causing a lattice contraction. This hypothesis is confirmed through XPS studies which indicated the presence of a higher fraction of Mn 4+ in LiMn 2O 4(P) than that present in LiMn 2O 4(S). A crystal shape algorithm was used to generate the crystal habits of lithium manganate from their XRD data leading to an understanding on the exposed hkl planes in these materials. From the atomic arrangement on the exposed hkl planes it is predicted that LiMn 2O 4(P) would be less prone to manganese dissolution and hence would possess a higher cycle life when compared to LiMn 2O 4(S).

  3. In-core fuel management code package validation for BWRs

    International Nuclear Information System (INIS)

    1995-12-01

    The main goal of the present CRP (Coordinated Research Programme) was to develop benchmarks which are appropriate to check and improve the fuel management computer code packages and their procedures. Therefore, benchmark specifications were established which included a set of realistic data for running in-core fuel management codes. Secondly, the results of measurements and/or operating data were also provided to verify and compare with these parameters as calculated by the in-core fuel management codes or code packages. For the BWR it was established that the Mexican Laguna Verde 1 BWR would serve as the model for providing data on the benchmark specifications. It was decided to provide results for the first 2 cycles of Unit 1 of the Laguna Verde reactor. The analyses of the above benchmarks are performed in two stages. In the first stage, the lattice parameters are generated as a function of burnup at different voids and with and without control rod. These lattice parameters form the input for 3-dimensional diffusion theory codes for over-all reactor analysis. The lattice calculations were performed using different methods, such as, Monte Carlo, 2-D integral transport theory methods. Supercell Model and transport-diffusion model with proper correction for burnable absorber. Thus the variety of results should provide adequate information for any institute or organization to develop competence to analyze In-core fuel management codes. 15 refs, figs and tabs

  4. Regulatory Benchmarking

    DEFF Research Database (Denmark)

    Agrell, Per J.; Bogetoft, Peter

    2017-01-01

    Benchmarking methods, and in particular Data Envelopment Analysis (DEA), have become well-established and informative tools for economic regulation. DEA is now routinely used by European regulators to set reasonable revenue caps for energy transmission and distribution system operators. The appli......Benchmarking methods, and in particular Data Envelopment Analysis (DEA), have become well-established and informative tools for economic regulation. DEA is now routinely used by European regulators to set reasonable revenue caps for energy transmission and distribution system operators....... The application of bench-marking in regulation, however, requires specific steps in terms of data validation, model specification and outlier detection that are not systematically documented in open publications, leading to discussions about regulatory stability and economic feasibility of these techniques...

  5. Regulatory Benchmarking

    DEFF Research Database (Denmark)

    Agrell, Per J.; Bogetoft, Peter

    2017-01-01

    Benchmarking methods, and in particular Data Envelopment Analysis (DEA), have become well-established and informative tools for economic regulation. DEA is now routinely used by European regulators to set reasonable revenue caps for energy transmission and distribution system operators. The appli......Benchmarking methods, and in particular Data Envelopment Analysis (DEA), have become well-established and informative tools for economic regulation. DEA is now routinely used by European regulators to set reasonable revenue caps for energy transmission and distribution system operators....... The application of benchmarking in regulation, however, requires specific steps in terms of data validation, model specification and outlier detection that are not systematically documented in open publications, leading to discussions about regulatory stability and economic feasibility of these techniques...

  6. Vver-1000 Mox core computational benchmark

    International Nuclear Information System (INIS)

    2006-01-01

    The NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, fuel performance and fuel cycle issues related to disposing of weapons-grade plutonium in mixed-oxide fuel. The objectives of the group are to provide NEA member countries with up-to-date information on, and to develop consensus regarding, core and fuel cycle issues associated with burning weapons-grade plutonium in thermal water reactors (PWR, BWR, VVER-1000, CANDU) and fast reactors (BN-600). These issues concern core physics, fuel performance and reliability, and the capability and flexibility of thermal water reactors and fast reactors to dispose of weapons-grade plutonium in standard fuel cycles. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close co-operation (jointly, in most cases) with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A prominent part of these activities include benchmark studies. At the time of preparation of this report, the following benchmarks were completed or in progress: VENUS-2 MOX Core Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); VVER-1000 LEU and MOX Benchmark (completed); KRITZ-2 Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); Hollow and Solid MOX Fuel Behaviour Benchmark (completed); PRIMO MOX Fuel Performance Benchmark (ongoing); VENUS-2 MOX-fuelled Reactor Dosimetry Calculation (ongoing); VVER-1000 In-core Self-powered Neutron Detector Calculational Benchmark (started); MOX Fuel Rod Behaviour in Fast Power Pulse Conditions (started); Benchmark on the VENUS Plutonium Recycling Experiments Configuration 7 (started). This report describes the detailed results of the benchmark investigating the physics of a whole VVER-1000 reactor core using two-thirds low-enriched uranium (LEU) and one-third MOX fuel. It contributes to the computer code certification process and to the

  7. Monte Carlo code criticality benchmark comparisons for waste packaging

    International Nuclear Information System (INIS)

    Alesso, H.P.; Annese, C.E.; Buck, R.M.; Pearson, J.S.; Lloyd, W.R.

    1992-07-01

    COG is a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The objective of this paper is to report on COG results for criticality benchmark experiments both on a Cray mainframe and on a HP 9000 workstation. COG has been recently ported to workstations to improve its accessibility to a wider community of users. COG has some similarities to a number of other computer codes used in the shielding and criticality community. The recently introduced high performance reduced instruction set (RISC) UNIX workstations provide computational power that approach mainframes at a fraction of the cost. A version of COG is currently being developed for the Hewlett Packard 9000/730 computer with a UNIX operating system. Subsequent porting operations will move COG to SUN, DEC, and IBM workstations. In addition, a CAD system for preparation of the geometry input for COG is being developed. In July 1977, Babcock ampersand Wilcox Co. (B ampersand W) was awarded a contract to conduct a series of critical experiments that simulated close-packed storage of LWR-type fuel. These experiments provided data for benchmarking and validating calculational methods used in predicting K-effective of nuclear fuel storage in close-packed, neutron poisoned arrays. Low enriched UO2 fuel pins in water-moderated lattices in fuel storage represent a challenging criticality calculation for Monte Carlo codes particularly when the fuel pins extend out of the water. COG and KENO calculational results of these criticality benchmark experiments are presented

  8. Benchmarking the expected stack manufacturing cost of next generation, intermediate-temperature protonic ceramic fuel cells with solid oxide fuel cell technology

    Science.gov (United States)

    Dubois, Alexis; Ricote, Sandrine; Braun, Robert J.

    2017-11-01

    Recent progress in the performance of intermediate temperature (500-600 °C) protonic ceramic fuel cells (PCFCs) has demonstrated both fuel flexibility and increasing power density that approach commercial application requirements. These developments may eventually position the technology as a viable alternative to solid oxide fuel cells (SOFCs) and molten carbonate fuel cells (MCFCs). The PCFCs investigated in this work are based on a BaZr0.8Y0.2O3-δ (BZY20) thin electrolyte supported by BZY20/Ni porous anodes, and a triple conducting cathode material comprised of BaCo0.4Fe0.4Zr0.1Y0.1O3-δ (BCFZY0.1). These cells are prepared using a low-cost solid-state reactive sintering (SSRS) process, and are capable of power densities of 0.156 W cm-2 at 500 °C operating directly from methane fuel. We develop a manufacturing cost model to estimate the Nth generation production costs of PCFC stack technology using high volume manufacturing processes and compare them to the state-of-the-art in SOFC technology. The low-cost cell manufacturing enabled by the SSRS technique compensates for the lower PCFC power density and the trade-off between operating temperature and efficiency enables the use of lower-cost stainless steel materials. PCFC stack production cost estimates are found to be as much as 27-37% lower at 550 °C than SOFCs operating at 800 °C.

  9. Use of WIMS-E lattice code for prediction of the transuranic source term for spent fuel dose estimation

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.

    1996-01-01

    A recent source term analysis has shown a discrepancy between ORIGEN2 transuranic isotopic production estimates and those produced with the WIMS-E lattice physics code. Excellent agreement between relevant experimental measurements and WIMS-E was shown, thus exposing an error in the cross section library used by ORIGEN2

  10. Generation of seven group cross section library for TRIGA LEU fuel in CITATION format and benchmarking some experimental and operational data

    International Nuclear Information System (INIS)

    Sarker, M.M.; Bhuiyan, S.I.; Akramuzzaman, M.

    2007-01-01

    The principal objective of this study is to validate the seven group cross section library in CITATION format for TRIGA LEU Fuel. This presentation deals with the 'generation of a cross section library for the CITATION and its validation. We used WIMSD-5B version for the generation of all group constants. The overall strategy is: (1) use WIMS package to generate few group neutron macroscopic cross section (cell constants) for all of the materials in the core and its immediate neighborhood (2) use 3-D code CITATION to perform the global analysis of the core to study: multiplication factor, neutron flux distribution and power peaking factors. Various options available in WIMS program were studied in depth to finalize the models to generate the most appropriate group constants. For the global analysis the code CITATION and a post processing program FCAP were chosen. Thus a seven group cross section library for the calculations of TRIGA Research Reactor was generated. To investigate the validity of the generated library a critical experiment of the TRIGA research reactor was benchmarked. (author)

  11. MOx Depletion Calculation Benchmark

    International Nuclear Information System (INIS)

    San Felice, Laurence; Eschbach, Romain; Dewi Syarifah, Ratna; Maryam, Seif-Eddine; Hesketh, Kevin

    2016-01-01

    Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of Reactor Systems (WPRS) has been established to study the reactor physics, fuel performance, radiation transport and shielding, and the uncertainties associated with modelling of these phenomena in present and future nuclear power systems. The WPRS has different expert groups to cover a wide range of scientific issues in these fields. The Expert Group on Reactor Physics and Advanced Nuclear Systems (EGRPANS) was created in 2011 to perform specific tasks associated with reactor physics aspects of present and future nuclear power systems. EGRPANS provides expert advice to the WPRS and the nuclear community on the development needs (data and methods, validation experiments, scenario studies) for different reactor systems and also provides specific technical information regarding: core reactivity characteristics, including fuel depletion effects; core power/flux distributions; Core dynamics and reactivity control. In 2013 EGRPANS published a report that investigated fuel depletion effects in a Pressurised Water Reactor (PWR). This was entitled 'International Comparison of a Depletion Calculation Benchmark on Fuel Cycle Issues' NEA/NSC/DOC(2013) that documented a benchmark exercise for UO 2 fuel rods. This report documents a complementary benchmark exercise that focused on PuO 2 /UO 2 Mixed Oxide (MOX) fuel rods. The results are especially relevant to the back-end of the fuel cycle, including irradiated fuel transport, reprocessing, interim storage and waste repository. Saint-Laurent B1 (SLB1) was the first French reactor to use MOx assemblies. SLB1 is a 900 MWe PWR, with 30% MOx fuel loading. The standard MOx assemblies, used in Saint-Laurent B1 reactor, include three zones with different plutonium enrichments, high Pu content (5.64%) in the center zone, medium Pu content (4.42%) in the intermediate zone and low Pu content (2.91%) in the peripheral zone

  12. Lattice QCD

    International Nuclear Information System (INIS)

    Hasenfratz, P.

    1983-01-01

    The author presents a general introduction to lattice gauge theories and discusses non-perturbative methods in the gauge sector. He then shows how the lattice works in obtaining the string tension in SU(2). Lattice QCD at finite physical temperature is discussed. Universality tests in SU(2) lattice QCD are presented. SU(3) pure gauge theory is briefly dealt with. Finally, fermions on the lattice are considered. (Auth.)

  13. Critical experiments with 4.31 wt % 235U-enriched UO2 rods in highly borated water lattices

    International Nuclear Information System (INIS)

    Durst, B.M.; Bierman, S.R.; Clayton, E.D.

    1982-08-01

    A series of critical experiments were performed with 4.31 wt % 235 U enriched UO 2 fuel rods immersed in water containing various concentrations of boron ranging up to 2.55 g/l. The boron was added in the form of boric acid (H 3 BO 3 ). Critical experimental data were obtained for two different lattice pitches wherein the water-to-uranium oxide volume ratios were 1.59 and 1.09. The experiments provide benchmarks on heavily borated systems for use in validating calculational techniques employed in analyzing fuel shipping casks and spent fuel storage systems that may utilize boron for criticality control

  14. Core Benchmarks Descriptions

    International Nuclear Information System (INIS)

    Pavlovichev, A.M.

    2001-01-01

    Actual regulations while designing of new fuel cycles for nuclear power installations comprise a calculational justification to be performed by certified computer codes. It guarantees that obtained calculational results will be within the limits of declared uncertainties that are indicated in a certificate issued by Gosatomnadzor of Russian Federation (GAN) and concerning a corresponding computer code. A formal justification of declared uncertainties is the comparison of calculational results obtained by a commercial code with the results of experiments or of calculational tests that are calculated with an uncertainty defined by certified precision codes of MCU type or of other one. The actual level of international cooperation provides an enlarging of the bank of experimental and calculational benchmarks acceptable for a certification of commercial codes that are being used for a design of fuel loadings with MOX fuel. In particular, the work is practically finished on the forming of calculational benchmarks list for a certification of code TVS-M as applied to MOX fuel assembly calculations. The results on these activities are presented

  15. Benchmarking the new JENDL-4.0 library on criticality experiments of a research reactor with oxide LEU (20 w/o) fuel, light water moderator and beryllium reflectors

    International Nuclear Information System (INIS)

    Liem, Peng Hong; Sembiring, Tagor Malem

    2012-01-01

    Highlights: ► Benchmark calculations of the new JENDL-4.0 library. ► Thermal research reactor with oxide LEU fuel, H 2 O moderator and Be reflector. ► JENDL-4.0 library shows better C/E values for criticality evaluations. - Abstract: Benchmark calculations of the new JENDL-4.0 library on the criticality experiments of a thermal research reactor with oxide low enriched uranium (LEU, 20 w/o) fuel, light water moderator and beryllium reflector (RSG GAS) have been conducted using a continuous energy Monte Carlo code, MVP-II. The JENDL-4.0 library shows better C/E values compared to the former library JENDL-3.3 and other world-widely used latest libraries (ENDF/B-VII.0 and JEFF-3.1).

  16. Evaluation of fire models for nuclear power plant applications. Benchmark exercise no. 4: Fuel pool fire inside a compartment - International panel report

    International Nuclear Information System (INIS)

    Klein-Hessling, W.; Roewekamp, M.; Riese, O.

    2006-11-01

    Fire simulations as well as their analytical validation procedures have gained more and more significance, particularly in the context of the fire safety analysis for operating nuclear power plants. Meanwhile, fire simulation models have been adapted as analytical tools for a risk oriented fire safety assessment. Calculated predictions can be used, on the one hand, for the improvements and upgrades of fire protection in nuclear power plants by the licensees and, on the other hand, as a tool for reproducible and clearly understandable estimations in assessing the available and/or foreseen fire protection measures by the authorities and their experts. For consideration of such aspects in the context of implementing new nuclear fire protection standards or of updating existing ones, an 'International Collaborative Project to Evaluate Fire Models for Nuclear Power Plant Applications' also known as the 'International Collaborative Fire Model Project' (ICFMP) was started in 1999. It has made use of the experience and knowledge of a variety of worldwide expert institutions in this field to assess and improve, if necessary, the state-of-the-art with respect to modeling fires in nuclear power plants and other nuclear installations. This document contains the results of the ICFMP Benchmark Exercise No. 4, where two fuel pool fire experiments in an enclosure with two different natural vent sizes have been considered. Analyzing the results of different fire simulation codes and code types provides some indications with respect to the uncertainty of the results. This information is especially important in setting uncertainty parameters in probabilistic risk studies and to provide general insights concerning the applicability and limitations in the application of different types of fire simulation codes for this type of fire scenario and boundary conditions. During the benchmark procedure the participants performed different types of calculations. These included totally blind

  17. Accuracy of cell calculation methods used for analysis of high conversion light water reactor lattice

    International Nuclear Information System (INIS)

    Jeong, Chang-Joon; Okumura, Keisuke; Ishiguro, Yukio; Tanaka, Ken-ichi

    1990-01-01

    Validation tests were made for the accuracy of cell calculation methods used in analyses of tight lattices of a mixed-oxide (MOX) fuel core in a high conversion light water reactor (HCLWR). A series of cell calculations was carried out for the lattices referred from an international HCLWR benchmark comparison, with emphasis placed on the resonance calculation methods; the NR, IR approximations, the collision probability method with ultra-fine energy group. Verification was also performed for the geometrical modelling; a hexagonal/cylindrical cell, and the boundary condition; mirror/white reflection. In the calculations, important reactor physics parameters, such as the neutron multiplication factor, the conversion ratio and the void coefficient, were evaluated using the above methods for various HCLWR lattices with different moderator to fuel volume ratios, fuel materials and fissile plutonium enrichments. The calculated results were compared with each other, and the accuracy and applicability of each method were clarified by comparison with continuous energy Monte Carlo calculations. It was verified that the accuracy of the IR approximation became worse when the neutron spectrum became harder. It was also concluded that the cylindrical cell model with the white boundary condition was not so suitable for MOX fuelled lattices, as for UO 2 fuelled lattices. (author)

  18. Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  20. Experiments on light water lattices with enriched uranium fuel; Analyse des donnees experimentales sur les reseaux a eau legere et uranium enrichi

    Energy Technology Data Exchange (ETDEWEB)

    Audinet, M [Societe des Forges et Ateliers du Creusot, 75 - Paris (France); Lamare, J de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Panossian, J [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    Experiments a light water lattices with slightly enriched uranium fuel, have been performed at Brookhaven and Bettis Plant Laboratories. The results are studied and compared with simple theories on reactor calculations. By taking into account shadow effects and non Maxwellian neutron spectrum, which are important in this kind of reactors, we have been able to explain the observed results fairly well. We can thus give a constituent set of formulas with which to calculate lattices similar to there we studied. (author) [French] Les resultats d'experiences effectuees aux Laboratoires de Brookbaven et de Bettis Plant, sur des reseaux heterogenes a eau legere et uranium metallique legerement enrichi, sont analyses et confrontes avec les theories simples du calcul de pile. En tenant compte des effets d'interaction et d'echauffement du spectre de neutrons qui sont importants dans ce type de reacteurs, on parvient a rendre compte convenablement des resultats observes. On a ainsi mis au point un formulaire permettant le calcul des reseaux quivpeuvent etre consideres comme assez semblables aux reseaux etudies. (auteur)

  1. On the thermal scattering law data for reactor lattice calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Mattes, M.

    2004-01-01

    Thermal scattering law data for hydrogen bound in water, hydrogen bound in zirconium hydride and deuterium bound in heavy water have been re-evaluated. The influence of the thermal scattering law data on critical lattices has been studied with detailed Monte Carlo calculations and a summary of results is presented for a numerical benchmark and for the TRIGA reactor benchmark. Systematics for a large sequence of benchmarks analysed with the WIMS-D lattice code are also presented. (author)

  2. Water droplet dynamic behavior during removal from a proton exchange membrane fuel cell gas diffusion layer by Lattice-Boltzmann method

    Energy Technology Data Exchange (ETDEWEB)

    Molaeimanesh, Golamreza; Akbari, Mohammad Hadi [Shiraz University, Shiraz (Iran, Islamic Republic of)

    2014-04-15

    A major challenge in the application of proton exchange membrane fuel cells (PEMFCs) is water management, with the flooding of electrodes as the main issue. The Lattice-Boltzmann method (LBM) is a relatively new technique that is superior in modeling the dynamic interface of multiphase fluid flow in complex microstructures such as non-homogeneous and anisotropic porous media of PEMFC electrodes. In this study, the dynamic behavior of a water droplet during removal from gas diffusion layer (GDL) of a PEMFC electrode with interdigitated flow field is simulated using LBM. The effects of GDL wettability and its spanwise and transverse gradients on the removal process are investigated. The results demonstrate great influence of wettability and its spanwise and transverse gradients on the dynamic behavior of droplets during the removal process. Although increasing the hydrophobicity of GDL results in better droplet removal, its increase beyond a critical value does not show a significant effect.

  3. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    2001-01-01

    The description of reactor lattice codes is carried out on the example of the WIMSD-5B code. The WIMS code in its various version is the most recognised lattice code. It is used in all parts of the world for calculations of research and power reactors. The version WIMSD-5B is distributed free of charge by NEA Data Bank. The description of its main features given in the present lecture follows the aspects defined previously for lattice calculations in the lecture on Reactor Lattice Transport Calculations. The spatial models are described, and the approach to the energy treatment is given. Finally the specific algorithm applied in fuel depletion calculations is outlined. (author)

  4. IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments

    International Nuclear Information System (INIS)

    2003-01-01

    plutonium utilisation in commercial reactors. A fourth report concerns Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel. The fifth report concerns critical experiments supporting underwater storage of tightly packed configurations of spent fuel pins. The sixth report entitled 'Physics verification program', covers principally a series of experiments to measure the effect of lattice heterogeneities. The seventh report concerns Development and Demonstration of an Advanced Extended-Burnup Fuel- Assembly Design Incorporating Urania-Gadolinia. The eighth report concerns 'Urania-Gadolinia: Nuclear Model Development and Critical Experiment Benchmark'. The purpose was development and verification within an extended-burnup program for pressurised water reactors of advanced fuel assembly design. The ninth report concerns the Characterization and Irradiation Program: Extended-Burnup Gadolinia Lead Test Assembly (Mark GdB). The goal of the program was to extend the burnup of pressurized water reactor fuel assemblies to 50,000 MWd/mtU batch average burnup. The tenth report concerns the Hot Cell Examination of Gadolinia Lead Test Assembly Rods After One Cycle of Irradiation as described in the eighth and ninth report. The eleventh report covering April 1986 through March 1987, combines the progress report for the program entitled Development of an Advanced Extended Burnup Fuel Assembly Design Incorporating Urania-Gadolinia, and the final progress report for the program entitled Qualification of the B and W Mark B Fuel Assembly for High Burnup. The twelfth report describes five lead test assemblies designed, manufactured, characterized, and inserted for irradiation in Oconee Unit 1 cycle 8

  5. Benchmark tests of JENDL-3.2 for thermal and fast reactors

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

    Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

  6. Lattice Boltzmann simulation on the liquid junction potential in a concentration fuel cell. Paper no. IGEC-1-060

    International Nuclear Information System (INIS)

    Park, J.; Huh, K.Y.; Li, X.

    2005-01-01

    The lattice Boltzmann method (LBM) is applied to investigate the liquid junction potential (LJP) at an interface between two electrolyte layers. The Poisson equation for electrostatic field is solved to extend the applicable range to micro and nano scales in which electroneutrality does not hold. The LBM solutions are validated against analytical and finite difference method (FDM) results for evolution of concentration, net charge density and electrostatic potential. Noticeable separation of the concentration profiles of positive and negative ions occurs for kd less than 67 in simulation, where k is the inverse of the thickness of electrical double layer and d is the system length. Parametric study is performed for the peak potential and the time to reach the peak with respect to kd and ξ which is the initial thickness ratio of the lower concentration to entire stream. Simple coding and easy parallelization will allow the LBM to make an efficient analysis tool for complex electrochemical systems. (author)

  7. Optimization of the fuel assembly for the Canadian SuperCritical Water-cooled Reactor (SCWR)

    Energy Technology Data Exchange (ETDEWEB)

    French, C., E-mail: Corey.French@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada); Bonin, H.; Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    An approach to develop a parametric optimization tool to support the Canadian Supercritical Water-cooled Reactor (SCWR) fuel design is presented in this work. The 2D benchmark lattices for 78-pin and 64-pin fuel assemblies are used as the initial models from which fuel performance and subsequent optimization stem from. A tandem optimization procedure is integrated which employs the steepest descent method. The physics codes WIMS-AECL, MCNP6 and SERPENT are used to calculate and verify select performance factors. The results are used as inputs to an optimization algorithm that yield optimal fresh fuel isotopic composition and lattice geometry. Preliminary results on verifications of infinite lattice reactivity are demonstrated in this paper. (author)

  8. Validation of WIMS-CANDU using Pin-Cell Lattices

    International Nuclear Information System (INIS)

    Kim, Won Young; Min, Byung Joo; Park, Joo Hwan

    2006-01-01

    The WIMS-CANDU is a lattice code which has a depletion capability for the analysis of reactor physics problems related to a design and safety. The WIMS-CANDU code has been developed from the WIMSD5B, a version of the WIMS code released from the OECD/NEA data bank in 1998. The lattice code POWDERPUFS-V (PPV) has been used for the physics design and analysis of a natural uranium fuel for the CANDU reactor. However since the application of PPV is limited to a fresh fuel due to its empirical correlations, the WIMS-AECL code has been developed by AECL to substitute the PPV. Also, the WIMS-CANDU code is being developed to perform the physics analysis of the present operating CANDU reactors as a replacement of PPV. As one of the developing work of WIMS-CANDU, the U 238 absorption cross-section in the nuclear data library of WIMS-CANDU was updated and WIMS-CANDU was validated using the benchmark problems for pin-cell lattices such as TRX-1, TRX-2, Bapl-1, Bapl-2 and Bapl-3. The results by the WIMS-CANDU and the WIMS-AECL were compared with the experimental data

  9. Lattice Boltzmann simulations of water transport in gas diffusion layer of a polymer electrolyte membrane fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Hao, Liang; Cheng, Ping [Ministry of Education Key Laboratory of Power Machinery and Engineering, School of Mechanical Engineering, Shanghai Jiaotong University, DongChuan Road 800, Shanghai 200240 (China)

    2010-06-15

    The effect of wettability on water transport dynamics in gas diffusion layer (GDL) is investigated by simulating water invasion in an initially gas-filled GDL using the multiphase free-energy lattice Boltzmann method (LBM). The results show that wettability plays a significant role on water saturation distribution in two-phase flow in the uniform wetting GDL. For highly hydrophobicity, the water transport falls in the regime of capillary fingering, while for neutral wettability, water transport exhibits the characteristic of stable displacement, although both processes are capillary force dominated flow with same capillary numbers. In addition, the introduction of hydrophilic paths in the GDL leads the water to flow through the hydrophilic pores preferentially. The resulting water saturation distributions show that the saturation in the GDL has little change after water breaks through the GDL, and further confirm that the selective introduction of hydrophilic passages in the GDL would facilitate the removal of liquid water more effectively, thus alleviating the flooding in catalyst layer (CL) and GDL. The LBM approach presented in this study provides an effective tool to investigate water transport phenomenon in the GDL at pore-scale level with wettability distribution taken into consideration. (author)

  10. Computer simulation of variform fuel assemblies using Dragon code

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun; Yao Dong

    2005-01-01

    The DRAGON is a cell code that developed for the CANDU reactor by the Ecole Polytechnique de Montreal of CANADA. Although, the DRAGON is mainly used to simulate the CANDU super-cell fuel assembly, it has an ability to simulate other geometries of the fuel assembly. However, only NEACRP benchmark problem of the BWR lattice cell was analyzed until now except for the CANDU reactor. We also need to develop the code to simulate the variform fuel assemblies, especially, for design of the advanced reactor. We validated that the cell code DRAGON is useful for simulating various kinds of the fuel assembly by analyzing the rod-shape fuel assembly of the PWR and the MTR plate-shape fuel assembly. Some other kinds of geometry of geometry were computed. Computational results show that the DRAGON is able to analyze variform fuel assembly problems and the precision is high. (authors)

  11. Benchmarking in Foodservice Operations

    National Research Council Canada - National Science Library

    Johnson, Bonnie

    1998-01-01

    The objective of this study was to identify usage of foodservice performance measures, important activities in foodservice benchmarking, and benchmarking attitudes, beliefs, and practices by foodservice directors...

  12. Polarization response of RHIC electron lens lattices

    International Nuclear Information System (INIS)

    Ranjbar, V. H.; Méot, F.; Bai, M.; Abell, D. T.; Meiser, D.

    2016-01-01

    Depolarization response for a system of two orthogonal snakes at irrational tunes is studied in depth using lattice independent spin integration. Particularly, we consider the effect of overlapping spin resonances in this system, to understand the impact of phase, tune, relative location and threshold strengths of the spin resonances. Furthermore, these results are benchmarked and compared to two dimensional direct tracking results for the RHIC e-lens lattice and the standard lattice. We then consider the effect of longitudinal motion via chromatic scans using direct six dimensional lattice tracking.

  13. Polarization response of RHIC electron lens lattices

    Directory of Open Access Journals (Sweden)

    V. H. Ranjbar

    2016-10-01

    Full Text Available Depolarization response for a system of two orthogonal snakes at irrational tunes is studied in depth using lattice independent spin integration. In particular we consider the effect of overlapping spin resonances in this system, to understand the impact of phase, tune, relative location and threshold strengths of the spin resonances. These results are benchmarked and compared to two dimensional direct tracking results for the RHIC e-lens lattice and the standard lattice. Finally we consider the effect of longitudinal motion via chromatic scans using direct six dimensional lattice tracking.

  14. Void lattices

    International Nuclear Information System (INIS)

    Chadderton, L.T.; Johnson, E.; Wohlenberg, T.

    1976-01-01

    Void lattices in metals apparently owe their stability to elastically anisotropic interactions. An ordered array of voids on the anion sublattice in fluorite does not fit so neatly into this scheme of things. Crowdions may play a part in the formation of the void lattice, and stability may derive from other sources. (Auth.)

  15. Lattice fermions

    International Nuclear Information System (INIS)

    Randjbar-Daemi, S.

    1995-12-01

    The so-called doubling problem in the lattice description of fermions led to a proof that under certain circumstances chiral gauge theories cannot be defined on the lattice. This is called the no-go theorem. It implies that if Γ/sub/A is defined on a lattice then its infrared limit, which should correspond to the quantum description of the classical action for the slowly varying fields on lattice scale, is inevitably a vector like theory. In particular, if not circumvented, the no-go theorem implies that there is no lattice formulation of the Standard Weinberg-Salam theory or SU(5) GUT, even though the fermions belong to anomaly-free representations of the gauge group. This talk aims to explain one possible attempt at bypassing the no-go theorem. 20 refs

  16. Lattice fermions

    Energy Technology Data Exchange (ETDEWEB)

    Randjbar-Daemi, S

    1995-12-01

    The so-called doubling problem in the lattice description of fermions led to a proof that under certain circumstances chiral gauge theories cannot be defined on the lattice. This is called the no-go theorem. It implies that if {Gamma}/sub/A is defined on a lattice then its infrared limit, which should correspond to the quantum description of the classical action for the slowly varying fields on lattice scale, is inevitably a vector like theory. In particular, if not circumvented, the no-go theorem implies that there is no lattice formulation of the Standard Weinberg-Salam theory or SU(5) GUT, even though the fermions belong to anomaly-free representations of the gauge group. This talk aims to explain one possible attempt at bypassing the no-go theorem. 20 refs.

  17. Teledyne Energy Systems, Inc., Proton Exchange Member (PEM) Fuel Cell Engineering Model Powerplant. Test Report: Initial Benchmark Tests in the Original Orientation

    Science.gov (United States)

    Loyselle, Patricia; Prokopius, Kevin

    2011-01-01

    Proton Exchange Membrane (PEM) fuel cell technology is the leading candidate to replace the alkaline fuel cell technology, currently used on the Shuttle, for future space missions. During a 5-yr development program, a PEM fuel cell powerplant was developed. This report details the initial performance evaluation test results of the powerplant.

  18. Large scale simulation of liquid water transport in a gas diffusion layer of polymer electrolyte membrane fuel cells using the lattice Boltzmann method

    Science.gov (United States)

    Sakaida, Satoshi; Tabe, Yutaka; Chikahisa, Takemi

    2017-09-01

    A method for the large-scale simulation with the lattice Boltzmann method (LBM) is proposed for liquid water movement in a gas diffusion layer (GDL) of polymer electrolyte membrane fuel cells. The LBM is able to analyze two-phase flows in complex structures, however the simulation domain is limited due to heavy computational loads. This study investigates a variety means to reduce computational loads and increase the simulation areas. One is applying an LBM treating two-phases as having the same density, together with keeping numerical stability with large time steps. The applicability of this approach is confirmed by comparing the results with rigorous simulations using actual density. The second is establishing the maximum limit of the Capillary number that maintains flow patterns similar to the precise simulation; this is attempted as the computational load is inversely proportional to the Capillary number. The results show that the Capillary number can be increased to 3.0 × 10-3, where the actual operation corresponds to Ca = 10-5∼10-8. The limit is also investigated experimentally using an enlarged scale model satisfying similarity conditions for the flow. Finally, a demonstration is made of the effects of pore uniformity in GDL as an example of a large-scale simulation covering a channel.

  19. Lattice strings

    International Nuclear Information System (INIS)

    Thorn, C.B.

    1988-01-01

    The possibility of studying non-perturbative effects in string theory using a world sheet lattice is discussed. The light-cone lattice string model of Giles and Thorn is studied numerically to assess the accuracy of ''coarse lattice'' approximations. For free strings a 5 by 15 lattice seems sufficient to obtain better than 10% accuracy for the bosonic string tachyon mass squared. In addition a crude lattice model simulating string like interactions is studied to find out how easily a coarse lattice calculation can pick out effects such as bound states which would qualitatively alter the spectrum of the free theory. The role of the critical dimension in obtaining a finite continuum limit is discussed. Instead of the ''gaussian'' lattice model one could use one of the vertex models, whose continuum limit is the same as a gaussian model on a torus of any radius. Indeed, any critical 2 dimensional statistical system will have a stringy continuum limit in the absence of string interactions. 8 refs., 1 fig. , 9 tabs

  20. AER benchmark specification sheet

    International Nuclear Information System (INIS)

    Aszodi, A.; Toth, S.

    2009-01-01

    In the VVER-440/213 type reactors, the core outlet temperature field is monitored with in-core thermocouples, which are installed above 210 fuel assemblies. These measured temperatures are used in determination of the fuel assembly powers and they have important role in the reactor power limitation. For these reasons, correct interpretation of the thermocouple signals is an important question. In order to interpret the signals in correct way, knowledge of the coolant mixing in the assembly heads is necessary. Computational Fluid Dynamics (CFD) codes and experiments can help to understand better these mixing processes and they can provide information which can support the more adequate interpretation of the thermocouple signals. This benchmark deals with the 3D CFD modeling of the coolant mixing in the heads of the profiled fuel assemblies with 12.2 mm rod pitch. Two assemblies of the 23rd cycle of the Paks NPP's Unit 3 are investigated. One of them has symmetrical pin power profile and another possesses inclined profile. (authors)

  1. AER Benchmark Specification Sheet

    International Nuclear Information System (INIS)

    Aszodi, A.; Toth, S.

    2009-01-01

    In the WWER-440/213 type reactors, the core outlet temperature field is monitored with in-core thermocouples, which are installed above 210 fuel assemblies. These measured temperatures are used in determination of the fuel assembly powers and they have important role in the reactor power limitation. For these reasons, correct interpretation of the thermocouple signals is an important question. In order to interpret the signals in correct way, knowledge of the coolant mixing in the assembly heads is necessary. Computational fluid dynamics codes and experiments can help to understand better these mixing processes and they can provide information which can support the more adequate interpretation of the thermocouple signals. This benchmark deals with the 3D computational fluid dynamics modeling of the coolant mixing in the heads of the profiled fuel assemblies with 12.2 mm rod pitch. Two assemblies of the twenty third cycle of the Paks NPPs Unit 3 are investigated. One of them has symmetrical pin power profile and another possesses inclined profile. (Authors)

  2. Benchmarking biofuels; Biobrandstoffen benchmarken

    Energy Technology Data Exchange (ETDEWEB)

    Croezen, H.; Kampman, B.; Bergsma, G.

    2012-03-15

    A sustainability benchmark for transport biofuels has been developed and used to evaluate the various biofuels currently on the market. For comparison, electric vehicles, hydrogen vehicles and petrol/diesel vehicles were also included. A range of studies as well as growing insight are making it ever clearer that biomass-based transport fuels may have just as big a carbon footprint as fossil fuels like petrol or diesel, or even bigger. At the request of Greenpeace Netherlands, CE Delft has brought together current understanding on the sustainability of fossil fuels, biofuels and electric vehicles, with particular focus on the performance of the respective energy carriers on three sustainability criteria, with the first weighing the heaviest: (1) Greenhouse gas emissions; (2) Land use; and (3) Nutrient consumption [Dutch] Greenpeace Nederland heeft CE Delft gevraagd een duurzaamheidsmeetlat voor biobrandstoffen voor transport te ontwerpen en hierop de verschillende biobrandstoffen te scoren. Voor een vergelijk zijn ook elektrisch rijden, rijden op waterstof en rijden op benzine of diesel opgenomen. Door onderzoek en voortschrijdend inzicht blijkt steeds vaker dat transportbrandstoffen op basis van biomassa soms net zoveel of zelfs meer broeikasgassen veroorzaken dan fossiele brandstoffen als benzine en diesel. CE Delft heeft voor Greenpeace Nederland op een rijtje gezet wat de huidige inzichten zijn over de duurzaamheid van fossiele brandstoffen, biobrandstoffen en elektrisch rijden. Daarbij is gekeken naar de effecten van de brandstoffen op drie duurzaamheidscriteria, waarbij broeikasgasemissies het zwaarst wegen: (1) Broeikasgasemissies; (2) Landgebruik; en (3) Nutriëntengebruik.

  3. Benchmarking and Performance Measurement.

    Science.gov (United States)

    Town, J. Stephen

    This paper defines benchmarking and its relationship to quality management, describes a project which applied the technique in a library context, and explores the relationship between performance measurement and benchmarking. Numerous benchmarking methods contain similar elements: deciding what to benchmark; identifying partners; gathering…

  4. Study of the lattice parameter evolution of PWR irradiated MOX fuel by X-Ray diffraction; Etude de l'evolution du parametre cristallin des combustibles MOX irradies en rep par la methode de diffraction des rayons X

    Energy Technology Data Exchange (ETDEWEB)

    Clavier, B

    1995-07-01

    Fuel irradiation leads to a swelling resulting from the formation of gaseous (Kr, Xe) or solid fission products which are found either in solution or as solid inclusions in the matrix. This phenomena has to be evaluated to be taken into account in fuel cladding Interaction. Fuel swelling was studied as a function of burn up by measuring the corresponding cell constant evolution by X-Ray diffraction. This study was realized on Mixed Oxide Fuels (MOX) irradiated in a Pressurized Water Reactor (PWR) at different burn-up for 3 initial Pu contents. Lattice parameter evolutions were followed as a function of burn-up for the irradiated fuel with and without an annealing thermal treatment. These experimental evolutions are compared to the theoretical evolutions calculated from the hard sphere model, using the fission product concentrations determined by the APPOLO computer code. Contribution of varying parameters influencing the unit cell value is discussed. Thermal treatment effects were checked by metallography, X-Ray diffraction and microprobe analysis. After thermal treatment, no structural change was observed but a decrease of the lattice parameter was measured. This modification results essentially from self-irradiation defect annealing and not from stoichiometry variations. Microprobe analysis showed that about 15% of the formed Molybdenum is in solid solution In the oxide matrix. Micrographs showed the existence of Pu packs in the oxide matrix which induces a broadening of diffraction lines. The RIETVELD method used to analyze the X-Ray patterns did not allow to characterize independently the Pu packs and the oxide matrix lattice parameters. Nevertheless, with this method, the presence of micro-strains in the irradiated nuclear fuel could be confirmed. (author)

  5. Benchmarking in the Netherlands

    International Nuclear Information System (INIS)

    1999-01-01

    In two articles an overview is given of the activities in the Dutch industry and energy sector with respect to benchmarking. In benchmarking operational processes of different competitive businesses are compared to improve your own performance. Benchmark covenants for energy efficiency between the Dutch government and industrial sectors contribute to a growth of the number of benchmark surveys in the energy intensive industry in the Netherlands. However, some doubt the effectiveness of the benchmark studies

  6. ISABELLE lattice

    International Nuclear Information System (INIS)

    Smith, L.

    1975-01-01

    An analysis is given of a number of variants of the basic lattice of the planned ISABELLE storage rings. The variants were formed by removing cells from the normal part of the lattice and juggling the lengths of magnets, cells, and insertions in order to maintain a rational relation of circumference to that of the AGS and approximately the same dispersion. Special insertions, correction windings, and the working line with nonlinear resonances are discussed

  7. Validation of VHTRC calculation benchmark of critical experiment using the MCB code

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2016-01-01

    Full Text Available The calculation benchmark problem Very High Temperature Reactor Critical (VHTR a pin-in-block type core critical assembly has been investigated with the Monte Carlo Burnup (MCB code in order to validate the latest version of Nuclear Data Library based on ENDF format. Executed benchmark has been made on the basis of VHTR benchmark available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments. This benchmark is useful for verifying the discrepancies in keff values between various libraries and experimental values. This allows to improve accuracy of the neutron transport calculations that may help in designing the high performance commercial VHTRs. Almost all safety parameters depend on the accuracy of neutron transport calculation results that, in turn depend on the accuracy of nuclear data libraries. Thus, evaluation of the libraries applicability to VHTR modelling is one of the important subjects. We compared the numerical experiment results with experimental measurements using two versions of available nuclear data (ENDF-B-VII.1 and JEFF-3.2 prepared for required temperatures. Calculations have been performed with the MCB code which allows to obtain very precise representation of complex VHTR geometry, including the double heterogeneity of a fuel element. In this paper, together with impact of nuclear data, we discuss also the impact of different lattice modelling inside the fuel pins. The discrepancies of keff have been successfully observed and show good agreement with each other and with the experimental data within the 1 σ range of the experimental uncertainty. Because some propagated discrepancies observed, we proposed appropriate corrections in experimental constants which can improve the reactivity coefficient dependency. Obtained results confirm the accuracy of the new Nuclear Data Libraries.

  8. Quantum lattice model solver HΦ

    Science.gov (United States)

    Kawamura, Mitsuaki; Yoshimi, Kazuyoshi; Misawa, Takahiro; Yamaji, Youhei; Todo, Synge; Kawashima, Naoki

    2017-08-01

    HΦ [aitch-phi ] is a program package based on the Lanczos-type eigenvalue solution applicable to a broad range of quantum lattice models, i.e., arbitrary quantum lattice models with two-body interactions, including the Heisenberg model, the Kitaev model, the Hubbard model and the Kondo-lattice model. While it works well on PCs and PC-clusters, HΦ also runs efficiently on massively parallel computers, which considerably extends the tractable range of the system size. In addition, unlike most existing packages, HΦ supports finite-temperature calculations through the method of thermal pure quantum (TPQ) states. In this paper, we explain theoretical background and user-interface of HΦ. We also show the benchmark results of HΦ on supercomputers such as the K computer at RIKEN Advanced Institute for Computational Science (AICS) and SGI ICE XA (Sekirei) at the Institute for the Solid State Physics (ISSP).

  9. IAEA sodium void reactivity benchmark calculations

    International Nuclear Information System (INIS)

    Hill, R.N.; Finck, P.J.

    1992-01-01

    In this paper, the IAEA-1 992 ''Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated

  10. Monte Carlo analysis of experiments on the reactivity temperature coefficient for UO2 and MOX light water moderated lattices

    International Nuclear Information System (INIS)

    Erradi, L.; Chetaine, A.; Chakir, E.; Kharchaf, A.; Elbardouni, T.; Elkhoukhi, T.

    2005-01-01

    In a previous work, we have analysed the main French experiments available on the reactivity temperature coefficient (RTC): CREOLE and MISTRAL experiments. In these experiments, the RTC has been measured in both UO 2 and UO 2 -PuO 2 PWR type lattices. Our calculations, using APOLLO2 code with CEA93 library based on JEF2.2 evaluation, have shown that the calculation error in UO 2 lattices is less than 1 pcm/C degrees which is considered as the target accuracy. On the other hand the calculation error in the MOX lattices is more significant in both low and high temperature ranges: an average error of -2 ± 0.5 pcm/C degrees is observed in low temperatures and an error of +3 ± 2 pcm/C degrees is obtained for temperatures higher than 250 C degrees. In the present work, we analysed additional experimental benchmarks on the RTC of UO 2 and MOX light water moderated lattices. To analyze these benchmarks and with the aim of minimizing uncertainties related to modelling of the experimental set up, we chose the Monte Carlo method which has the advantage of taking into account in the most exact manner the geometry of the experimental configurations. This analysis shows for the UO 2 lattices, a maximum experiment-calculation deviation of about 0,7 pcm/C degrees, which is below the target accuracy for this type of lattices. For the KAMINI experiment, which relates to the measurement of the RTC in a light water moderated lattice using U-233 as fuel our analysis shows that the ENDF/B6 library gives the best result, with an experiment-calculation deviation of the order of -0,16 pcm/C degrees. The analysis of the benchmarks using MOX fuel made it possible to highlight a discrepancy between experiment and calculation on the RTC of about -0.7 pcm/C degrees (for a range of temperatures going from 20 to 248 C degrees) and -1,2 pcm/C degrees (for a range of temperatures going from 20 to 80 C degrees). This result, in particular the tendency which has the error to decrease when the

  11. Supersymmetric lattices

    International Nuclear Information System (INIS)

    Catterall, Simon

    2013-01-01

    Discretization of supersymmetric theories is an old problem in lattice field theory. It has resisted solution until quite recently when new ideas drawn from orbifold constructions and topological field theory have been brought to bear on the question. The result has been the creation of a new class of lattice gauge theory in which the lattice action is invariant under one or more supersymmetries. The resultant theories are local and free of doublers and in the case of Yang-Mills theories also possess exact gauge invariance. In principle they form the basis for a truly non-perturbative definition of the continuum supersymmetric field theory. In this talk these ideas are reviewed with particular emphasis being placed on N = 4 super Yang-Mills theory.

  12. Benchmarking of nuclear economics tools

    International Nuclear Information System (INIS)

    Moore, Megan; Korinny, Andriy; Shropshire, David; Sadhankar, Ramesh

    2017-01-01

    Highlights: • INPRO and GIF economic tools exhibited good alignment in total capital cost estimation. • Subtle discrepancies in the cost result from differences in financing and the fuel cycle assumptions. • A common set of assumptions was found to reduce the discrepancies to 1% or less. • Opportunities for harmonisation of economic tools exists. - Abstract: Benchmarking of the economics methodologies developed by the Generation IV International Forum (GIF) and the International Atomic Energy Agency’s International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), was performed for three Generation IV nuclear energy systems. The Economic Modeling Working Group of GIF developed an Excel based spreadsheet package, G4ECONS (Generation 4 Excel-based Calculation Of Nuclear Systems), to calculate the total capital investment cost (TCIC) and the levelised unit energy cost (LUEC). G4ECONS is sufficiently generic in the sense that it can accept the types of projected input, performance and cost data that are expected to become available for Generation IV systems through various development phases and that it can model both open and closed fuel cycles. The Nuclear Energy System Assessment (NESA) Economic Support Tool (NEST) was developed to enable an economic analysis using the INPRO methodology to easily calculate outputs including the TCIC, LUEC and other financial figures of merit including internal rate of return, return of investment and net present value. NEST is also Excel based and can be used to evaluate nuclear reactor systems using the open fuel cycle, MOX (mixed oxide) fuel recycling and closed cycles. A Super Critical Water-cooled Reactor system with an open fuel cycle and two Fast Reactor systems, one with a break-even fuel cycle and another with a burner fuel cycle, were selected for the benchmarking exercise. Published data on capital and operating costs were used for economics analyses using G4ECONS and NEST tools. Both G4ECONS and

  13. Comparison of RSYST and WIMSD-4 performance for gadolinium poisoned lattices

    Energy Technology Data Exchange (ETDEWEB)

    Kulikowska, T; Szczesna, B; Sadowska, B

    1992-06-01

    The participation in the Co-ordinated Research Programme on `Safe Core Management with Burnable Absorbers in VVERs` has created a possibility of validation of our basic calculational tools for advanced lattice calculations. A systematic analysis of the performance of WIMSD-4 and the recently adapted RSYST modular systems has been carried out on the basis of two benchmarks with gadolinium bearing pins. The report consists of a detailed comparison of methods and models available in RSYST and WIMSD-4 followed by calculational results and their discussion. Finally, the conclusions are drawn concerning the applicability of the two codes for clean fuel and gadolinium poisoned reactor lattices. (author). 26 refs, 19 figs, 19 tabs.

  14. Aquatic Life Benchmarks

    Data.gov (United States)

    U.S. Environmental Protection Agency — The Aquatic Life Benchmarks is an EPA-developed set of criteria for freshwater species. These benchmarks are based on toxicity values reviewed by EPA and used in the...

  15. Hydrogen production using the waste heat of Benchmark pressurized Molten carbonate fuel cell system via combination of organic Rankine cycle and proton exchange membrane (PEM) electrolysis

    International Nuclear Information System (INIS)

    Nami, Hossein; Akrami, Ehsan; Ranjbar, Faramarz

    2017-01-01

    Highlights: • Waste heat of the Benchmark system recovered using an ORC. • An integrated system is proposed to produce power and hydrogen. • The effects of some decision parameters on the produced hydrogen have investigated. - Abstract: Energy and exergy analyses are carried out for hydrogen production via combination of Benchmark system and organic Rankine cycle (ORC) coupled with a proton exchange membrane electrolyzer. A parametric study is reported and effects of such organic Rankine cycle significant variables as evaporator temperature, pinch point temperature difference in the evaporator and degree of superheat at the ORC turbine inlet on the rate of produced hydrogen, sustainability index, overall exergy efficiency and organic Rankine cycle net produced power are investigated. It is observed that the rate of produced hydrogen and overall exergy efficiency of the proposed combined system take the maximum value to change in the evaporator temperature. Also, it is revealed that increasing the pinch point temperature difference in the evaporator decreases the rate of produced hydrogen and the overall exergy efficiency of the system. Furthermore, the effects on the rate of produced hydrogen and the overall exergy efficiency of the degree of superheat at the ORC turbine inlet are the same as the effect of pinch point temperature difference.

  16. Benchmarking for Higher Education.

    Science.gov (United States)

    Jackson, Norman, Ed.; Lund, Helen, Ed.

    The chapters in this collection explore the concept of benchmarking as it is being used and developed in higher education (HE). Case studies and reviews show how universities in the United Kingdom are using benchmarking to aid in self-regulation and self-improvement. The chapters are: (1) "Introduction to Benchmarking" (Norman Jackson…

  17. Lattice overview

    International Nuclear Information System (INIS)

    Creutz, M.

    1984-01-01

    After reviewing some recent developments in supercomputer access, the author discusses a few areas where perturbation theory and lattice gauge simulations make contact. The author concludes with a brief discussion of a deterministic dynamics for the Ising model. This may be useful for numerical studies of nonequilibrium phenomena. 13 references

  18. Benchmarking the internal combustion engine and hydrogen

    International Nuclear Information System (INIS)

    Wallace, J.S.

    2006-01-01

    The internal combustion engine is a cost-effective and highly reliable energy conversion technology. Exhaust emission regulations introduced in the 1970's triggered extensive research and development that has significantly improved in-use fuel efficiency and dramatically reduced exhaust emissions. The current level of gasoline vehicle engine development is highlighted and representative emissions and efficiency data are presented as benchmarks. The use of hydrogen fueling for IC engines has been investigated over many decades and the benefits and challenges arising are well-known. The current state of hydrogen-fueled engine development will be reviewed and evaluated against gasoline-fueled benchmarks. The prospects for further improvements to hydrogen-fueled IC engines will be examined. While fuel cells are projected to offer greater energy efficiency than IC engines and zero emissions, the availability of fuel cells in quantity at reasonable cost is a barrier to their widespread adaptation for the near future. In their current state of development, hydrogen fueled IC engines are an effective technology to create demand for hydrogen fueling infrastructure until fuel cells become available in commercial quantities. During this transition period, hydrogen fueled IC engines can achieve PZEV/ULSLEV emissions. (author)

  19. Benchmarking semantic web technology

    CERN Document Server

    García-Castro, R

    2009-01-01

    This book addresses the problem of benchmarking Semantic Web Technologies; first, from a methodological point of view, proposing a general methodology to follow in benchmarking activities over Semantic Web Technologies and, second, from a practical point of view, presenting two international benchmarking activities that involved benchmarking the interoperability of Semantic Web technologies using RDF(S) as the interchange language in one activity and OWL in the other.The book presents in detail how the different resources needed for these interoperability benchmarking activities were defined:

  20. Benchmarking in University Toolbox

    Directory of Open Access Journals (Sweden)

    Katarzyna Kuźmicz

    2015-06-01

    Full Text Available In the face of global competition and rising challenges that higher education institutions (HEIs meet, it is imperative to increase innovativeness and efficiency of their management. Benchmarking can be the appropriate tool to search for a point of reference necessary to assess institution’s competitive position and learn from the best in order to improve. The primary purpose of the paper is to present in-depth analysis of benchmarking application in HEIs worldwide. The study involves indicating premises of using benchmarking in HEIs. It also contains detailed examination of types, approaches and scope of benchmarking initiatives. The thorough insight of benchmarking applications enabled developing classification of benchmarking undertakings in HEIs. The paper includes review of the most recent benchmarking projects and relating them to the classification according to the elaborated criteria (geographical range, scope, type of data, subject, support and continuity. The presented examples were chosen in order to exemplify different approaches to benchmarking in higher education setting. The study was performed on the basis of the published reports from benchmarking projects, scientific literature and the experience of the author from the active participation in benchmarking projects. The paper concludes with recommendations for university managers undertaking benchmarking, derived on the basis of the conducted analysis.

  1. Analysis of an OECD/NEA high-temperature reactor benchmark

    International Nuclear Information System (INIS)

    Hosking, J. G.; Newton, T. D.; Koeberl, O.; Morris, P.; Goluoglu, S.; Tombakoglu, T.; Colak, U.; Sartori, E.

    2006-01-01

    This paper describes analyses of the OECD/NEA HTR benchmark organized by the 'Working Party on the Scientific Issues of Reactor Systems (WPRS)', formerly the 'Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles'. The benchmark was specifically designed to provide inter-comparisons for plutonium and thorium fuels when used in HTR systems. Calculations considering uranium fuel have also been included in the benchmark, in order to identify any increased uncertainties when using plutonium or thorium fuels. The benchmark consists of five phases, which include cell and whole-core calculations. Analysis of the benchmark has been performed by a number of international participants, who have used a range of deterministic and Monte Carlo code schemes. For each of the benchmark phases, neutronics parameters have been evaluated. Comparisons are made between the results of the benchmark participants, as well as comparisons between the predictions of the deterministic calculations and those from detailed Monte Carlo calculations. (authors)

  2. Lattice cell burnup calculation

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1977-01-01

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

  3. Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2000-01-01

    This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable

  4. Benchmarking criticality safety calculations with subcritical experiments

    International Nuclear Information System (INIS)

    Mihalczo, J.T.

    1984-06-01

    Calculation of the neutron multiplication factor at delayed criticality may be necessary for benchmarking calculations but it may not be sufficient. The use of subcritical experiments to benchmark criticality safety calculations could result in substantial savings in fuel material costs for experiments. In some cases subcritical configurations could be used to benchmark calculations where sufficient fuel to achieve delayed criticality is not available. By performing a variety of measurements with subcritical configurations, much detailed information can be obtained which can be compared directly with calculations. This paper discusses several measurements that can be performed with subcritical assemblies and presents examples that include comparisons between calculation and experiment where possible. Where not, examples from critical experiments have been used but the measurement methods could also be used for subcritical experiments

  5. Lattice QCD on fine lattices

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, Stefan [DESY (Germany). Neumann Inst. for Computing

    2016-11-01

    These configurations are currently in use in many on-going projects carried out by researchers throughout Europe. In particular this data will serve as an essential input into the computation of the coupling constant of QCD, where some of the simulations are still on-going. But also projects computing the masses of hadrons and investigating their structure are underway as well as activities in the physics of heavy quarks. As this initial project of gauge field generation has been successful, it is worthwhile to extend the currently available ensembles with further points in parameter space. These will allow to further study and control systematic effects like the ones introduced by the finite volume, the non-physical quark masses and the finite lattice spacing. In particular certain compromises have still been made in the region where pion masses and lattice spacing are both small. This is because physical pion masses require larger lattices to keep the effects of the finite volume under control. At light pion masses, a precise control of the continuum extrapolation is therefore difficult, but certainly a main goal of future simulations. To reach this goal, algorithmic developments as well as faster hardware will be needed.

  6. Calculus of a reactor VVER-1000 benchmark

    International Nuclear Information System (INIS)

    Dourougie, C.

    1998-01-01

    In the framework of the FMDP (Fissile Materials Disposition Program between the US and Russian, a benchmark was tested. The pin cells contain low enriched uranium (LEU) and mixed oxide fuels (MOX). The calculations are done for a wide range of temperatures and solute boron concentrations, in accidental conditions. (A.L.B.)

  7. MCNP neutron benchmarks

    International Nuclear Information System (INIS)

    Hendricks, J.S.; Whalen, D.J.; Cardon, D.A.; Uhle, J.L.

    1991-01-01

    Over 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The new and significant aspects of this work are as follows: These calculations are the first attempt at a validation program for MCNP and the first official benchmarking of version 4 of the code. We believe the chosen set of benchmarks is a comprehensive set that may be useful for benchmarking other radiation transport codes and data libraries. These calculations provide insight into how well neutron transport calculations can be expected to model a wide variety of problems

  8. RISKIND verification and benchmark comparisons

    International Nuclear Information System (INIS)

    Biwer, B.M.; Arnish, J.J.; Chen, S.Y.; Kamboj, S.

    1997-08-01

    This report presents verification calculations and benchmark comparisons for RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the population from exposures associated with the transportation of spent nuclear fuel and other radioactive materials. Spreadsheet calculations were performed to verify the proper operation of the major options and calculational steps in RISKIND. The program is unique in that it combines a variety of well-established models into a comprehensive treatment for assessing risks from the transportation of radioactive materials. Benchmark comparisons with other validated codes that incorporate similar models were also performed. For instance, the external gamma and neutron dose rate curves for a shipping package estimated by RISKIND were compared with those estimated by using the RADTRAN 4 code and NUREG-0170 methodology. Atmospheric dispersion of released material and dose estimates from the GENII and CAP88-PC codes. Verification results have shown the program to be performing its intended function correctly. The benchmark results indicate that the predictions made by RISKIND are within acceptable limits when compared with predictions from similar existing models

  9. RISKIND verification and benchmark comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Biwer, B.M.; Arnish, J.J.; Chen, S.Y.; Kamboj, S.

    1997-08-01

    This report presents verification calculations and benchmark comparisons for RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the population from exposures associated with the transportation of spent nuclear fuel and other radioactive materials. Spreadsheet calculations were performed to verify the proper operation of the major options and calculational steps in RISKIND. The program is unique in that it combines a variety of well-established models into a comprehensive treatment for assessing risks from the transportation of radioactive materials. Benchmark comparisons with other validated codes that incorporate similar models were also performed. For instance, the external gamma and neutron dose rate curves for a shipping package estimated by RISKIND were compared with those estimated by using the RADTRAN 4 code and NUREG-0170 methodology. Atmospheric dispersion of released material and dose estimates from the GENII and CAP88-PC codes. Verification results have shown the program to be performing its intended function correctly. The benchmark results indicate that the predictions made by RISKIND are within acceptable limits when compared with predictions from similar existing models.

  10. Assessment of the dry process fuel sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

  11. Assessment of the dry process fuel sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

  12. Determining the asymptotic buckling for the reference RB reactor lattice

    International Nuclear Information System (INIS)

    Martinc, R.; Sotic, O.

    1969-01-01

    Material buckling was measured for reference lattice of the heavy water reflected system with 2% enriched uranium fuel. Experiments were done for cores with lattice pitch values: 8, 8√2, i 16 cm. Each of these cores had heavy water reflector, as well as active reflector - heavy water lattice with natural uranium fuel. The core was reflected by natural uranium lattice in order to approach asymptotic regime in the central zone. Buckling values obtained with the natural uranium lattice as reflector are, as a rule, lower then in case of heavy water reflector [sr

  13. Development of a reference scheme for MOX lattice physics calculations

    International Nuclear Information System (INIS)

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

    1998-01-01

    The US program to dispose of weapons-grade Pu could involve the irradiation of mixed-oxide (MOX) fuel assemblies in commercial light water reactors. This will require licensing acceptance because of the modifications to the core safety characteristics. In particular, core neutronics will be significantly modified, thus making it necessary to validate the standard suites of neutronics codes for that particular application. Validation criteria are still unclear, but it seems reasonable to expect that the same level of accuracy will be expected for MOX as that which has been achieved for UO 2 . Commercial lattice physics codes are invariably claimed to be accurate for MOX analysis but often lack independent confirmation of their performance on a representative experimental database. Argonne National Laboratory (ANL) has started implementing a public domain suite of codes to provide for a capability to perform independent assessments of MOX core analyses. The DRAGON lattice code was chosen, and fine group ENDF/B-VI.04 and JEF-2.2 libraries have been developed. The objective of this work is to validate the DRAGON algorithms with respect to continuous-energy Monte Carlo for a suite of realistic UO 2 -MOX benchmark cases, with the aim of establishing a reference DRAGON scheme with a demonstrated high level of accuracy and no computing resource constraints. Using this scheme as a reference, future work will be devoted to obtaining simpler and less costly schemes that preserve accuracy as much as possible

  14. Efficient LBM visual simulation on face-centered cubic lattices.

    Science.gov (United States)

    Petkov, Kaloian; Qiu, Feng; Fan, Zhe; Kaufman, Arie E; Mueller, Klaus

    2009-01-01

    The Lattice Boltzmann method (LBM) for visual simulation of fluid flow generally employs cubic Cartesian (CC) lattices such as the D3Q13 and D3Q19 lattices for the particle transport. However, the CC lattices lead to suboptimal representation of the simulation space. We introduce the face-centered cubic (FCC) lattice, fD3Q13, for LBM simulations. Compared to the CC lattices, the fD3Q13 lattice creates a more isotropic sampling of the simulation domain and its single lattice speed (i.e., link length) simplifies the computations and data storage. Furthermore, the fD3Q13 lattice can be decomposed into two independent interleaved lattices, one of which can be discarded, which doubles the simulation speed. The resulting LBM simulation can be efficiently mapped to the GPU, further increasing the computational performance. We show the numerical advantages of the FCC lattice on channeled flow in 2D and the flow-past-a-sphere benchmark in 3D. In both cases, the comparison is against the corresponding CC lattices using the analytical solutions for the systems as well as velocity field visualizations. We also demonstrate the performance advantages of the fD3Q13 lattice for interactive simulation and rendering of hot smoke in an urban environment using thermal LBM.

  15. Benchmarking af kommunernes sagsbehandling

    DEFF Research Database (Denmark)

    Amilon, Anna

    Fra 2007 skal Ankestyrelsen gennemføre benchmarking af kommuernes sagsbehandlingskvalitet. Formålet med benchmarkingen er at udvikle praksisundersøgelsernes design med henblik på en bedre opfølgning og at forbedre kommunernes sagsbehandling. Dette arbejdspapir diskuterer metoder for benchmarking...

  16. Internet based benchmarking

    DEFF Research Database (Denmark)

    Bogetoft, Peter; Nielsen, Kurt

    2005-01-01

    We discuss the design of interactive, internet based benchmarking using parametric (statistical) as well as nonparametric (DEA) models. The user receives benchmarks and improvement potentials. The user is also given the possibility to search different efficiency frontiers and hereby to explore...

  17. The Drill Down Benchmark

    NARCIS (Netherlands)

    P.A. Boncz (Peter); T. Rühl (Tim); F. Kwakkel

    1998-01-01

    textabstractData Mining places specific requirements on DBMS query performance that cannot be evaluated satisfactorily using existing OLAP benchmarks. The DD Benchmark - defined here - provides a practical case and yardstick to explore how well a DBMS is able to support Data Mining applications. It

  18. Benchmarking Tool Kit.

    Science.gov (United States)

    Canadian Health Libraries Association.

    Nine Canadian health libraries participated in a pilot test of the Benchmarking Tool Kit between January and April, 1998. Although the Tool Kit was designed specifically for health libraries, the content and approach are useful to other types of libraries as well. Used to its full potential, benchmarking can provide a common measuring stick to…

  19. Verification of HELIOS-MASTER system through benchmark of critical experiments

    International Nuclear Information System (INIS)

    Kim, H. Y.; Kim, K. Y.; Cho, B. O.; Lee, C. C.; Zee, S. O.

    1999-01-01

    The HELIOS-MASTER code system is verified through the benchmark of the critical experiments that were performed by RRC 'Kurchatov Institute' with water-moderated hexagonally pitched lattices of highly enriched Uranium fuel rods (80w/o). We also used the same input by using the MCNP code that was described in the evaluation report, and compared our results with those of the evaluation report. HELIOS, developed by Scandpower A/S, is a two-dimensional transport program for the generation of group cross-sections, and MASTER, developed by KAERI, is a three-dimensional nuclear design and analysis code based on the two-group diffusion theory. It solves neutronics model with the AFEN (Analytic Function Expansion Nodal) method for hexagonal geometry. The results show that the HELIOS-MASTER code system is fast and accurate enough to be used as nuclear core analysis tool for hexagonal geometry

  20. Calculation of the 5th AER dynamic benchmark with APROS

    International Nuclear Information System (INIS)

    Puska, E.K.; Kontio, H.

    1998-01-01

    The model used for calculation of the 5th AER dynamic benchmark with APROS code is presented. In the calculation of the 5th AER dynamic benchmark the three-dimensional neutronics model of APROS was used. The core was divided axially into 20 nodes according to the specifications of the benchmark and each six identical fuel assemblies were placed into one one-dimensional thermal hydraulic channel. The five-equation thermal hydraulic model was used in the benchmark. The plant process and automation was described with a generic VVER-440 plant model created by IVO PE. (author)

  1. How Activists Use Benchmarks

    DEFF Research Database (Denmark)

    Seabrooke, Leonard; Wigan, Duncan

    2015-01-01

    Non-governmental organisations use benchmarks as a form of symbolic violence to place political pressure on firms, states, and international organisations. The development of benchmarks requires three elements: (1) salience, that the community of concern is aware of the issue and views...... are put to the test. The first is a reformist benchmarking cycle where organisations defer to experts to create a benchmark that conforms with the broader system of politico-economic norms. The second is a revolutionary benchmarking cycle driven by expert-activists that seek to contest strong vested...... interests and challenge established politico-economic norms. Differentiating these cycles provides insights into how activists work through organisations and with expert networks, as well as how campaigns on complex economic issues can be mounted and sustained....

  2. EGS4 benchmark program

    International Nuclear Information System (INIS)

    Yasu, Y.; Hirayama, H.; Namito, Y.; Yashiro, S.

    1995-01-01

    This paper proposes EGS4 Benchmark Suite which consists of three programs called UCSAMPL4, UCSAMPL4I and XYZDOS. This paper also evaluates optimization methods of recent RISC/UNIX systems, such as IBM, HP, DEC, Hitachi and Fujitsu, for the benchmark suite. When particular compiler option and math library were included in the evaluation process, system performed significantly better. Observed performance of some of the RISC/UNIX systems were beyond some so-called Mainframes of IBM, Hitachi or Fujitsu. The computer performance of EGS4 Code System on an HP9000/735 (99MHz) was defined to be the unit of EGS4 Unit. The EGS4 Benchmark Suite also run on various PCs such as Pentiums, i486 and DEC alpha and so forth. The performance of recent fast PCs reaches that of recent RISC/UNIX systems. The benchmark programs have been evaluated with correlation of industry benchmark programs, namely, SPECmark. (author)

  3. Benchmark calculation of nuclear design code for HCLWR

    International Nuclear Information System (INIS)

    Suzuki, Katsuo; Saji, Etsuro; Gakuhari, Kazuhiko; Akie, Hiroshi; Takano, Hideki; Ishiguro, Yukio.

    1986-01-01

    In the calculation of the lattice cell for High Conversion Light Water Reactors, big differences of nuclear design parameters appear between the results obtained by various methods and nuclear data libraries. The validity of the calculation can be verified by the critical experiment. The benchmark calculation is also efficient for the estimation of the validity in wide range of lattice parameters and burnup. As we do not have many measured data. The benchmark calculations were done by JAERI and MAPI, using SRAC and WIMS-E respectively. The problem covered the wide range of lattice parameters, i.e., from tight lattice to the current PWR lattice. The comparison was made on the effective multiplication factor, conversion ratio, and reaction rate of each nuclide, including burnup and void effects. The difference of the result is largest at the tightest lattice. But even at that lattice, the difference of the effective multiplication factor is only 1.4 %. The main cause of the difference is the neutron absorption rate U-238 in resonance energy region. The difference of other nuclear design parameters and their cause were also grasped. (author)

  4. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  5. Area of Lattice Polygons

    Science.gov (United States)

    Scott, Paul

    2006-01-01

    A lattice is a (rectangular) grid of points, usually pictured as occurring at the intersections of two orthogonal sets of parallel, equally spaced lines. Polygons that have lattice points as vertices are called lattice polygons. It is clear that lattice polygons come in various shapes and sizes. A very small lattice triangle may cover just 3…

  6. Verification and validation benchmarks.

    Energy Technology Data Exchange (ETDEWEB)

    Oberkampf, William Louis; Trucano, Timothy Guy

    2007-02-01

    Verification and validation (V&V) are the primary means to assess the accuracy and reliability of computational simulations. V&V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V&V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the level of

  7. Verification and validation benchmarks

    International Nuclear Information System (INIS)

    Oberkampf, William Louis; Trucano, Timothy Guy

    2007-01-01

    Verification and validation (V and V) are the primary means to assess the accuracy and reliability of computational simulations. V and V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V and V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the

  8. Verification and validation benchmarks

    International Nuclear Information System (INIS)

    Oberkampf, William L.; Trucano, Timothy G.

    2008-01-01

    Verification and validation (V and V) are the primary means to assess the accuracy and reliability of computational simulations. V and V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V and V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the

  9. Ad hoc committee on reactor physics benchmarks

    International Nuclear Information System (INIS)

    Diamond, D.J.; Mosteller, R.D.; Gehin, J.C.

    1996-01-01

    In the spring of 1994, an ad hoc committee on reactor physics benchmarks was formed under the leadership of two American Nuclear Society (ANS) organizations. The ANS-19 Standards Subcommittee of the Reactor Physics Division and the Computational Benchmark Problem Committee of the Mathematics and Computation Division had both seen a need for additional benchmarks to help validate computer codes used for light water reactor (LWR) neutronics calculations. Although individual organizations had employed various means to validate the reactor physics methods that they used for fuel management, operations, and safety, additional work in code development and refinement is under way, and to increase accuracy, there is a need for a corresponding increase in validation. Both organizations thought that there was a need to promulgate benchmarks based on measured data to supplement the LWR computational benchmarks that have been published in the past. By having an organized benchmark activity, the participants also gain by being able to discuss their problems and achievements with others traveling the same route

  10. Benchmarking of the 99-group ANSL-V library

    International Nuclear Information System (INIS)

    Wright, R.Q.; Ford, W.E. III; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Westfall, R.M.

    1987-01-01

    The purpose of this paper is to present thermal benchmark data testing results for the BAPL-1, TRX-1, and SEEP-1 lattices, using selected processed cross-sections from the ANSL-V 99-group library. 7 refs., 1 tab

  11. Benchmarking and the laboratory

    Science.gov (United States)

    Galloway, M; Nadin, L

    2001-01-01

    This article describes how benchmarking can be used to assess laboratory performance. Two benchmarking schemes are reviewed, the Clinical Benchmarking Company's Pathology Report and the College of American Pathologists' Q-Probes scheme. The Clinical Benchmarking Company's Pathology Report is undertaken by staff based in the clinical management unit, Keele University with appropriate input from the professional organisations within pathology. Five annual reports have now been completed. Each report is a detailed analysis of 10 areas of laboratory performance. In this review, particular attention is focused on the areas of quality, productivity, variation in clinical practice, skill mix, and working hours. The Q-Probes scheme is part of the College of American Pathologists programme in studies of quality assurance. The Q-Probes scheme and its applicability to pathology in the UK is illustrated by reviewing two recent Q-Probe studies: routine outpatient test turnaround time and outpatient test order accuracy. The Q-Probes scheme is somewhat limited by the small number of UK laboratories that have participated. In conclusion, as a result of the government's policy in the UK, benchmarking is here to stay. Benchmarking schemes described in this article are one way in which pathologists can demonstrate that they are providing a cost effective and high quality service. Key Words: benchmarking • pathology PMID:11477112

  12. Studies of thermal-reactor benchmark-data interpretation: experimental corrections

    International Nuclear Information System (INIS)

    Sher, R.; Fiarman, S.

    1976-10-01

    Experimental values of integral parameters of the lattices studied in this report, i.e., the MIT(D 2 O) and TRX benchmark lattices have been re-examined and revised. The revisions correct several systematic errors that have been previously ignored or considered insignificant. These systematic errors are discussed in detail. The final corrected values are presented

  13. ZZ ECN-BUBEBO, ECN-Petten Burnup Benchmark Book, Inventories, Afterheat

    International Nuclear Information System (INIS)

    Kloosterman, Jan Leen

    1999-01-01

    Description of program or function: Contains experimental benchmarks which can be used for the validation of burnup code systems and accompanied data libraries. Although the benchmarks presented here are thoroughly described in literature, it is in many cases not straightforward to retrieve unambiguously the correct input data and corresponding results from the benchmark Descriptions. Furthermore, results which can easily be measured, are sometimes difficult to calculate because of conversions to be made. Therefore, emphasis has been put to clarify the input of the benchmarks and to present the benchmark results in such a way that they can easily be calculated and compared. For more thorough Descriptions of the benchmarks themselves, the literature referred to here should be consulted. This benchmark book is divided in 11 chapters/files containing the following in text and tabular form: chapter 1: Introduction; chapter 2: Burnup Credit Criticality Benchmark Phase 1-B; chapter 3: Yankee-Rowe Core V Fuel Inventory Study; chapter 4: H.B. Robinson Unit 2 Fuel Inventory Study; chapter 5: Turkey Point Unit 3 Fuel Inventory Study; chapter 6: Turkey Point Unit 3 Afterheat Power Study; chapter 7: Dickens Benchmark on Fission Product Energy Release of U-235; chapter 8: Dickens Benchmark on Fission Product Energy Release of Pu-239; chapter 9: Yarnell Benchmark on Decay Heat Measurements of U-233; chapter 10: Yarnell Benchmark on Decay Heat Measurements of U-235; chapter 11: Yarnell Benchmark on Decay Heat Measurements of Pu-239

  14. Shielding benchmark problems, (2)

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Shin, Kazuo; Tada, Keiko.

    1980-02-01

    Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)

  15. Toxicological Benchmarks for Wildlife

    Energy Technology Data Exchange (ETDEWEB)

    Sample, B.E. Opresko, D.M. Suter, G.W.

    1993-01-01

    Ecological risks of environmental contaminants are evaluated by using a two-tiered process. In the first tier, a screening assessment is performed where concentrations of contaminants in the environment are compared to no observed adverse effects level (NOAEL)-based toxicological benchmarks. These benchmarks represent concentrations of chemicals (i.e., concentrations presumed to be nonhazardous to the biota) in environmental media (water, sediment, soil, food, etc.). While exceedance of these benchmarks does not indicate any particular level or type of risk, concentrations below the benchmarks should not result in significant effects. In practice, when contaminant concentrations in food or water resources are less than these toxicological benchmarks, the contaminants may be excluded from further consideration. However, if the concentration of a contaminant exceeds a benchmark, that contaminant should be retained as a contaminant of potential concern (COPC) and investigated further. The second tier in ecological risk assessment, the baseline ecological risk assessment, may use toxicological benchmarks as part of a weight-of-evidence approach (Suter 1993). Under this approach, based toxicological benchmarks are one of several lines of evidence used to support or refute the presence of ecological effects. Other sources of evidence include media toxicity tests, surveys of biota (abundance and diversity), measures of contaminant body burdens, and biomarkers. This report presents NOAEL- and lowest observed adverse effects level (LOAEL)-based toxicological benchmarks for assessment of effects of 85 chemicals on 9 representative mammalian wildlife species (short-tailed shrew, little brown bat, meadow vole, white-footed mouse, cottontail rabbit, mink, red fox, and whitetail deer) or 11 avian wildlife species (American robin, rough-winged swallow, American woodcock, wild turkey, belted kingfisher, great blue heron, barred owl, barn owl, Cooper's hawk, and red

  16. LATTICE: an interactive lattice computer code

    International Nuclear Information System (INIS)

    Staples, J.

    1976-10-01

    LATTICE is a computer code which enables an interactive user to calculate the functions of a synchrotron lattice. This program satisfies the requirements at LBL for a simple interactive lattice program by borrowing ideas from both TRANSPORT and SYNCH. A fitting routine is included

  17. Thermal reactor benchmark tests on JENDL-2

    International Nuclear Information System (INIS)

    Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

    1983-11-01

    A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

  18. Diagnostic Algorithm Benchmarking

    Science.gov (United States)

    Poll, Scott

    2011-01-01

    A poster for the NASA Aviation Safety Program Annual Technical Meeting. It describes empirical benchmarking on diagnostic algorithms using data from the ADAPT Electrical Power System testbed and a diagnostic software framework.

  19. Benchmarking Swiss electricity grids

    International Nuclear Information System (INIS)

    Walti, N.O.; Weber, Ch.

    2001-01-01

    This extensive article describes a pilot benchmarking project initiated by the Swiss Association of Electricity Enterprises that assessed 37 Swiss utilities. The data collected from these utilities on a voluntary basis included data on technical infrastructure, investments and operating costs. These various factors are listed and discussed in detail. The assessment methods and rating mechanisms that provided the benchmarks are discussed and the results of the pilot study are presented that are to form the basis of benchmarking procedures for the grid regulation authorities under the planned Switzerland's electricity market law. Examples of the practical use of the benchmarking methods are given and cost-efficiency questions still open in the area of investment and operating costs are listed. Prefaces by the Swiss Association of Electricity Enterprises and the Swiss Federal Office of Energy complete the article

  20. Benchmarking and Regulation

    DEFF Research Database (Denmark)

    Agrell, Per J.; Bogetoft, Peter

    . The application of benchmarking in regulation, however, requires specific steps in terms of data validation, model specification and outlier detection that are not systematically documented in open publications, leading to discussions about regulatory stability and economic feasibility of these techniques...

  1. Financial Integrity Benchmarks

    Data.gov (United States)

    City of Jackson, Mississippi — This data compiles standard financial integrity benchmarks that allow the City to measure its financial standing. It measure the City's debt ratio and bond ratings....

  2. Benchmarking in Foodservice Operations

    National Research Council Canada - National Science Library

    Johnson, Bonnie

    1998-01-01

    .... The design of this study included two parts: (1) eleven expert panelists involved in a Delphi technique to identify and rate importance of foodservice performance measures and rate the importance of benchmarking activities, and (2...

  3. ZZ WPPR, Pu Recycling Benchmark Results

    International Nuclear Information System (INIS)

    Lutz, D.; Mattes, M.; Delpech, Marc; Juanola, Marc

    2002-01-01

    Description of program or function: The NEA NSC Working Party on Physics of Plutonium Recycling has commissioned a series of benchmarks covering: - Plutonium recycling in pressurized-water reactors; - Void reactivity effect in pressurized-water reactors; - Fast Plutonium-burner reactors: beginning of life; - Plutonium recycling in fast reactors; - Multiple recycling in advanced pressurized-water reactors. The results have been published (see references). ZZ-WPPR-1-A/B contains graphs and tables relative to the PWR Mox pin cell benchmark, representing typical fuel for plutonium recycling, one corresponding to a first cycle, the second for a fifth cycle. These computer readable files contain the complete set of results, while the printed report contains only a subset. ZZ-WPPR-2-CYC1 are the results from cycle 1 of the multiple recycling benchmarks

  4. Monte Carlo analysis of experiments on the reactivity temperature coefficient for UO2 and MOX light water moderated lattices

    International Nuclear Information System (INIS)

    Chakir, E.; Erradi, L.; Bardouni, T El.; Khoukhi, T El.; Boukhal, H.; Meroun, O.; Bakkari, B El

    2007-01-01

    Full text: In a previous work, we have analysed the main french experiments available on the reactivity temperature coefficient (RTC) : CREAOLE and Mistral experiments. In these experiments, the RTC has been measured in both UO2 and UO2-PuO2 PWR type lattices. Our calculations, using APPOLO2 code with CEA93 library based on JEF2.2 evaluation, have shown that the calculation error in UO2 lattices is less than 1 pcm/Deg C which is considered as the target accuracy. On the other hand the calculation error in the MOX lattices is more significant in both low and high temperature ranges : an average error of -2 ± 0.5 pcm/Deg C is observed in low temperatures and an error of +3±2 pcm/Deg C is obtained for temperature higher than 250Deg C. In the present work, we analysed additional experimental benchmarks on the RTC of UO2 and MOX light water moderated lattices. To analyze these benchmarks and with the aim of minimizing uncertainties related to modelling of the experimental set up, we chose the Monte Carlo Method which has the advantage of taking into account in the most exact manner the geometry of the experimental configurations. Thus we have used the code MCNP5, for its recognized power and its availability. This analysis shows for the UO2 lattices, an average experiment-calculation deviation of about 0,5 pcm/Deg C, which is largely below the target accuracy for this type of lattices, that we estimate at approximately 1 pcm/Deg C. For the KAMINI experiment, which relates to the measurement of the RTC in light water moderated lattice using U-233 as fuel our analysis shows that the Endf/B6 library gives the best result, with an experiment -calculation deviation of the order of -0,16 pcm/Deg C. The analysis of the benchmarks using MOX fuel made it possible to highlight a discrepancy between experiment and calculation on the RTC of about -0.7pcm/Deg C ( for a range of temperature going from 20 to 248 Deg C) and -1.2 pcm/Deg C ( for a range of temperature going from 20 to

  5. MFTF TOTAL benchmark

    International Nuclear Information System (INIS)

    Choy, J.H.

    1979-06-01

    A benchmark of the TOTAL data base management system as applied to the Mirror Fusion Test Facility (MFTF) data base was implemented and run in February and March of 1979. The benchmark was run on an Interdata 8/32 and involved the following tasks: (1) data base design, (2) data base generation, (3) data base load, and (4) develop and implement programs to simulate MFTF usage of the data base

  6. Accelerator shielding benchmark problems

    International Nuclear Information System (INIS)

    Hirayama, H.; Ban, S.; Nakamura, T.

    1993-01-01

    Accelerator shielding benchmark problems prepared by Working Group of Accelerator Shielding in the Research Committee on Radiation Behavior in the Atomic Energy Society of Japan were compiled by Radiation Safety Control Center of National Laboratory for High Energy Physics. Twenty-five accelerator shielding benchmark problems are presented for evaluating the calculational algorithm, the accuracy of computer codes and the nuclear data used in codes. (author)

  7. Shielding benchmark problems

    International Nuclear Information System (INIS)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Kawai, Masayoshi; Nakazawa, Masaharu.

    1978-09-01

    Shielding benchmark problems were prepared by the Working Group of Assessment of Shielding Experiments in the Research Comittee on Shielding Design of the Atomic Energy Society of Japan, and compiled by the Shielding Laboratory of Japan Atomic Energy Research Institute. Twenty-one kinds of shielding benchmark problems are presented for evaluating the calculational algorithm and the accuracy of computer codes based on the discrete ordinates method and the Monte Carlo method and for evaluating the nuclear data used in the codes. (author)

  8. Calculus of a reactor VVER-1000 benchmark; Calcul d'un benchmark de reacteur VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Dourougie, C

    1998-07-01

    In the framework of the FMDP (Fissile Materials Disposition Program between the US and Russian, a benchmark was tested. The pin cells contain low enriched uranium (LEU) and mixed oxide fuels (MOX). The calculations are done for a wide range of temperatures and solute boron concentrations, in accidental conditions. (A.L.B.)

  9. Benchmarking electricity distribution

    Energy Technology Data Exchange (ETDEWEB)

    Watts, K. [Department of Justice and Attorney-General, QLD (Australia)

    1995-12-31

    Benchmarking has been described as a method of continuous improvement that involves an ongoing and systematic evaluation and incorporation of external products, services and processes recognised as representing best practice. It is a management tool similar to total quality management (TQM) and business process re-engineering (BPR), and is best used as part of a total package. This paper discusses benchmarking models and approaches and suggests a few key performance indicators that could be applied to benchmarking electricity distribution utilities. Some recent benchmarking studies are used as examples and briefly discussed. It is concluded that benchmarking is a strong tool to be added to the range of techniques that can be used by electricity distribution utilities and other organizations in search of continuous improvement, and that there is now a high level of interest in Australia. Benchmarking represents an opportunity for organizations to approach learning from others in a disciplined and highly productive way, which will complement the other micro-economic reforms being implemented in Australia. (author). 26 refs.

  10. Simulating colloid hydrodynamics with lattice Boltzmann methods

    International Nuclear Information System (INIS)

    Cates, M E; Stratford, K; Adhikari, R; Stansell, P; Desplat, J-C; Pagonabarraga, I; Wagner, A J

    2004-01-01

    We present a progress report on our work on lattice Boltzmann methods for colloidal suspensions. We focus on the treatment of colloidal particles in binary solvents and on the inclusion of thermal noise. For a benchmark problem of colloids sedimenting and becoming trapped by capillary forces at a horizontal interface between two fluids, we discuss the criteria for parameter selection, and address the inevitable compromise between computational resources and simulation accuracy

  11. Thermal and fast reactor benchmark testing of ENDF/B-6.4

    International Nuclear Information System (INIS)

    Liu Guisheng

    1999-01-01

    The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

  12. Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries; Definicion y Analisis de Benchmarks de Reactores de Agua Pesada para Pruebas de Nuevas Bibliotecas de Datos Wims-D

    Energy Technology Data Exchange (ETDEWEB)

    Leszczynski, Francisco [Comision Nacional de Energia Atomica, Centro Atomico Bariloche (Argentina)

    2000-07-01

    This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable.

  13. The KMAT: Benchmarking Knowledge Management.

    Science.gov (United States)

    de Jager, Martha

    Provides an overview of knowledge management and benchmarking, including the benefits and methods of benchmarking (e.g., competitive, cooperative, collaborative, and internal benchmarking). Arthur Andersen's KMAT (Knowledge Management Assessment Tool) is described. The KMAT is a collaborative benchmarking tool, designed to help organizations make…

  14. Lattice gauge theory

    International Nuclear Information System (INIS)

    Mack, G.

    1982-01-01

    After a description of a pure Yang-Mills theory on a lattice, the author considers a three-dimensional pure U(1) lattice gauge theory. Thereafter he discusses the exact relation between lattice gauge theories with the gauge groups SU(2) and SO(3). Finally he presents Monte Carlo data on phase transitions in SU(2) and SO(3) lattice gauge models. (HSI)

  15. Benchmarking the Netherlands. Benchmarking for growth

    International Nuclear Information System (INIS)

    2003-01-01

    This is the fourth edition of the Ministry of Economic Affairs' publication 'Benchmarking the Netherlands', which aims to assess the competitiveness of the Dutch economy. The methodology and objective of the benchmarking remain the same. The basic conditions for economic activity (institutions, regulation, etc.) in a number of benchmark countries are compared in order to learn from the solutions found by other countries for common economic problems. This publication is devoted entirely to the potential output of the Dutch economy. In other words, its ability to achieve sustainable growth and create work over a longer period without capacity becoming an obstacle. This is important because economic growth is needed to increase prosperity in the broad sense and meeting social needs. Prosperity in both a material (per capita GDP) and immaterial (living environment, environment, health, etc) sense, in other words. The economy's potential output is determined by two structural factors: the growth of potential employment and the structural increase in labour productivity. Analysis by the Netherlands Bureau for Economic Policy Analysis (CPB) shows that in recent years the increase in the capacity for economic growth has been realised mainly by increasing the supply of labour and reducing the equilibrium unemployment rate. In view of the ageing of the population in the coming years and decades the supply of labour is unlikely to continue growing at the pace we have become accustomed to in recent years. According to a number of recent studies, to achieve a respectable rate of sustainable economic growth the aim will therefore have to be to increase labour productivity. To realise this we have to focus on for six pillars of economic policy: (1) human capital, (2) functioning of markets, (3) entrepreneurship, (4) spatial planning, (5) innovation, and (6) sustainability. These six pillars determine the course for economic policy aiming at higher productivity growth. Throughout

  16. Benchmarking the Netherlands. Benchmarking for growth

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-01-01

    This is the fourth edition of the Ministry of Economic Affairs' publication 'Benchmarking the Netherlands', which aims to assess the competitiveness of the Dutch economy. The methodology and objective of the benchmarking remain the same. The basic conditions for economic activity (institutions, regulation, etc.) in a number of benchmark countries are compared in order to learn from the solutions found by other countries for common economic problems. This publication is devoted entirely to the potential output of the Dutch economy. In other words, its ability to achieve sustainable growth and create work over a longer period without capacity becoming an obstacle. This is important because economic growth is needed to increase prosperity in the broad sense and meeting social needs. Prosperity in both a material (per capita GDP) and immaterial (living environment, environment, health, etc) sense, in other words. The economy's potential output is determined by two structural factors: the growth of potential employment and the structural increase in labour productivity. Analysis by the Netherlands Bureau for Economic Policy Analysis (CPB) shows that in recent years the increase in the capacity for economic growth has been realised mainly by increasing the supply of labour and reducing the equilibrium unemployment rate. In view of the ageing of the population in the coming years and decades the supply of labour is unlikely to continue growing at the pace we have become accustomed to in recent years. According to a number of recent studies, to achieve a respectable rate of sustainable economic growth the aim will therefore have to be to increase labour productivity. To realise this we have to focus on for six pillars of economic policy: (1) human capital, (2) functioning of markets, (3) entrepreneurship, (4) spatial planning, (5) innovation, and (6) sustainability. These six pillars determine the course for economic policy aiming at higher productivity

  17. Measurements and analyses on reactivity effects of absorber rods in a light-water moderated UO2 lattices

    International Nuclear Information System (INIS)

    Murakami, Kiyonobu; Miyoshi, Yoshinori; Hirose, Hideyuki; Suzaki, Takenori

    1985-03-01

    Reactivity effects and reactivity-interference effects of absorber rods were measured with a cylindrical core aiming to obtain bench-marks for verification of the calculational methods. The core consisted of 2.6 w/o enriched UO 2 fuel rods lattice of which water-to-fuel volume ratio was 1.83. In the experiment, the critical water levels were measured changing neutron absorber content of absorber rods and the distance between two absorber rods in the core center. Monte Calro codes KENO-IV and MULTI-KENO were used to calculate reactivity worthes of absorber rods. The calculational results of effective multiplication factors ranged from 0.978 to 0.999 for the 60 cases of critical cores with inserted absorber rods. The calculational results of absorber worthes agreed to the experimental results within twice of the standerd deviation accompanied with the Monte Calro calculation. (author)

  18. Benchmarking in Mobarakeh Steel Company

    Directory of Open Access Journals (Sweden)

    Sasan Ghasemi

    2008-05-01

    Full Text Available Benchmarking is considered as one of the most effective ways of improving performance incompanies. Although benchmarking in business organizations is a relatively new concept and practice, ithas rapidly gained acceptance worldwide. This paper introduces the benchmarking project conducted in Esfahan’s Mobarakeh Steel Company, as the first systematic benchmarking project conducted in Iran. It aimsto share the process deployed for the benchmarking project in this company and illustrate how the projectsystematic implementation led to succes.

  19. Benchmarking in Mobarakeh Steel Company

    OpenAIRE

    Sasan Ghasemi; Mohammad Nazemi; Mehran Nejati

    2008-01-01

    Benchmarking is considered as one of the most effective ways of improving performance in companies. Although benchmarking in business organizations is a relatively new concept and practice, it has rapidly gained acceptance worldwide. This paper introduces the benchmarking project conducted in Esfahan's Mobarakeh Steel Company, as the first systematic benchmarking project conducted in Iran. It aims to share the process deployed for the benchmarking project in this company and illustrate how th...

  20. Lattices with unique complements

    CERN Document Server

    Saliĭ, V N

    1988-01-01

    The class of uniquely complemented lattices properly contains all Boolean lattices. However, no explicit example of a non-Boolean lattice of this class has been found. In addition, the question of whether this class contains any complete non-Boolean lattices remains unanswered. This book focuses on these classical problems of lattice theory and the various attempts to solve them. Requiring no specialized knowledge, the book is directed at researchers and students interested in general algebra and mathematical logic.

  1. Once-through thorium fuel cycle evaluation for TVA's Browns Ferry-3 Boiling Water Reactor

    International Nuclear Information System (INIS)

    Hopkins, G.C.

    1982-05-01

    This report documents benchmark evaluations to test thorium lattice predictive methods and neutron cross sections against available data and summarizes specific evaluations of the once-through thorium cycle when applied to the Browns Ferry-3 BWR. It was concluded that appreciable uncertainties in thorium cycle nuclear data cloud the ability to reliably predict the fuel cycle performance and that power reactor irradiations of ThO 2 rods in BWRs are desirable to resolve uncertainties. Benchmark evaluations indicated that the ENDF/B-IV data used in the evaluations should cause an underprediction of U-233/ThO 2 fuel reactivity, and, therefore, the results of the preliminary evaluations completed under the program should be conservative

  2. Shielding Benchmark Computational Analysis

    International Nuclear Information System (INIS)

    Hunter, H.T.; Slater, C.O.; Holland, L.B.; Tracz, G.; Marshall, W.J.; Parsons, J.L.

    2000-01-01

    Over the past several decades, nuclear science has relied on experimental research to verify and validate information about shielding nuclear radiation for a variety of applications. These benchmarks are compared with results from computer code models and are useful for the development of more accurate cross-section libraries, computer code development of radiation transport modeling, and building accurate tests for miniature shielding mockups of new nuclear facilities. When documenting measurements, one must describe many parts of the experimental results to allow a complete computational analysis. Both old and new benchmark experiments, by any definition, must provide a sound basis for modeling more complex geometries required for quality assurance and cost savings in nuclear project development. Benchmarks may involve one or many materials and thicknesses, types of sources, and measurement techniques. In this paper the benchmark experiments of varying complexity are chosen to study the transport properties of some popular materials and thicknesses. These were analyzed using three-dimensional (3-D) models and continuous energy libraries of MCNP4B2, a Monte Carlo code developed at Los Alamos National Laboratory, New Mexico. A shielding benchmark library provided the experimental data and allowed a wide range of choices for source, geometry, and measurement data. The experimental data had often been used in previous analyses by reputable groups such as the Cross Section Evaluation Working Group (CSEWG) and the Organization for Economic Cooperation and Development/Nuclear Energy Agency Nuclear Science Committee (OECD/NEANSC)

  3. Different Activation Techniques for the Study of Epithermal Spectra, Applied to Heavy Water Lattices of Varying Fuel-To-Moderator Ratio

    Energy Technology Data Exchange (ETDEWEB)

    Sokolowski, E K

    1966-06-15

    Spectral indices at the cell boundary have been studied as functions of lattice pitch in the reference core of the Swedish R0 reactor. Epithermal indices were determined by activation of In{sup 115}, employing three different techniques: the two-foil, the cadmium ratio and the sandwich foil methods. The latter of these has the advantage of being independent of assumptions about foil cross sections or spectral functions, and it gives a spectrum index that lends itself readily to comparisons with theoretical multigroup calculations. Alternatively the results can be expressed in terms of the Westcott parameters r and T{sub n} when this is justified by the spectral conditions. The agreement between the three methods investigated is generally good. Good agreement is also found with multigroup collision.

  4. Benchmarking of the PHOENIX-P/ANC [Advanced Nodal Code] advanced nuclear design system

    International Nuclear Information System (INIS)

    Nguyen, T.Q.; Liu, Y.S.; Durston, C.; Casadei, A.L.

    1988-01-01

    At Westinghouse, an advanced neutronic methods program was designed to improve the quality of the predictions, enhance flexibility in designing advanced fuel and related products, and improve design lead time. Extensive benchmarking data is presented to demonstrate the accuracy of the Advanced Nodal Code (ANC) and the PHOENIX-P advanced lattice code. Qualification data to demonstrate the accuracy of ANC include comparison of key physics parameters against a fine-mesh diffusion theory code, TORTISE. Benchmarking data to demonstrate the validity of the PHOENIX-P methodologies include comparison of physics predictions against critical experiments, isotopics measurements and measured power distributions from spatial criticals. The accuracy of the PHOENIX-P/ANC Advanced Design System is demonstrated by comparing predictions of hot zero power physics parameters and hot full power core follow against measured data from operating reactors. The excellent performance of this system for a broad range of comparisons establishes the basis for implementation of these tools for core design, licensing and operational follow of PWR [pressurized water reactor] cores at Westinghouse

  5. Deviating From the Benchmarks

    DEFF Research Database (Denmark)

    Rocha, Vera; Van Praag, Mirjam; Carneiro, Anabela

    This paper studies three related questions: To what extent otherwise similar startups employ different quantities and qualities of human capital at the moment of entry? How persistent are initial human capital choices over time? And how does deviating from human capital benchmarks influence firm......, founders human capital, and the ownership structure of startups (solo entrepreneurs versus entrepreneurial teams). We then study the survival implications of exogenous deviations from these benchmarks, based on spline models for survival data. Our results indicate that (especially negative) deviations from...... the benchmark can be substantial, are persistent over time, and hinder the survival of firms. The implications may, however, vary according to the sector and the ownership structure at entry. Given the stickiness of initial choices, wrong human capital decisions at entry turn out to be a close to irreversible...

  6. HPCG Benchmark Technical Specification

    Energy Technology Data Exchange (ETDEWEB)

    Heroux, Michael Allen [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Dongarra, Jack [Univ. of Tennessee, Knoxville, TN (United States); Luszczek, Piotr [Univ. of Tennessee, Knoxville, TN (United States)

    2013-10-01

    The High Performance Conjugate Gradient (HPCG) benchmark [cite SNL, UTK reports] is a tool for ranking computer systems based on a simple additive Schwarz, symmetric Gauss-Seidel preconditioned conjugate gradient solver. HPCG is similar to the High Performance Linpack (HPL), or Top 500, benchmark [1] in its purpose, but HPCG is intended to better represent how today’s applications perform. In this paper we describe the technical details of HPCG: how it is designed and implemented, what code transformations are permitted and how to interpret and report results.

  7. Benchmarking for Best Practice

    CERN Document Server

    Zairi, Mohamed

    1998-01-01

    Benchmarking for Best Practice uses up-to-the-minute case-studies of individual companies and industry-wide quality schemes to show how and why implementation has succeeded. For any practitioner wanting to establish best practice in a wide variety of business areas, this book makes essential reading. .It is also an ideal textbook on the applications of TQM since it describes concepts, covers definitions and illustrates the applications with first-hand examples. Professor Mohamed Zairi is an international expert and leading figure in the field of benchmarking. His pioneering work in this area l

  8. Benchmarking Danish Industries

    DEFF Research Database (Denmark)

    Gammelgaard, Britta; Bentzen, Eric; Aagaard Andreassen, Mette

    2003-01-01

    compatible survey. The International Manufacturing Strategy Survey (IMSS) doesbring up the question of supply chain management, but unfortunately, we did not have access to thedatabase. Data from the members of the SCOR-model, in the form of benchmarked performance data,may exist, but are nonetheless...... not public. The survey is a cooperative project "Benchmarking DanishIndustries" with CIP/Aalborg University, the Danish Technological University, the DanishTechnological Institute and Copenhagen Business School as consortia partners. The project has beenfunded by the Danish Agency for Trade and Industry...

  9. [Do you mean benchmarking?].

    Science.gov (United States)

    Bonnet, F; Solignac, S; Marty, J

    2008-03-01

    The purpose of benchmarking is to settle improvement processes by comparing the activities to quality standards. The proposed methodology is illustrated by benchmark business cases performed inside medical plants on some items like nosocomial diseases or organization of surgery facilities. Moreover, the authors have built a specific graphic tool, enhanced with balance score numbers and mappings, so that the comparison between different anesthesia-reanimation services, which are willing to start an improvement program, is easy and relevant. This ready-made application is even more accurate as far as detailed tariffs of activities are implemented.

  10. RB reactor benchmark cores

    International Nuclear Information System (INIS)

    Pesic, M.

    1998-01-01

    A selected set of the RB reactor benchmark cores is presented in this paper. The first results of validation of the well-known Monte Carlo MCNP TM code and adjoining neutron cross section libraries are given. They confirm the idea for the proposal of the new U-D 2 O criticality benchmark system and support the intention to include this system in the next edition of the recent OECD/NEA Project: International Handbook of Evaluated Criticality Safety Experiment, in near future. (author)

  11. Integral nuclear fuel element assembly

    International Nuclear Information System (INIS)

    Schluderberg, D. C.

    1985-01-01

    An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

  12. Status on benchmark testing of CENDL-3

    CERN Document Server

    Liu Ping

    2002-01-01

    CENDL-3, the newest version of China Evaluated Nuclear Data Library has been finished, and distributed for some benchmarks analysis recently. The processing was carried out using the NJOY nuclear data processing code system. The calculations and analysis of benchmarks were done with Monte Carlo code MCNP and reactor lattice code WIMSD5A. The calculated results were compared with the experimental results based on ENDF/B6. In most thermal and fast uranium criticality benchmarks, the calculated k sub e sub f sub f values with CENDL-3 were in good agreements with experimental results. In the plutonium fast cores, the k sub e sub f sub f values were improved significantly with CENDL-3. This is duo to reevaluation of the fission spectrum and elastic angular distributions of sup 2 sup 3 sup 9 Pu and sup 2 sup 4 sup 0 Pu. CENDL-3 underestimated the k sub e sub f sub f values compared with other evaluated data libraries for most spherical or cylindrical assemblies of plutonium or uranium with beryllium

  13. Calculational methods for lattice cells

    International Nuclear Information System (INIS)

    Askew, J.R.

    1980-01-01

    At the current stage of development, direct simulation of all the processes involved in the reactor to the degree of accuracy required is not an economic proposition, and this is achieved by progressive synthesis of models for parts of the full space/angle/energy neutron behaviour. The split between reactor and lattice calculations is one such simplification. Most reactors are constructed of repetitions of similar geometric units, the fuel elements, having broadly similar properties. Thus the provision of detailed predictions of their behaviour is an important step towards overall modelling. We shall be dealing with these lattice methods in this series of lectures, but will refer back from time to time to their relationship with overall reactor calculation The lattice cell is itself composed of somewhat similar sub-units, the fuel pins, and will itself often rely upon a further break down of modelling. Construction of a good model depends upon the identification, on physical and mathematical grounds, of the most helpful division of the calculation at this level

  14. Benchmarking and Performance Management

    Directory of Open Access Journals (Sweden)

    Adrian TANTAU

    2010-12-01

    Full Text Available The relevance of the chosen topic is explained by the meaning of the firm efficiency concept - the firm efficiency means the revealed performance (how well the firm performs in the actual market environment given the basic characteristics of the firms and their markets that are expected to drive their profitability (firm size, market power etc.. This complex and relative performance could be due to such things as product innovation, management quality, work organization, some other factors can be a cause even if they are not directly observed by the researcher. The critical need for the management individuals/group to continuously improve their firm/company’s efficiency and effectiveness, the need for the managers to know which are the success factors and the competitiveness determinants determine consequently, what performance measures are most critical in determining their firm’s overall success. Benchmarking, when done properly, can accurately identify both successful companies and the underlying reasons for their success. Innovation and benchmarking firm level performance are critical interdependent activities. Firm level variables, used to infer performance, are often interdependent due to operational reasons. Hence, the managers need to take the dependencies among these variables into account when forecasting and benchmarking performance. This paper studies firm level performance using financial ratio and other type of profitability measures. It uses econometric models to describe and then propose a method to forecast and benchmark performance.

  15. Surveys and Benchmarks

    Science.gov (United States)

    Bers, Trudy

    2012-01-01

    Surveys and benchmarks continue to grow in importance for community colleges in response to several factors. One is the press for accountability, that is, for colleges to report the outcomes of their programs and services to demonstrate their quality and prudent use of resources, primarily to external constituents and governing boards at the state…

  16. KAERI results for BN600 full MOX benchmark (Phase 4)

    International Nuclear Information System (INIS)

    Lee, Kibog Lee

    2003-01-01

    The purpose of this document is to report the results of KAERI's calculation for the Phase-4 of BN-600 full MOX fueled core benchmark analyses according to the RCM report of IAEA CRP Action on U pdated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. T he BN-600 full MOX core model is based on the specification in the document, F ull MOX Model (Phase4. doc ) . This document addresses the calculational methods employed in the benchmark analyses and benchmark results carried out by KAERI

  17. Epithermal and Thermal Spectrum Indices in Heavy Water Lattices

    Energy Technology Data Exchange (ETDEWEB)

    Sokolowski, E K; Jonsson, A

    1967-05-15

    Spectral indices have been measured by foil activation technique in a number of different D{sub 2}O-moderated lattices in the Swedish zero power reactor R0 and the pressurized exponential assembly TZ. In most cases the fuel was in the form of single rods, distributed uniformly in the lattice. Parameters in these cases were lattice pitch and fuel composition. A 31-rod cluster lattice was also investigated, with the moderator temperature varying up to 210 deg C. On the basis of these measurements, as well as measurements on cluster lattices, reported by other investigators, it has been possible to derive simple correlations for the spectral indices, which seem to be of fairly general validity for D{sub 2}O lattices. The experimental results have also been compared to calculations with the multigroup collision probability program FLEF.

  18. Epithermal and Thermal Spectrum Indices in Heavy Water Lattices

    International Nuclear Information System (INIS)

    Sokolowski, E.K.; Jonsson, A.

    1967-05-01

    Spectral indices have been measured by foil activation technique in a number of different D 2 O-moderated lattices in the Swedish zero power reactor R0 and the pressurized exponential assembly TZ. In most cases the fuel was in the form of single rods, distributed uniformly in the lattice. Parameters in these cases were lattice pitch and fuel composition. A 31-rod cluster lattice was also investigated, with the moderator temperature varying up to 210 deg C. On the basis of these measurements, as well as measurements on cluster lattices, reported by other investigators, it has been possible to derive simple correlations for the spectral indices, which seem to be of fairly general validity for D 2 O lattices. The experimental results have also been compared to calculations with the multigroup collision probability program FLEF

  19. New integrable lattice hierarchies

    International Nuclear Information System (INIS)

    Pickering, Andrew; Zhu Zuonong

    2006-01-01

    In this Letter we give a new integrable four-field lattice hierarchy, associated to a new discrete spectral problem. We obtain our hierarchy as the compatibility condition of this spectral problem and an associated equation, constructed herein, for the time-evolution of eigenfunctions. We consider reductions of our hierarchy, which also of course admit discrete zero curvature representations, in detail. We find that our hierarchy includes many well-known integrable hierarchies as special cases, including the Toda lattice hierarchy, the modified Toda lattice hierarchy, the relativistic Toda lattice hierarchy, and the Volterra lattice hierarchy. We also obtain here a new integrable two-field lattice hierarchy, to which we give the name of Suris lattice hierarchy, since the first equation of this hierarchy has previously been given by Suris. The Hamiltonian structure of the Suris lattice hierarchy is obtained by means of a trace identity formula

  20. Dark matter, constrained minimal supersymmetric standard model, and lattice QCD.

    Science.gov (United States)

    Giedt, Joel; Thomas, Anthony W; Young, Ross D

    2009-11-13

    Recent lattice measurements have given accurate estimates of the quark condensates in the proton. We use these results to significantly improve the dark matter predictions in benchmark models within the constrained minimal supersymmetric standard model. The predicted spin-independent cross sections are at least an order of magnitude smaller than previously suggested and our results have significant consequences for dark matter searches.

  1. Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

    International Nuclear Information System (INIS)

    Bess, John D.; Montierth, Leland; Köberl, Oliver

    2014-01-01

    Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the 235 U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

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

    International Nuclear Information System (INIS)

    Mertyurek, Ugur; Gauld, Ian C.

    2016-01-01

    Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  3. Precise determination of lattice phase shifts and mixing angles

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Bing-Nan, E-mail: b.lu@fz-juelich.de [Institute for Advanced Simulation, Institut für Kernphysik, and Jülich Center for Hadron Physics, Forschungszentrum Jülich, D-52425 Jülich (Germany); Lähde, Timo A. [Institute for Advanced Simulation, Institut für Kernphysik, and Jülich Center for Hadron Physics, Forschungszentrum Jülich, D-52425 Jülich (Germany); Lee, Dean [Department of Physics, North Carolina State University, Raleigh, NC 27695 (United States); Meißner, Ulf-G. [Helmholtz-Institut für Strahlen- und Kernphysik and Bethe Center for Theoretical Physics, Universität Bonn, D-53115 Bonn (Germany); Institute for Advanced Simulation, Institut für Kernphysik, and Jülich Center for Hadron Physics, Forschungszentrum Jülich, D-52425 Jülich (Germany); JARA – High Performance Computing, Forschungszentrum Jülich, D-52425 Jülich (Germany)

    2016-09-10

    We introduce a general and accurate method for determining lattice phase shifts and mixing angles, which is applicable to arbitrary, non-cubic lattices. Our method combines angular momentum projection, spherical wall boundaries and an adjustable auxiliary potential. This allows us to construct radial lattice wave functions and to determine phase shifts at arbitrary energies. For coupled partial waves, we use a complex-valued auxiliary potential that breaks time-reversal invariance. We benchmark our method using a system of two spin-1/2 particles interacting through a finite-range potential with a strong tensor component. We are able to extract phase shifts and mixing angles for all angular momenta and energies, with precision greater than that of extant methods. We discuss a wide range of applications from nuclear lattice simulations to optical lattice experiments.

  4. Impact of the thermal scattering law of H in H_2O on the isothermal temperatures reactivity coefficients for UOX and MOX fuel lattices in cold operating conditions

    International Nuclear Information System (INIS)

    Scotta, J.P.; Noguere, G.; Bernard, D.; Santamarina, A.; Damian, J.I.M.

    2016-01-01

    The contribution of the thermal scattering law of hydrogen in light water to isothermal temperature reactivity coefficients for UOX and MOX lattices was studied in the frame of the MISTRAL critical experiments carried out in the zero power reactor EOLE of CEA Cadarache (France). The interpretation of the core residual reactivity measured between 6 to 80 C. degrees (by step of 5 C. degrees) was performed with the Monte-Carlo code TRIPOLI-4"R. The nuclear data from the JEFF-3.1.1 library were used in the calculations. 3 different thermal scattering laws of hydrogen in light water were tested in order to evaluate their impact on the MISTRAL calculations. The thermal scattering laws of interest were firstly those recommended in JEFF-3.1.1 and ENDF/BVII.1 and also that recently produced at the atomic center of Bariloche (CAB, Argentina) with molecular dynamic simulations. The present work indicates that the calculation-to-experimental bias is (0.4 ± 0.3) pcm/C. degree in the UOX core and (1.0 ± 0.3) pcm/C. degree in the MOX cores, when the JEFF-3.1.1 library is used. An improvement is observed over the whole temperature range with the CAB model. The calculation-to-experimental bias vanishes for the UOX core (0.02 pcm/C. degree) and becomes close to 0.7 pcm/C. degree for the MOX cores. The magnitude of these bias have to be connected to the typical value of the temperature reactivity coefficient that ranges from 5 pcm/C. degree at Beginning Of Cycle (BOC) up to 50 pcm/C. degrees at End Of Cycle (EOC), in PWR conditions. (authors)

  5. Towards the reanalysis of void coefficients measurements at proteus for high conversion light water reactor lattices

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Koeberl, O.; Perret, G. [Paul Scherrer Institut PSI, 5232 Villigen (Switzerland)

    2012-07-01

    High Conversion Light Water Reactors (HCLWR) allows a better usage of fuel resources thanks to a higher breeding ratio than standard LWR. Their uses together with the current fleet of LWR constitute a fuel cycle thoroughly studied in Japan and the US today. However, one of the issues related to HCLWR is their void reactivity coefficient (VRC), which can be positive. Accurate predictions of void reactivity coefficient in HCLWR conditions and their comparisons with representative experiments are therefore required. In this paper an inter comparison of modern codes and cross-section libraries is performed for a former Benchmark on Void Reactivity Effect in PWRs conducted by the OECD/NEA. It shows an overview of the k-inf values and their associated VRC obtained for infinite lattice calculations with UO{sub 2} and highly enriched MOX fuel cells. The codes MCNPX2.5, TRIPOLI4.4 and CASMO-5 in conjunction with the libraries ENDF/B-VI.8, -VII.0, JEF-2.2 and JEFF-3.1 are used. A non-negligible spread of results for voided conditions is found for the high content MOX fuel. The spread of eigenvalues for the moderated and voided UO{sub 2} fuel are about 200 pcm and 700 pcm, respectively. The standard deviation for the VRCs for the UO{sub 2} fuel is about 0.7% while the one for the MOX fuel is about 13%. This work shows that an appropriate treatment of the unresolved resonance energy range is an important issue for the accurate determination of the void reactivity effect for HCLWR. A comparison to experimental results is needed to resolve the presented discrepancies. (authors)

  6. Benchmarking i den offentlige sektor

    DEFF Research Database (Denmark)

    Bukh, Per Nikolaj; Dietrichson, Lars; Sandalgaard, Niels

    2008-01-01

    I artiklen vil vi kort diskutere behovet for benchmarking i fraværet af traditionelle markedsmekanismer. Herefter vil vi nærmere redegøre for, hvad benchmarking er med udgangspunkt i fire forskellige anvendelser af benchmarking. Regulering af forsyningsvirksomheder vil blive behandlet, hvorefter...

  7. Burn-up TRIGA Mark II benchmark experiment

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  8. Cloud benchmarking for performance

    OpenAIRE

    Varghese, Blesson; Akgun, Ozgur; Miguel, Ian; Thai, Long; Barker, Adam

    2014-01-01

    Date of Acceptance: 20/09/2014 How can applications be deployed on the cloud to achieve maximum performance? This question has become significant and challenging with the availability of a wide variety of Virtual Machines (VMs) with different performance capabilities in the cloud. The above question is addressed by proposing a six step benchmarking methodology in which a user provides a set of four weights that indicate how important each of the following groups: memory, processor, computa...

  9. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sasaki, Y.; Tashima, J.

    1975-01-01

    A description is given of nuclear reactor fuel assemblies arranged in the form of a lattice wherein there is attached to the interface of one of two adjacent fuel assemblies a plate spring having a concave portion curved toward said interface and to the interface of the other fuel assembly a plate spring having a convex portion curved away from said interface

  10. Coolant Void Reactivity Analysis of CANDU Lattice

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    Models of CANDU-6 and ACR-700 fuel lattices were constructed for a single bundle and 2 by 2 checkerboard to understand the physics related to CVR. Also, a familiar four factor formula was used to predict the specific contributions to reactivity change in order to achieve an understanding of the physics issues related to the CVR. At the same time, because the situation of coolant voiding should bring about a change of neutron behavior, the spectral changes and neutron current were also analyzed. The models of the CANDU- 6 and ACR-700 fuel lattices were constructed using the Monte Carlo code MCNP6 using the ENDF/B-VII.0 continuous energy cross section library based on the specification from AECL. The CANDU fuel lattice was searched through sensitivity studies of each design parameter such as fuel enrichment, fuel pitch, and types of burnable absorber for obtaining better behavior in terms of CVR. Unlike the single channel coolant voiding, the ACR-700 bundle has a positive reactivity change upon 2x2 checkerboard coolant voiding. Because of the new path for neutron moderation, the neutrons from the voided channel move to the no-void channel where they lose energy and come back to the voided channel as thermal neutrons. This phenomenon causes the positive CVR when checkerboard voiding occurs. The sensitivity study revealed the effects of the moderator to fuel volume ratio, fuel enrichment, and burnable absorber on the CVR. A fuel bundle with low moderator to fuel volume ratio and high fuel enrichment can help achieve negative CVR.

  11. KENO-IV code benchmark calculation, (6)

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Naito, Yoshitaka; Yamakawa, Yasuhiro.

    1980-11-01

    A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multigroup constants library MGCL. The present report describes the results of a benchmark test using criticality experiments about Plutonium fuel in various shape. In all, 33 cases of experiments have been calculated for Pu(NO 3 ) 4 aqueous solution, Pu metal or PuO 2 -polystyrene compact in various shape (sphere, cylinder, rectangular parallelepiped). The effective multiplication factors calculated for the 33 cases distribute widely between 0.955 and 1.045 due to wide range of system variables. (author)

  12. Generalized isothermic lattices

    International Nuclear Information System (INIS)

    Doliwa, Adam

    2007-01-01

    We study multi-dimensional quadrilateral lattices satisfying simultaneously two integrable constraints: a quadratic constraint and the projective Moutard constraint. When the lattice is two dimensional and the quadric under consideration is the Moebius sphere one obtains, after the stereographic projection, the discrete isothermic surfaces defined by Bobenko and Pinkall by an algebraic constraint imposed on the (complex) cross-ratio of the circular lattice. We derive the analogous condition for our generalized isothermic lattices using Steiner's projective structure of conics, and we present basic geometric constructions which encode integrability of the lattice. In particular, we introduce the Darboux transformation of the generalized isothermic lattice and we derive the corresponding Bianchi permutability principle. Finally, we study two-dimensional generalized isothermic lattices, in particular geometry of their initial boundary value problem

  13. Benchmarking reference services: an introduction.

    Science.gov (United States)

    Marshall, J G; Buchanan, H S

    1995-01-01

    Benchmarking is based on the common sense idea that someone else, either inside or outside of libraries, has found a better way of doing certain things and that your own library's performance can be improved by finding out how others do things and adopting the best practices you find. Benchmarking is one of the tools used for achieving continuous improvement in Total Quality Management (TQM) programs. Although benchmarking can be done on an informal basis, TQM puts considerable emphasis on formal data collection and performance measurement. Used to its full potential, benchmarking can provide a common measuring stick to evaluate process performance. This article introduces the general concept of benchmarking, linking it whenever possible to reference services in health sciences libraries. Data collection instruments that have potential application in benchmarking studies are discussed and the need to develop common measurement tools to facilitate benchmarking is emphasized.

  14. Fuel assembly

    International Nuclear Information System (INIS)

    Bando, Masaru.

    1993-01-01

    As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

  15. Fast burner reactor benchmark results from the NEA working party on physics of plutonium recycle

    International Nuclear Information System (INIS)

    Hill, R.N.; Wade, D.C.; Palmiotti, G.

    1995-01-01

    As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed; fuel cycle scenarios using either PUREX/TRUEX (oxide fuel) or pyrometallurgical (metal fuel) separation technologies were specified. These benchmarks were designed to evaluate the nuclear performance and radiotoxicity impact of a transuranic-burning fast reactor system. International benchmark results are summarized in this paper; and key conclusions are highlighted

  16. Analysis on void reactivity of DCA lattice

    International Nuclear Information System (INIS)

    Min, B. J.; Noh, K. H.; Choi, H. B.; Yang, M. K.

    2001-01-01

    In case of loss of coolant accident, the void reactivity of CANDU fuel provides the positive reactivity and increases the reactor power rapidly. Therefore, it is required to secure credibility of the void reactivity for the design and analysis of reactor, which motivated a study to assess the measurement data of void reactivity. The assessment of lattice code was performed with the experimental data of void reactivity at 30, 70, 87 and 100% of void fractions. The infinite multiplication factors increased in four types of fuels as the void fractions of them grow. The infinite multiplication factors of uranium fuels are almost within 1%, but those of Pu fuels are over 10% by the results of WIMS-AECL and MCNP-4B codes. Moreover, coolant void reactivity of the core loaded with plutonium fuel is more negative compared with that with uranium fuel because of spectrum hardening resulting from large void fraction

  17. ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    2017-06-19

    The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared to a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output

  18. Partially-reflected water-moderated square-piteched U(6.90)O2 fuel rod lattices with 0.67 fuel to water volume ratio (0.800 CM Pitch)

    Energy Technology Data Exchange (ETDEWEB)

    Harms, Gary A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5% 235U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90%)O2 fuel rods.

  19. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Wakamatsu, Mitsuo.

    1974-01-01

    Object: To improve a circulating flow passage of coolant so as to be able to accurately detect the temperature of coolant, rare gases contained, and the like. Structure: A fuel assembly comprising a flow regulating lattice provided with a plurality of communication holes in an axial direction, said lattice being positioned at the upper end of an outer tube in which nuclear fuel elements are received, and a neutron shielding body having a plurality of spiral coolant flow passages disposed between the lattice and the nuclear fuel elements, whereby a coolant comprised of liquid sodium or the like, which moves up passing through the coolant flow passages and the flow regulating passage, is regulated and passed through a detector mounted at the upper part of the flow regulating lattice to detect coolant temperature, flow rate, and rare gases or the like as the origin of nuclear fission contained in the coolant due to breakage of fuel elements. (Kamimura, M.)

  20. MCNP simulation of the TRIGA Mark II benchmark experiment

    International Nuclear Information System (INIS)

    Jeraj, R.; Glumac, B.; Maucec, M.

    1996-01-01

    The complete 3D MCNP model of the TRIGA Mark II reactor is presented. It enables precise calculations of some quantities of interest in a steady-state mode of operation. Calculational results are compared to the experimental results gathered during reactor reconstruction in 1992. Since the operating conditions were well defined at that time, the experimental results can be used as a benchmark. It may be noted that this benchmark is one of very few high enrichment benchmarks available. In our simulations experimental conditions were thoroughly simulated: fuel elements and control rods were precisely modeled as well as entire core configuration and the vicinity of the core. ENDF/B-VI and ENDF/B-V libraries were used. Partial results of benchmark calculations are presented. Excellent agreement of core criticality, excess reactivity and control rod worths can be observed. (author)

  1. Benchmarking HIV health care

    DEFF Research Database (Denmark)

    Podlekareva, Daria; Reekie, Joanne; Mocroft, Amanda

    2012-01-01

    ABSTRACT: BACKGROUND: State-of-the-art care involving the utilisation of multiple health care interventions is the basis for an optimal long-term clinical prognosis for HIV-patients. We evaluated health care for HIV-patients based on four key indicators. METHODS: Four indicators of health care we...... document pronounced regional differences in adherence to guidelines and can help to identify gaps and direct target interventions. It may serve as a tool for assessment and benchmarking the clinical management of HIV-patients in any setting worldwide....

  2. Benchmarking Cloud Storage Systems

    OpenAIRE

    Wang, Xing

    2014-01-01

    With the rise of cloud computing, many cloud storage systems like Dropbox, Google Drive and Mega have been built to provide decentralized and reliable file storage. It is thus of prime importance to know their features, performance, and the best way to make use of them. In this context, we introduce BenchCloud, a tool designed as part of this thesis to conveniently and efficiently benchmark any cloud storage system. First, we provide a study of six commonly-used cloud storage systems to ident...

  3. The COST Benchmark

    DEFF Research Database (Denmark)

    Jensen, Christian Søndergaard; Tiesyte, Dalia; Tradisauskas, Nerius

    2006-01-01

    An infrastructure is emerging that enables the positioning of populations of on-line, mobile service users. In step with this, research in the management of moving objects has attracted substantial attention. In particular, quite a few proposals now exist for the indexing of moving objects...... takes into account that the available positions of the moving objects are inaccurate, an aspect largely ignored in previous indexing research. The concepts of data and query enlargement are introduced for addressing inaccuracy. As proof of concepts of the benchmark, the paper covers the application...

  4. Calculation methods for advanced concept light water reactor lattices

    International Nuclear Information System (INIS)

    Carmona, S.

    1986-01-01

    In the last few years s several advanced concepts for fuel rod lattices have been studied. Improved fuel utilization is one of the major aims in the development of new fuel rod designs and lattice modifications. By these changes s better performance in fuel economics s fuel burnup and material endurance can be achieved in the frame of the well-known basic Light Water Reactor technology. Among the new concepts involved in these studies that have attracted serious attention are lattices consisting of arrays of annular rods duplex pellet rods or tight multicells. These new designs of fuel rods and lattices present several computational problems. The treatment of resonance shielded cross sections is a crucial point in the analyses of these advanced concepts . The purpose of this study was to assess adequate approximation methods for calculating as accurately as possible, resonance shielding for these new lattices. Although detailed and exact computational methods for the evaluation of the resonance shielding in these lattices are possible, they are quite inefficient when used in lattice codes. The computer time and memory required for this kind of computations are too large to be used in an acceptable routine manner. In order to over- come these limitations and to make the analyses possible with reasonable use of computer resources s approximation methods are necessary. Usual approximation methods, for the resonance energy regions used in routine lattice computer codes, can not adequately handle the evaluation of these new fuel rod lattices. The main contribution of the present work to advanced lattice concepts is the development of an equivalence principle for the calculation of resonance shielding in the annular fuel pellet zone of duplex pellets; the duplex pellet in this treatment consists of two fuel zones with the same absorber isotope in both regions. In the transition from a single duplex rod to an infinite array of this kind of fuel rods, the similarity of the

  5. RECENT ADDITIONS OF CRITICALITY SAFETY RELATED INTEGRAL BENCHMARK DATA TO THE ICSBEP AND IRPHEP HANDBOOKS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Sartori

    2009-09-01

    High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions.

  6. Criticality safety benchmarking of PASC-3 and ECNJEF1.1

    International Nuclear Information System (INIS)

    Li, J.

    1992-09-01

    To validate the code system PASC-3 and the multigroup cross section library ECNJEF1.1 on various applications many benchmarks are required. This report presents the results of critically safety benchmarking for five calculational and four experimental benchmarks. These benchmarks are related to the transport package of fissile materials such as spent fuel. The fissile nuclides in these benchmarks are 235 U and 239 Pu. The modules of PASC-3 which have been used for the calculations are BONAMI, NITAWL and KENO.5A. The final results for the experimental benchmarks do agree well with experimental data. For the calculational benchmarks the results presented here are in reasonable agreement with the results from other investigations. (author). 8 refs.; 20 figs.; 5 tabs

  7. Recent additions of criticality safety related integral benchmark data to the ICSBEP and IRPHEP handbooks

    International Nuclear Information System (INIS)

    Briggs, J. B.; Scott, L.; Rugama, Y.; Sartori, E.

    2009-01-01

    High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions. (authors)

  8. REcent Additions Of Criticality Safety Related Integral Benchmark Data To The Icsbep And Irphep Handbooks

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Scott, Lori; Rugama, Yolanda; Sartori, Enrico

    2009-01-01

    High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions.

  9. On techniques of ATR lattice computation

    International Nuclear Information System (INIS)

    1997-08-01

    Lattice computation is to compute the average nuclear constants of unit fuel lattice which are required for computing core nuclear characteristics such as core power distribution and reactivity characteristics. The main nuclear constants are infinite multiplying rate, neutron movement area, cross section for diffusion computation, local power distribution and isotope composition. As for the lattice computation code, WIMS-ATR is used, which is based on the WIMS-D code developed in U.K., and for the purpose of heightening the accuracy of analysis, which was improved by adding heavy water scattering cross section considering the temperature dependence by Honeck model. For the computation of the neutron absorption by control rods, LOIEL BLUE code is used. The extrapolation distance of neutron flux on control rod surfaces is computed by using THERMOS and DTF codes, and the lattice constants of adjoining lattices are computed by using the WIMS-ATR code. As for the WIMS-ATR code, the computation flow and nuclear data library, and as for the LOIEL BLUE code, the computation flow are explained. The local power distribution in fuel assemblies determined by the WIMS-ATR code was verified with the measured data, and the results are reported. (K.I.)

  10. Lattice theory for nonspecialists

    International Nuclear Information System (INIS)

    Hari Dass, N.D.

    1984-01-01

    These lectures were delivered as part of the academic training programme at the NIKHEF-H. These lectures were intended primarily for experimentalists, and theorists not specializing in lattice methods. The goal was to present the essential spirit behind the lattice approach and consequently the author has concentrated mostly on issues of principle rather than on presenting a large amount of detail. In particular, the author emphasizes the deep theoretical infra-structure that has made lattice studies meaningful. At the same time, he has avoided the use of heavy formalisms as they tend to obscure the basic issues for people trying to approach this subject for the first time. The essential ideas are illustrated with elementary soluble examples not involving complicated mathematics. The following subjects are discussed: three ways of solving the harmonic oscillator problem; latticization; gauge fields on a lattice; QCD observables; how to solve lattice theories. (Auth.)

  11. Lattice gauge theories

    International Nuclear Information System (INIS)

    Creutz, M.

    1983-04-01

    In the last few years lattice gauge theory has become the primary tool for the study of nonperturbative phenomena in gauge theories. The lattice serves as an ultraviolet cutoff, rendering the theory well defined and amenable to numerical and analytical work. Of course, as with any cutoff, at the end of a calculation one must consider the limit of vanishing lattice spacing in order to draw conclusions on the physical continuum limit theory. The lattice has the advantage over other regulators that it is not tied to the Feynman expansion. This opens the possibility of other approximation schemes than conventional perturbation theory. Thus Wilson used a high temperature expansion to demonstrate confinement in the strong coupling limit. Monte Carlo simulations have dominated the research in lattice gauge theory for the last four years, giving first principle calculations of nonperturbative parameters characterizing the continuum limit. Some of the recent results with lattice calculations are reviewed

  12. JENDL-3.3 thermal reactor benchmark test

    International Nuclear Information System (INIS)

    Akie, Hiroshi

    2001-01-01

    Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

  13. Wilson Dslash Kernel From Lattice QCD Optimization

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Balint [Jefferson Lab, Newport News, VA; Smelyanskiy, Mikhail [Parallel Computing Lab, Intel Corporation, California, USA; Kalamkar, Dhiraj D. [Parallel Computing Lab, Intel Corporation, India; Vaidyanathan, Karthikeyan [Parallel Computing Lab, Intel Corporation, India

    2015-07-01

    Lattice Quantum Chromodynamics (LQCD) is a numerical technique used for calculations in Theoretical Nuclear and High Energy Physics. LQCD is traditionally one of the first applications ported to many new high performance computing architectures and indeed LQCD practitioners have been known to design and build custom LQCD computers. Lattice QCD kernels are frequently used as benchmarks (e.g. 168.wupwise in the SPEC suite) and are generally well understood, and as such are ideal to illustrate several optimization techniques. In this chapter we will detail our work in optimizing the Wilson-Dslash kernels for Intel Xeon Phi, however, as we will show the technique gives excellent performance on regular Xeon Architecture as well.

  14. On Traveling Waves in Lattices: The Case of Riccati Lattices

    Science.gov (United States)

    Dimitrova, Zlatinka

    2012-09-01

    The method of simplest equation is applied for analysis of a class of lattices described by differential-difference equations that admit traveling-wave solutions constructed on the basis of the solution of the Riccati equation. We denote such lattices as Riccati lattices. We search for Riccati lattices within two classes of lattices: generalized Lotka-Volterra lattices and generalized Holling lattices. We show that from the class of generalized Lotka-Volterra lattices only the Wadati lattice belongs to the class of Riccati lattices. Opposite to this many lattices from the Holling class are Riccati lattices. We construct exact traveling wave solutions on the basis of the solution of Riccati equation for three members of the class of generalized Holling lattices.

  15. Benchmarking multimedia performance

    Science.gov (United States)

    Zandi, Ahmad; Sudharsanan, Subramania I.

    1998-03-01

    With the introduction of faster processors and special instruction sets tailored to multimedia, a number of exciting applications are now feasible on the desktops. Among these is the DVD playback consisting, among other things, of MPEG-2 video and Dolby digital audio or MPEG-2 audio. Other multimedia applications such as video conferencing and speech recognition are also becoming popular on computer systems. In view of this tremendous interest in multimedia, a group of major computer companies have formed, Multimedia Benchmarks Committee as part of Standard Performance Evaluation Corp. to address the performance issues of multimedia applications. The approach is multi-tiered with three tiers of fidelity from minimal to full compliant. In each case the fidelity of the bitstream reconstruction as well as quality of the video or audio output are measured and the system is classified accordingly. At the next step the performance of the system is measured. In many multimedia applications such as the DVD playback the application needs to be run at a specific rate. In this case the measurement of the excess processing power, makes all the difference. All these make a system level, application based, multimedia benchmark very challenging. Several ideas and methodologies for each aspect of the problems will be presented and analyzed.

  16. A benchmarking study

    Directory of Open Access Journals (Sweden)

    H. Groessing

    2015-02-01

    Full Text Available A benchmark study for permeability measurement is presented. In the past studies of other research groups which focused on the reproducibility of 1D-permeability measurements showed high standard deviations of the gained permeability values (25%, even though a defined test rig with required specifications was used. Within this study, the reproducibility of capacitive in-plane permeability testing system measurements was benchmarked by comparing results of two research sites using this technology. The reproducibility was compared by using a glass fibre woven textile and carbon fibre non crimped fabric (NCF. These two material types were taken into consideration due to the different electrical properties of glass and carbon with respect to dielectric capacitive sensors of the permeability measurement systems. In order to determine the unsaturated permeability characteristics as function of fibre volume content the measurements were executed at three different fibre volume contents including five repetitions. It was found that the stability and reproducibility of the presentedin-plane permeability measurement system is very good in the case of the glass fibre woven textiles. This is true for the comparison of the repetition measurements as well as for the comparison between the two different permeameters. These positive results were confirmed by a comparison to permeability values of the same textile gained with an older generation permeameter applying the same measurement technology. Also it was shown, that a correct determination of the grammage and the material density are crucial for correct correlation of measured permeability values and fibre volume contents.

  17. Benchmarking Using Basic DBMS Operations

    Science.gov (United States)

    Crolotte, Alain; Ghazal, Ahmad

    The TPC-H benchmark proved to be successful in the decision support area. Many commercial database vendors and their related hardware vendors used these benchmarks to show the superiority and competitive edge of their products. However, over time, the TPC-H became less representative of industry trends as vendors keep tuning their database to this benchmark-specific workload. In this paper, we present XMarq, a simple benchmark framework that can be used to compare various software/hardware combinations. Our benchmark model is currently composed of 25 queries that measure the performance of basic operations such as scans, aggregations, joins and index access. This benchmark model is based on the TPC-H data model due to its maturity and well-understood data generation capability. We also propose metrics to evaluate single-system performance and compare two systems. Finally we illustrate the effectiveness of this model by showing experimental results comparing two systems under different conditions.

  18. Lattice degeneracies of fermions

    International Nuclear Information System (INIS)

    Raszillier, H.

    1983-10-01

    We present a detailed description of the minimal degeneracies of geometric (Kaehler) fermions on all the lattices of maximal symmetries in n = 1, ..., 4 dimensions. We also determine the isolated orbits of the maximal symmetry groups, which are related to the minimal numbers of ''naive'' fermions on the reciprocals of these lattices. It turns out that on the self-reciprocal lattices the minimal numbers of naive fermions are equal to the minimal numbers of degrees of freedom of geometric fermions. The description we give relies on the close connection of the maximal lattice symmetry groups with (affine) Weyl groups of root systems of (semi-) simple Lie algebras. (orig.)

  19. Twisted mass lattice QCD

    International Nuclear Information System (INIS)

    Shindler, A.

    2007-07-01

    I review the theoretical foundations, properties as well as the simulation results obtained so far of a variant of the Wilson lattice QCD formulation: Wilson twisted mass lattice QCD. Emphasis is put on the discretization errors and on the effects of these discretization errors on the phase structure for Wilson-like fermions in the chiral limit. The possibility to use in lattice simulations different lattice actions for sea and valence quarks to ease the renormalization patterns of phenomenologically relevant local operators, is also discussed. (orig.)

  20. Twisted mass lattice QCD

    Energy Technology Data Exchange (ETDEWEB)

    Shindler, A. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany). John von Neumann-Inst. fuer Computing NIC

    2007-07-15

    I review the theoretical foundations, properties as well as the simulation results obtained so far of a variant of the Wilson lattice QCD formulation: Wilson twisted mass lattice QCD. Emphasis is put on the discretization errors and on the effects of these discretization errors on the phase structure for Wilson-like fermions in the chiral limit. The possibility to use in lattice simulations different lattice actions for sea and valence quarks to ease the renormalization patterns of phenomenologically relevant local operators, is also discussed. (orig.)

  1. Low-energy scattering on the lattice

    International Nuclear Information System (INIS)

    Bour Bour, Shahin

    2014-01-01

    In this thesis we present precision benchmark calculations for two-component fermions in the unitarity limit using an ab initio method, namely Hamiltonian lattice formalism. We calculate the ground state energy for unpolarized four particles (Fermi gas) in a periodic cube as a fraction of the ground state energy of the non-interacting system for two independent representations of the lattice Hamiltonians. We obtain the values 0.211(2) and 0.210(2). These results are in full agreement with the Euclidean lattice and fixed-node diffusion Monte Carlo calculations. We also give an expression for the energy corrections to the binding energy of a bound state in a moving frame. These corrections contain information about the mass and number of the constituents and are topological in origin and will have a broad applications to the lattice calculations of nucleons, nuclei, hadronic molecules and cold atoms. As one of its applications we use this expression and determine the low-energy parameters for the fermion dimer elastic scattering in shallow binding limit. For our lattice calculations we use Luescher's finite volume method. From the lattice calculations we find κa fd =1.174(9) and κr fd =-0.029(13), where κ represents the binding momentum of dimer and a fd (r fd ) denotes the scattering length (effective-range). These results are confirmed by the continuum calculations using the Skorniakov-Ter-Martirosian integral equation which gives 1.17907(1) and -0.0383(3) for the scattering length and effective range, respectively.

  2. Benchmark and physics testing of LIFE-4C. Summary

    International Nuclear Information System (INIS)

    Liu, Y.Y.

    1984-06-01

    LIFE-4C is a steady-state/transient analysis code developed for performance evaluation of carbide [(U,Pu)C and UC] fuel elements in advanced LMFBRs. This paper summarizes selected results obtained during a crucial step in the development of LIFE-4C - benchmark and physics testing

  3. Benchmarking & European Sustainable Transport Policies

    DEFF Research Database (Denmark)

    Gudmundsson, H.

    2003-01-01

    , Benchmarking is one of the management tools that have recently been introduced in the transport sector. It is rapidly being applied to a wide range of transport operations, services and policies. This paper is a contribution to the discussion of the role of benchmarking in the future efforts to...... contribution to the discussions within the Eusponsored BEST Thematic Network (Benchmarking European Sustainable Transport) which ran from 2000 to 2003....

  4. Benchmarking in Czech Higher Education

    OpenAIRE

    Plaček Michal; Ochrana František; Půček Milan

    2015-01-01

    The first part of this article surveys the current experience with the use of benchmarking at Czech universities specializing in economics and management. The results indicate that collaborative benchmarking is not used on this level today, but most actors show some interest in its introduction. The expression of the need for it and the importance of benchmarking as a very suitable performance-management tool in less developed countries are the impetus for the second part of our article. Base...

  5. Power reactor pressure vessel benchmarks

    International Nuclear Information System (INIS)

    Rahn, F.J.

    1978-01-01

    A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

  6. S/sub n/ analysis of the TRX metal lattices with ENDF/B version III data

    International Nuclear Information System (INIS)

    Wheeler, F.J.

    1975-01-01

    Two critical assemblies, designated as thermal-reactor benchmarks TRX-1 and TRX-2 for ENDF/B data testing, were analyzed using the one-dimensional S/sub n/-theory code SCAMP. The two assemblies were simple lattices of aluminum-clad, uranium-metal fuel rods in triangular arrays with D 2 O as moderator and reflector. The fuel was low-enriched (1.3 percent 235 U), 0.387-inch in diameter and had an active height of 48 inches. The volume ratio of water to uranium was 2.35 for the TRX-1 lattice and 4.02 for TRX-2. Full-core S/sub n/ calculations based on Version III data were performed for these assemblies and the results obtained were compared with the measured values of the multiplication factors, the ratio of epithermal-to-thermal neutron capture in 238 U, the ratio of epithermal-to-thermal fission in 235 U, the ratio of 238 U fission to 235 U fission, and the ratio of capture in 238 U to fission in 235 U. Reaction rates were obtained from a central region of the full-core problems. Multigroup cross sections for the reactor calculation were obtained from S/sub n/ cell calculations with resonance self-shielding calculated using the RABBLE treatment. The results of the analyses are generally consistent with results obtained by other investigators

  7. Sensitivity Analysis of OECD Benchmark Tests in BISON

    Energy Technology Data Exchange (ETDEWEB)

    Swiler, Laura Painton [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schmidt, Rodney C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Williamson, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This report summarizes a NEAMS (Nuclear Energy Advanced Modeling and Simulation) project focused on sensitivity analysis of a fuels performance benchmark problem. The benchmark problem was defined by the Uncertainty Analysis in Modeling working group of the Nuclear Science Committee, part of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD ). The benchmark problem involv ed steady - state behavior of a fuel pin in a Pressurized Water Reactor (PWR). The problem was created in the BISON Fuels Performance code. Dakota was used to generate and analyze 300 samples of 17 input parameters defining core boundary conditions, manuf acturing tolerances , and fuel properties. There were 24 responses of interest, including fuel centerline temperatures at a variety of locations and burnup levels, fission gas released, axial elongation of the fuel pin, etc. Pearson and Spearman correlatio n coefficients and Sobol' variance - based indices were used to perform the sensitivity analysis. This report summarizes the process and presents results from this study.

  8. Simplified two and three dimensional HTTR benchmark problems

    International Nuclear Information System (INIS)

    Zhang Zhan; Rahnema, Farzad; Zhang Dingkang; Pounders, Justin M.; Ougouag, Abderrafi M.

    2011-01-01

    To assess the accuracy of diffusion or transport methods for reactor calculations, it is desirable to create heterogeneous benchmark problems that are typical of whole core configurations. In this paper we have created two and three dimensional numerical benchmark problems typical of high temperature gas cooled prismatic cores. Additionally, a single cell and single block benchmark problems are also included. These problems were derived from the HTTR start-up experiment. Since the primary utility of the benchmark problems is in code-to-code verification, minor details regarding geometry and material specification of the original experiment have been simplified while retaining the heterogeneity and the major physics properties of the core from a neutronics viewpoint. A six-group material (macroscopic) cross section library has been generated for the benchmark problems using the lattice depletion code HELIOS. Using this library, Monte Carlo solutions are presented for three configurations (all-rods-in, partially-controlled and all-rods-out) for both the 2D and 3D problems. These solutions include the core eigenvalues, the block (assembly) averaged fission densities, local peaking factors, the absorption densities in the burnable poison and control rods, and pin fission density distribution for selected blocks. Also included are the solutions for the single cell and single block problems.

  9. Stylized whole-core benchmark of the Integral Inherently Safe Light Water Reactor (I2S-LWR) concept

    International Nuclear Information System (INIS)

    Hon, Ryan; Kooreman, Gabriel; Rahnema, Farzad; Petrovic, Bojan

    2017-01-01

    Highlights: • A stylized benchmark specification of the I2S-LWR core. • A library of cross sections were generated in both 8 and 47 groups. • Monte Carlo solutions generated for the 8 group library using MCNP5. • Cross sections and pin fission densities provided in journal’s repository. - Abstract: The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) is a pressurized water reactor (PWR) concept under development by a multi-institutional team led by Georgia Tech. The core is similar in size to small 2-loop PWRs while having the power level of current large reactors (∼1000 MWe) but using uranium silicide fuel and advanced stainless steel cladding. A stylized benchmark specification of the I 2 S-LWR core has been developed in order to test whole-core neutronics codes and methods. For simplification the core was split into 57 distinct material regions for cross section generation. Cross sections were generated using the lattice physics code HELIOS version 1.10 in both 8 and 47 groups. Monte Carlo solutions, including eigenvalue and pin fission densities, were generated for the 8 group library using MCNP5. Due to space limitations in this paper, the full cross section library and normalized pin fission density results are provided in the journal’s electronic repository.

  10. Model based energy benchmarking for glass furnace

    International Nuclear Information System (INIS)

    Sardeshpande, Vishal; Gaitonde, U.N.; Banerjee, Rangan

    2007-01-01

    Energy benchmarking of processes is important for setting energy efficiency targets and planning energy management strategies. Most approaches used for energy benchmarking are based on statistical methods by comparing with a sample of existing plants. This paper presents a model based approach for benchmarking of energy intensive industrial processes and illustrates this approach for industrial glass furnaces. A simulation model for a glass furnace is developed using mass and energy balances, and heat loss equations for the different zones and empirical equations based on operating practices. The model is checked with field data from end fired industrial glass furnaces in India. The simulation model enables calculation of the energy performance of a given furnace design. The model results show the potential for improvement and the impact of different operating and design preferences on specific energy consumption. A case study for a 100 TPD end fired furnace is presented. An achievable minimum energy consumption of about 3830 kJ/kg is estimated for this furnace. The useful heat carried by glass is about 53% of the heat supplied by the fuel. Actual furnaces operating at these production scales have a potential for reduction in energy consumption of about 20-25%

  11. A thermo mechanical benchmark calculation of a hexagonal can in the BTI accident with INCA code

    International Nuclear Information System (INIS)

    Zucchini, A.

    1988-01-01

    The thermomechanical behaviour of an hexagonal can in a benchmark problem (simulating the conditions of a BTI accident in a fuel assembly) is examined by means of the INCA code and the results systematically compared with those of ADINA

  12. Lattice Boltzmann Pore-Scale Investigation of Coupled Physical-electrochemical Processes in C/Pt and Non-Precious Metal Cathode Catalyst Layers in Proton Exchange Membrane Fuel Cells

    International Nuclear Information System (INIS)

    Chen, Li; Wu, Gang; Holby, Edward F; Zelenay, Piotr; Tao, Wen-Quan; Kang, Qinjun

    2015-01-01

    Highlights: • Nanoscale structures of catalyst layer are reconstructed. • Pore-scale simulation is performed to predict macroscopic transport properties. • Reactive transport in catalyst layer with non-precious metal and Pt catalysts is studied. • Mesopores rather than micropores are required to enhance mass transport. - Abstract: High-resolution porous structures of catalyst layers (CLs) including non-precious metal catalysts (NPMCs) or Pt for proton exchange membrane fuel cells are reconstructed using the quartet structure generation set. The nanoscale structures are analyzed in terms of pore size distribution, specific surface area, and phase connectivity. Pore-scale simulation methods based on the lattice Boltzmann method are developed to predict the macroscopic transport properties in CLs. The non-uniform distribution of ionomer in CL generates more tortuous pathways for reactant transport, greatly reducing the effective diffusivity. The tortuosity of CLs is much higher than that adopted by the Bruggeman equation. Knudsen diffusion plays a significant role in oxygen diffusion and significantly reduces the effective diffusivity. Reactive transport inside the CLs is also investigated. Although the reactive surface area of the non-precious metal catalyst (NPMC) CL is much higher than that of the Pt CL, the oxygen reaction rate is lower in the NPMC CL due to the much lower reaction rate coefficient. Although pores of a few nanometers in size can increase the number of reactive sites in NPMC CLs, they contribute little to enhance the mass transport. Mesopores, which are a few tens of nanometers or larger in size, are shown to be required in order to increase the mass transport rate

  13. Benchmarking Academic Anatomic Pathologists

    Directory of Open Access Journals (Sweden)

    Barbara S. Ducatman MD

    2016-10-01

    Full Text Available The most common benchmarks for faculty productivity are derived from Medical Group Management Association (MGMA or Vizient-AAMC Faculty Practice Solutions Center ® (FPSC databases. The Association of Pathology Chairs has also collected similar survey data for several years. We examined the Association of Pathology Chairs annual faculty productivity data and compared it with MGMA and FPSC data to understand the value, inherent flaws, and limitations of benchmarking data. We hypothesized that the variability in calculated faculty productivity is due to the type of practice model and clinical effort allocation. Data from the Association of Pathology Chairs survey on 629 surgical pathologists and/or anatomic pathologists from 51 programs were analyzed. From review of service assignments, we were able to assign each pathologist to a specific practice model: general anatomic pathologists/surgical pathologists, 1 or more subspecialties, or a hybrid of the 2 models. There were statistically significant differences among academic ranks and practice types. When we analyzed our data using each organization’s methods, the median results for the anatomic pathologists/surgical pathologists general practice model compared to MGMA and FPSC results for anatomic and/or surgical pathology were quite close. Both MGMA and FPSC data exclude a significant proportion of academic pathologists with clinical duties. We used the more inclusive FPSC definition of clinical “full-time faculty” (0.60 clinical full-time equivalent and above. The correlation between clinical full-time equivalent effort allocation, annual days on service, and annual work relative value unit productivity was poor. This study demonstrates that effort allocations are variable across academic departments of pathology and do not correlate well with either work relative value unit effort or reported days on service. Although the Association of Pathology Chairs–reported median work relative

  14. Self-benchmarking Guide for Cleanrooms: Metrics, Benchmarks, Actions

    Energy Technology Data Exchange (ETDEWEB)

    Mathew, Paul; Sartor, Dale; Tschudi, William

    2009-07-13

    This guide describes energy efficiency metrics and benchmarks that can be used to track the performance of and identify potential opportunities to reduce energy use in laboratory buildings. This guide is primarily intended for personnel who have responsibility for managing energy use in existing laboratory facilities - including facilities managers, energy managers, and their engineering consultants. Additionally, laboratory planners and designers may also use the metrics and benchmarks described in this guide for goal-setting in new construction or major renovation. This guide provides the following information: (1) A step-by-step outline of the benchmarking process. (2) A set of performance metrics for the whole building as well as individual systems. For each metric, the guide provides a definition, performance benchmarks, and potential actions that can be inferred from evaluating this metric. (3) A list and descriptions of the data required for computing the metrics. This guide is complemented by spreadsheet templates for data collection and for computing the benchmarking metrics. This guide builds on prior research supported by the national Laboratories for the 21st Century (Labs21) program, supported by the U.S. Department of Energy and the U.S. Environmental Protection Agency. Much of the benchmarking data are drawn from the Labs21 benchmarking database and technical guides. Additional benchmark data were obtained from engineering experts including laboratory designers and energy managers.

  15. Self-benchmarking Guide for Laboratory Buildings: Metrics, Benchmarks, Actions

    Energy Technology Data Exchange (ETDEWEB)

    Mathew, Paul; Greenberg, Steve; Sartor, Dale

    2009-07-13

    This guide describes energy efficiency metrics and benchmarks that can be used to track the performance of and identify potential opportunities to reduce energy use in laboratory buildings. This guide is primarily intended for personnel who have responsibility for managing energy use in existing laboratory facilities - including facilities managers, energy managers, and their engineering consultants. Additionally, laboratory planners and designers may also use the metrics and benchmarks described in this guide for goal-setting in new construction or major renovation. This guide provides the following information: (1) A step-by-step outline of the benchmarking process. (2) A set of performance metrics for the whole building as well as individual systems. For each metric, the guide provides a definition, performance benchmarks, and potential actions that can be inferred from evaluating this metric. (3) A list and descriptions of the data required for computing the metrics. This guide is complemented by spreadsheet templates for data collection and for computing the benchmarking metrics. This guide builds on prior research supported by the national Laboratories for the 21st Century (Labs21) program, supported by the U.S. Department of Energy and the U.S. Environmental Protection Agency. Much of the benchmarking data are drawn from the Labs21 benchmarking database and technical guides. Additional benchmark data were obtained from engineering experts including laboratory designers and energy managers.

  16. Shielding benchmark test

    International Nuclear Information System (INIS)

    Kawai, Masayoshi

    1984-01-01

    Iron data in JENDL-2 have been tested by analyzing shielding benchmark experiments for neutron transmission through iron block performed at KFK using CF-252 neutron source and at ORNL using collimated neutron beam from reactor. The analyses are made by a shielding analysis code system RADHEAT-V4 developed at JAERI. The calculated results are compared with the measured data. As for the KFK experiments, the C/E values are about 1.1. For the ORNL experiments, the calculated values agree with the measured data within an accuracy of 33% for the off-center geometry. The d-t neutron transmission measurements through carbon sphere made at LLNL are also analyzed preliminarily by using the revised JENDL data for fusion neutronics calculation. (author)

  17. Nuclear lattice simulations

    Directory of Open Access Journals (Sweden)

    Epelbaum E.

    2010-04-01

    Full Text Available We review recent progress on nuclear lattice simulations using chiral effective field theory. We discuss lattice results for dilute neutron matter at next-to-leading order, three-body forces at next-to-next-toleading order, isospin-breaking and Coulomb effects, and the binding energy of light nuclei.

  18. Lattice Higgs models

    International Nuclear Information System (INIS)

    Jersak, J.

    1986-01-01

    This year has brought a sudden interest in lattice Higgs models. After five years of only modest activity we now have many new results obtained both by analytic and Monte Carlo methods. This talk is a review of the present state of lattice Higgs models with particular emphasis on the recent development

  19. Benchmarking monthly homogenization algorithms

    Science.gov (United States)

    Venema, V. K. C.; Mestre, O.; Aguilar, E.; Auer, I.; Guijarro, J. A.; Domonkos, P.; Vertacnik, G.; Szentimrey, T.; Stepanek, P.; Zahradnicek, P.; Viarre, J.; Müller-Westermeier, G.; Lakatos, M.; Williams, C. N.; Menne, M.; Lindau, R.; Rasol, D.; Rustemeier, E.; Kolokythas, K.; Marinova, T.; Andresen, L.; Acquaotta, F.; Fratianni, S.; Cheval, S.; Klancar, M.; Brunetti, M.; Gruber, C.; Prohom Duran, M.; Likso, T.; Esteban, P.; Brandsma, T.

    2011-08-01

    The COST (European Cooperation in Science and Technology) Action ES0601: Advances in homogenization methods of climate series: an integrated approach (HOME) has executed a blind intercomparison and validation study for monthly homogenization algorithms. Time series of monthly temperature and precipitation were evaluated because of their importance for climate studies and because they represent two important types of statistics (additive and multiplicative). The algorithms were validated against a realistic benchmark dataset. The benchmark contains real inhomogeneous data as well as simulated data with inserted inhomogeneities. Random break-type inhomogeneities were added to the simulated datasets modeled as a Poisson process with normally distributed breakpoint sizes. To approximate real world conditions, breaks were introduced that occur simultaneously in multiple station series within a simulated network of station data. The simulated time series also contained outliers, missing data periods and local station trends. Further, a stochastic nonlinear global (network-wide) trend was added. Participants provided 25 separate homogenized contributions as part of the blind study as well as 22 additional solutions submitted after the details of the imposed inhomogeneities were revealed. These homogenized datasets were assessed by a number of performance metrics including (i) the centered root mean square error relative to the true homogeneous value at various averaging scales, (ii) the error in linear trend estimates and (iii) traditional contingency skill scores. The metrics were computed both using the individual station series as well as the network average regional series. The performance of the contributions depends significantly on the error metric considered. Contingency scores by themselves are not very informative. Although relative homogenization algorithms typically improve the homogeneity of temperature data, only the best ones improve precipitation data

  20. Benchmarking foreign electronics technologies

    Energy Technology Data Exchange (ETDEWEB)

    Bostian, C.W.; Hodges, D.A.; Leachman, R.C.; Sheridan, T.B.; Tsang, W.T.; White, R.M.

    1994-12-01

    This report has been drafted in response to a request from the Japanese Technology Evaluation Center`s (JTEC) Panel on Benchmarking Select Technologies. Since April 1991, the Competitive Semiconductor Manufacturing (CSM) Program at the University of California at Berkeley has been engaged in a detailed study of quality, productivity, and competitiveness in semiconductor manufacturing worldwide. The program is a joint activity of the College of Engineering, the Haas School of Business, and the Berkeley Roundtable on the International Economy, under sponsorship of the Alfred P. Sloan Foundation, and with the cooperation of semiconductor producers from Asia, Europe and the United States. Professors David A. Hodges and Robert C. Leachman are the project`s Co-Directors. The present report for JTEC is primarily based on data and analysis drawn from that continuing program. The CSM program is being conducted by faculty, graduate students and research staff from UC Berkeley`s Schools of Engineering and Business, and Department of Economics. Many of the participating firms are represented on the program`s Industry Advisory Board. The Board played an important role in defining the research agenda. A pilot study was conducted in 1991 with the cooperation of three semiconductor plants. The research plan and survey documents were thereby refined. The main phase of the CSM benchmarking study began in mid-1992 and will continue at least through 1997. reports are presented on the manufacture of integrated circuits; data storage; wireless technology; human-machine interfaces; and optoelectronics. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  1. SSI and structural benchmarks

    International Nuclear Information System (INIS)

    Philippacopoulos, A.J.; Miller, C.A.; Costantino, C.J.; Graves, H.

    1987-01-01

    This paper presents the latest results of the ongoing program entitled, Standard Problems for Structural Computer Codes, currently being worked on at BNL for the USNRC, Office of Nuclear Regulatory Research. During FY 1986, efforts were focussed on three tasks, namely, (1) an investigation of ground water effects on the response of Category I structures, (2) the Soil-Structure Interaction Workshop and (3) studies on structural benchmarks associated with Category I structures. The objective of the studies on ground water effects is to verify the applicability and the limitations of the SSI methods currently used by the industry in performing seismic evaluations of nuclear plants which are located at sites with high water tables. In a previous study by BNL (NUREG/CR-4588), it has been concluded that the pore water can influence significantly the soil-structure interaction process. This result, however, is based on the assumption of fully saturated soil profiles. Consequently, the work was further extended to include cases associated with variable water table depths. In this paper, results related to cut-off depths beyond which the pore water effects can be ignored in seismic calculations, are addressed. Comprehensive numerical data are given for soil configurations typical to those encountered in nuclear plant sites. These data were generated by using a modified version of the SLAM code which is capable of handling problems related to the dynamic response of saturated soils. Further, the paper presents some key aspects of the Soil-Structure Interaction Workshop (NUREG/CP-0054) which was held in Bethesda, MD on June 1, 1986. Finally, recent efforts related to the task on the structural benchmarks are described

  2. Effects of uncertainties of experimental data in the benchmarking of a computer code

    International Nuclear Information System (INIS)

    Meulemeester, E. de; Bouffioux, P.; Demeester, J.

    1980-01-01

    Fuel rod performance modelling is sometimes taken in an academical way. The experience of the COMETHE code development since 1967 has clearly shown that benchmarking was the most important part of modelling development. Unfortunately, it requires well characterized data. Although, the two examples presented here were not intended for benchmarking, as the COMETHE calculations were only performed for an interpretation of the results, they illustrate the effects of a lack of fuel characterization and of the power history uncertainties

  3. Benchmarking of SIMULATE-3 on engineering workstations

    International Nuclear Information System (INIS)

    Karlson, C.F.; Reed, M.L.; Webb, J.R.; Elzea, J.D.

    1990-01-01

    The nuclear fuel management department of Arizona Public Service Company (APS) has evaluated various computer platforms for a departmental engineering and business work-station local area network (LAN). Historically, centralized mainframe computer systems have been utilized for engineering calculations. Increasing usage and the resulting longer response times on the company mainframe system and the relative cost differential between a mainframe upgrade and workstation technology justified the examination of current workstations. A primary concern was the time necessary to turn around routine reactor physics reload and analysis calculations. Computers ranging from a Definicon 68020 processing board in an AT compatible personal computer up to an IBM 3090 mainframe were benchmarked. The SIMULATE-3 advanced nodal code was selected for benchmarking based on its extensive use in nuclear fuel management. SIMULATE-3 is used at APS for reload scoping, design verification, core follow, and providing predictions of reactor behavior under nominal conditions and planned reactor maneuvering, such as axial shape control during start-up and shutdown

  4. Review for session K - benchmarks

    International Nuclear Information System (INIS)

    McCracken, A.K.

    1980-01-01

    Eight of the papers to be considered in Session K are directly concerned, at least in part, with the Pool Critical Assembly (P.C.A.) benchmark at Oak Ridge. The remaining seven papers in this session, the subject of this review, are concerned with a variety of topics related to the general theme of Benchmarks and will be considered individually

  5. Internal Benchmarking for Institutional Effectiveness

    Science.gov (United States)

    Ronco, Sharron L.

    2012-01-01

    Internal benchmarking is an established practice in business and industry for identifying best in-house practices and disseminating the knowledge about those practices to other groups in the organization. Internal benchmarking can be done with structures, processes, outcomes, or even individuals. In colleges or universities with multicampuses or a…

  6. Entropy-based benchmarking methods

    NARCIS (Netherlands)

    Temurshoev, Umed

    2012-01-01

    We argue that benchmarking sign-volatile series should be based on the principle of movement and sign preservation, which states that a bench-marked series should reproduce the movement and signs in the original series. We show that the widely used variants of Denton (1971) method and the growth

  7. Benchmark simulation models, quo vadis?

    DEFF Research Database (Denmark)

    Jeppsson, U.; Alex, J; Batstone, D. J.

    2013-01-01

    As the work of the IWA Task Group on Benchmarking of Control Strategies for wastewater treatment plants (WWTPs) is coming to an end, it is essential to disseminate the knowledge gained. For this reason, all authors of the IWA Scientific and Technical Report on benchmarking have come together to p...

  8. EPA's Benchmark Dose Modeling Software

    Science.gov (United States)

    The EPA developed the Benchmark Dose Software (BMDS) as a tool to help Agency risk assessors facilitate applying benchmark dose (BMD) method’s to EPA’s human health risk assessment (HHRA) documents. The application of BMD methods overcomes many well know limitations ...

  9. Benchmark for Strategic Performance Improvement.

    Science.gov (United States)

    Gohlke, Annette

    1997-01-01

    Explains benchmarking, a total quality management tool used to measure and compare the work processes in a library with those in other libraries to increase library performance. Topics include the main groups of upper management, clients, and staff; critical success factors for each group; and benefits of benchmarking. (Author/LRW)

  10. Benchmarking: A Process for Improvement.

    Science.gov (United States)

    Peischl, Thomas M.

    One problem with the outcome-based measures used in higher education is that they measure quantity but not quality. Benchmarking, or the use of some external standard of quality to measure tasks, processes, and outputs, is partially solving that difficulty. Benchmarking allows for the establishment of a systematic process to indicate if outputs…

  11. Benchmark job – Watch out!

    CERN Multimedia

    Staff Association

    2017-01-01

    On 12 December 2016, in Echo No. 259, we already discussed at length the MERIT and benchmark jobs. Still, we find that a couple of issues warrant further discussion. Benchmark job – administrative decision on 1 July 2017 On 12 January 2017, the HR Department informed all staff members of a change to the effective date of the administrative decision regarding benchmark jobs. The benchmark job title of each staff member will be confirmed on 1 July 2017, instead of 1 May 2017 as originally announced in HR’s letter on 18 August 2016. Postponing the administrative decision by two months will leave a little more time to address the issues related to incorrect placement in a benchmark job. Benchmark job – discuss with your supervisor, at the latest during the MERIT interview In order to rectify an incorrect placement in a benchmark job, it is essential that the supervisor and the supervisee go over the assigned benchmark job together. In most cases, this placement has been done autom...

  12. Calculation of the fifth atomic energy research dynamic benchmark with APROS

    International Nuclear Information System (INIS)

    Puska Eija Karita; Kontio Harii

    1998-01-01

    The band-out presents the model used for calculation of the fifth atomic energy research dynamic benchmark with APROS code. In the calculation of the fifth atomic energy research dynamic benchmark the three-dimensional neutronics model of APROS was used. The core was divided axially into 20 nodes according to the specifications of the benchmark and each six identical fuel assemblies were placed into one one-dimensional thermal hydraulic channel. The five-equation thermal hydraulic model was used in the benchmark. The plant process and automation was described with a generic WWER-440 plant model created by IVO Power Engineering Ltd. - Finland. (Author)

  13. On singularities of lattice varieties

    OpenAIRE

    Mukherjee, Himadri

    2013-01-01

    Toric varieties associated with distributive lattices arise as a fibre of a flat degeneration of a Schubert variety in a minuscule. The singular locus of these varieties has been studied by various authors. In this article we prove that the number of diamonds incident on a lattice point $\\a$ in a product of chain lattices is more than or equal to the codimension of the lattice. Using this we also show that the lattice varieties associated with product of chain lattices is smooth.

  14. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-15

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology.

  15. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-01

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology

  16. Benchmarking: applications to transfusion medicine.

    Science.gov (United States)

    Apelseth, Torunn Oveland; Molnar, Laura; Arnold, Emmy; Heddle, Nancy M

    2012-10-01

    Benchmarking is as a structured continuous collaborative process in which comparisons for selected indicators are used to identify factors that, when implemented, will improve transfusion practices. This study aimed to identify transfusion medicine studies reporting on benchmarking, summarize the benchmarking approaches used, and identify important considerations to move the concept of benchmarking forward in the field of transfusion medicine. A systematic review of published literature was performed to identify transfusion medicine-related studies that compared at least 2 separate institutions or regions with the intention of benchmarking focusing on 4 areas: blood utilization, safety, operational aspects, and blood donation. Forty-five studies were included: blood utilization (n = 35), safety (n = 5), operational aspects of transfusion medicine (n = 5), and blood donation (n = 0). Based on predefined criteria, 7 publications were classified as benchmarking, 2 as trending, and 36 as single-event studies. Three models of benchmarking are described: (1) a regional benchmarking program that collects and links relevant data from existing electronic sources, (2) a sentinel site model where data from a limited number of sites are collected, and (3) an institutional-initiated model where a site identifies indicators of interest and approaches other institutions. Benchmarking approaches are needed in the field of transfusion medicine. Major challenges include defining best practices and developing cost-effective methods of data collection. For those interested in initiating a benchmarking program, the sentinel site model may be most effective and sustainable as a starting point, although the regional model would be the ideal goal. Copyright © 2012 Elsevier Inc. All rights reserved.

  17. Towards a physical interpretation of the entropic lattice Boltzmann method

    Science.gov (United States)

    Malaspinas, Orestis; Deville, Michel; Chopard, Bastien

    2008-12-01

    The entropic lattice Boltzmann method (ELBM) is one among several different versions of the lattice Boltzmann method for the simulation of hydrodynamics. The collision term of the ELBM is characterized by a nonincreasing H function, guaranteed by a variable relaxation time. We propose here an analysis of the ELBM using the Chapman-Enskog expansion. We show that it can be interpreted as some kind of subgrid model, where viscosity correction scales like the strain rate tensor. We confirm our analytical results by the numerical computations of the relaxation time modifications on the two-dimensional dipole-wall interaction benchmark.

  18. MoMaS reactive transport benchmark using PFLOTRAN

    Science.gov (United States)

    Park, H.

    2017-12-01

    MoMaS benchmark was developed to enhance numerical simulation capability for reactive transport modeling in porous media. The benchmark was published in late September of 2009; it is not taken from a real chemical system, but realistic and numerically challenging tests. PFLOTRAN is a state-of-art massively parallel subsurface flow and reactive transport code that is being used in multiple nuclear waste repository projects at Sandia National Laboratories including Waste Isolation Pilot Plant and Used Fuel Disposition. MoMaS benchmark has three independent tests with easy, medium, and hard chemical complexity. This paper demonstrates how PFLOTRAN is applied to this benchmark exercise and shows results of the easy benchmark test case which includes mixing of aqueous components and surface complexation. Surface complexations consist of monodentate and bidentate reactions which introduces difficulty in defining selectivity coefficient if the reaction applies to a bulk reference volume. The selectivity coefficient becomes porosity dependent for bidentate reaction in heterogeneous porous media. The benchmark is solved by PFLOTRAN with minimal modification to address the issue and unit conversions were made properly to suit PFLOTRAN.

  19. Benchmarking to improve the quality of cystic fibrosis care.

    Science.gov (United States)

    Schechter, Michael S

    2012-11-01

    Benchmarking involves the ascertainment of healthcare programs with most favorable outcomes as a means to identify and spread effective strategies for delivery of care. The recent interest in the development of patient registries for patients with cystic fibrosis (CF) has been fueled in part by an interest in using them to facilitate benchmarking. This review summarizes reports of how benchmarking has been operationalized in attempts to improve CF care. Although certain goals of benchmarking can be accomplished with an exclusive focus on registry data analysis, benchmarking programs in Germany and the United States have supplemented these data analyses with exploratory interactions and discussions to better understand successful approaches to care and encourage their spread throughout the care network. Benchmarking allows the discovery and facilitates the spread of effective approaches to care. It provides a pragmatic alternative to traditional research methods such as randomized controlled trials, providing insights into methods that optimize delivery of care and allowing judgments about the relative effectiveness of different therapeutic approaches.

  20. Simulation of turbulent flow over staggered tube bundles using multi-relaxation time lattice Boltzmann method

    International Nuclear Information System (INIS)

    Park, Jong Woon; Choi, Hyun Gyung

    2014-01-01

    A turbulent fluid flow over staggered tube bundles is of great interest in many engineering fields including nuclear fuel rods, heat exchangers and especially a gas cooled reactor lower plenum. Computational methods have evolved for the simulation of such flow for decades and lattice Boltzmann method (LBM) is one of the attractive methods due to its sound physical basis and ease of computerization including parallelization. In this study to find computational performance of the LBM in turbulent flows over staggered tubes, a fluid flow analysis code employing multi-relaxation time lattice Boltzmann method (MRT-LBM) is developed based on a 2-dimensional D2Q9 lattice model and classical sub-grid eddy viscosity model of Smagorinsky. As a first step, fundamental performance MRT-LBM is investigated against a standard problem of a flow past a cylinder at low Reynolds number in terms of drag forces. As a major step, benchmarking of the MRT-LBM is performed over a turbulent flow through staggered tube bundles at Reynolds number of 18,000. For a flow past a single cylinder, the accuracy is validated against existing experimental data and previous computations in terms of drag forces on the cylinder. Mainly, the MRT-LBM computation for a flow through staggered tube bundles is performed and compared with experimental data and general purpose computational fluid dynamic (CFD) analyses with standard k-ω turbulence and large eddy simulation (LES) equipped with turbulence closures of Smagrinsky-Lilly and wall-adapting local eddy-viscosity (WALE) model. The agreement between the experimental and the computational results from the present MRT-LBM is found to be reasonably acceptable and even comparable to the LES whereas the computational efficiency is superior. (orig.)

  1. Simulation of turbulent flow over staggered tube bundles using multi-relaxation time lattice Boltzmann method

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Woon; Choi, Hyun Gyung [Dongguk Univ., Gyeongju (Korea, Republic of). Nuclear and Energy Engineering Dept.

    2014-02-15

    A turbulent fluid flow over staggered tube bundles is of great interest in many engineering fields including nuclear fuel rods, heat exchangers and especially a gas cooled reactor lower plenum. Computational methods have evolved for the simulation of such flow for decades and lattice Boltzmann method (LBM) is one of the attractive methods due to its sound physical basis and ease of computerization including parallelization. In this study to find computational performance of the LBM in turbulent flows over staggered tubes, a fluid flow analysis code employing multi-relaxation time lattice Boltzmann method (MRT-LBM) is developed based on a 2-dimensional D2Q9 lattice model and classical sub-grid eddy viscosity model of Smagorinsky. As a first step, fundamental performance MRT-LBM is investigated against a standard problem of a flow past a cylinder at low Reynolds number in terms of drag forces. As a major step, benchmarking of the MRT-LBM is performed over a turbulent flow through staggered tube bundles at Reynolds number of 18,000. For a flow past a single cylinder, the accuracy is validated against existing experimental data and previous computations in terms of drag forces on the cylinder. Mainly, the MRT-LBM computation for a flow through staggered tube bundles is performed and compared with experimental data and general purpose computational fluid dynamic (CFD) analyses with standard k-ω turbulence and large eddy simulation (LES) equipped with turbulence closures of Smagrinsky-Lilly and wall-adapting local eddy-viscosity (WALE) model. The agreement between the experimental and the computational results from the present MRT-LBM is found to be reasonably acceptable and even comparable to the LES whereas the computational efficiency is superior. (orig.)

  2. Benchmarking school nursing practice: the North West Regional Benchmarking Group

    OpenAIRE

    Littler, Nadine; Mullen, Margaret; Beckett, Helen; Freshney, Alice; Pinder, Lynn

    2016-01-01

    It is essential that the quality of care is reviewed regularly through robust processes such as benchmarking to ensure all outcomes and resources are evidence-based so that children and young people’s needs are met effectively. This article provides an example of the use of benchmarking in school nursing practice. Benchmarking has been defined as a process for finding, adapting and applying best practices (Camp, 1994). This concept was first adopted in the 1970s ‘from industry where it was us...

  3. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kahler, A.C.; Herman, M.; Kahler,A.C.; MacFarlane,R.E.; Mosteller,R.D.; Kiedrowski,B.C.; Frankle,S.C.; Chadwick,M.B.; McKnight,R.D.; Lell,R.M.; Palmiotti,G.; Hiruta,H.; Herman,M.; Arcilla,R.; Mughabghab,S.F.; Sublet,J.C.; Trkov,A.; Trumbull,T.H.; Dunn,M.

    2011-12-01

    The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also

  4. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kahler, A. [Los Alamos National Laboratory (LANL); Macfarlane, R E [Los Alamos National Laboratory (LANL); Mosteller, R D [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Frankle, S C [Los Alamos National Laboratory (LANL); Chadwick, M. B. [Los Alamos National Laboratory (LANL); Mcknight, R D [Argonne National Laboratory (ANL); Lell, R M [Argonne National Laboratory (ANL); Palmiotti, G [Idaho National Laboratory (INL); Hiruta, h [Idaho National Laboratory (INL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Arcilla, r [Brookhaven National Laboratory (BNL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Sublet, J C [Culham Science Center, Abington, UK; Trkov, A. [Jozef Stefan Institute, Slovenia; Trumbull, T H [Knolls Atomic Power Laboratory; Dunn, Michael E [ORNL

    2011-01-01

    The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unrnoderated and uranium reflected (235)U and (239)Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as (236)U; (238,242)Pu and (241,243)Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical

  5. Benchmarking Nuclear Power Plants

    International Nuclear Information System (INIS)

    Jakic, I.

    2016-01-01

    One of the main tasks an owner have is to keep its business competitive on the market while delivering its product. Being owner of nuclear power plant bear the same (or even more complex and stern) responsibility due to safety risks and costs. In the past, nuclear power plant managements could (partly) ignore profit or it was simply expected and to some degree assured through the various regulatory processes governing electricity rate design. It is obvious now that, with the deregulation, utility privatization and competitive electricity market, key measure of success used at nuclear power plants must include traditional metrics of successful business (return on investment, earnings and revenue generation) as well as those of plant performance, safety and reliability. In order to analyze business performance of (specific) nuclear power plant, benchmarking, as one of the well-established concept and usual method was used. Domain was conservatively designed, with well-adjusted framework, but results have still limited application due to many differences, gaps and uncertainties. (author).

  6. Virtual machine performance benchmarking.

    Science.gov (United States)

    Langer, Steve G; French, Todd

    2011-10-01

    The attractions of virtual computing are many: reduced costs, reduced resources and simplified maintenance. Any one of these would be compelling for a medical imaging professional attempting to support a complex practice on limited resources in an era of ever tightened reimbursement. In particular, the ability to run multiple operating systems optimized for different tasks (computational image processing on Linux versus office tasks on Microsoft operating systems) on a single physical machine is compelling. However, there are also potential drawbacks. High performance requirements need to be carefully considered if they are to be executed in an environment where the running software has to execute through multiple layers of device drivers before reaching the real disk or network interface. Our lab has attempted to gain insight into the impact of virtualization on performance by benchmarking the following metrics on both physical and virtual platforms: local memory and disk bandwidth, network bandwidth, and integer and floating point performance. The virtual performance metrics are compared to baseline performance on "bare metal." The results are complex, and indeed somewhat surprising.

  7. Benchmarking in academic pharmacy departments.

    Science.gov (United States)

    Bosso, John A; Chisholm-Burns, Marie; Nappi, Jean; Gubbins, Paul O; Ross, Leigh Ann

    2010-10-11

    Benchmarking in academic pharmacy, and recommendations for the potential uses of benchmarking in academic pharmacy departments are discussed in this paper. Benchmarking is the process by which practices, procedures, and performance metrics are compared to an established standard or best practice. Many businesses and industries use benchmarking to compare processes and outcomes, and ultimately plan for improvement. Institutions of higher learning have embraced benchmarking practices to facilitate measuring the quality of their educational and research programs. Benchmarking is used internally as well to justify the allocation of institutional resources or to mediate among competing demands for additional program staff or space. Surveying all chairs of academic pharmacy departments to explore benchmarking issues such as department size and composition, as well as faculty teaching, scholarly, and service productivity, could provide valuable information. To date, attempts to gather this data have had limited success. We believe this information is potentially important, urge that efforts to gather it should be continued, and offer suggestions to achieve full participation.

  8. The spent fuel safety experiment

    International Nuclear Information System (INIS)

    Harmms, G.A.; Davis, F.J.; Ford, J.T.

    1995-01-01

    The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

  9. Criticality Analysis Of TCA Critical Lattices With MNCP-4C Monte Carlo Calculation

    International Nuclear Information System (INIS)

    Zuhair

    2002-01-01

    The use of uranium-plutonium mixed oxide (MOX) fuel in electric generation light water reactor (PWR, BWR) is being planned in Japan. Therefore, the accuracy evaluations of neutronic analysis code for MOX cores have been employed by many scientists and reactor physicists. Benchmark evaluations for TCA was done using various calculation methods. The Monte Carlo become the most reliable method to predict criticality of various reactor types. In this analysis, the MCNP-4C code was chosen because various superiorities the code has. All in all, the MCNP-4C calculation for TCA core with 38 MOX critical lattice configurations gave the results with high accuracy. The JENDL-3.2 library showed significantly closer results to the ENDF/B-V. The k eff values calculated with the ENDF/B-VI library gave underestimated results. The ENDF/B-V library gave the best estimation. It can be concluded that MCNP-4C calculation, especially with ENDF/B-V and JENDL-3.2 libraries, for MOX fuel utilized NPP design in reactor core is the best choice

  10. Development of common user data model for APOLLO3 and MARBLE and application to benchmark problems

    International Nuclear Information System (INIS)

    Yokoyama, Kenji

    2009-07-01

    A Common User Data Model, CUDM, has been developed for the purpose of benchmark calculations between APOLLO3 and MARBLE code systems. The current version of CUDM was designed for core calculation benchmark problems with 3-dimensional Cartesian, 3-D XYZ, geometry. CUDM is able to manage all input/output data such as 3-D XYZ geometry, effective macroscopic cross section, effective multiplication factor and neutron flux. In addition, visualization tools for geometry and neutron flux were included. CUDM was designed by the object-oriented technique and implemented using Python programming language. Based on the CUDM, a prototype system for a benchmark calculation, CUDM-benchmark, was also developed. The CUDM-benchmark supports input/output data conversion for IDT solver in APOLLO3, and TRITAC and SNT solvers in MARBLE. In order to evaluate pertinence of CUDM, the CUDM-benchmark was applied to benchmark problems proposed by T. Takeda, G. Chiba and I. Zmijarevic. It was verified that the CUDM-benchmark successfully reproduced the results calculated with reference input data files, and provided consistent results among all the solvers by using one common input data defined by CUDM. In addition, a detailed benchmark calculation for Chiba benchmark was performed by using the CUDM-benchmark. Chiba benchmark is a neutron transport benchmark problem for fast criticality assembly without homogenization. This benchmark problem consists of 4 core configurations which have different sodium void regions, and each core configuration is defined by more than 5,000 fuel/material cells. In this application, it was found that the results by IDT and SNT solvers agreed well with the reference results by Monte-Carlo code. In addition, model effects such as quadrature set effect, S n order effect and mesh size effect were systematically evaluated and summarized in this report. (author)

  11. PHISICS/RELAP5-3D RESULTS FOR EXERCISES II-1 AND II-2 OF THE OECD/NEA MHTGR-350 BENCHMARK

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Laboratory

    2016-03-01

    The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) High-Temperature Gas-Cooled Reactor (HTGR) Methods group currently leads the Modular High-Temperature Gas-Cooled Reactor (MHTGR) 350 benchmark. The benchmark consists of a set of lattice-depletion, steady-state, and transient problems that can be used by HTGR simulation groups to assess the performance of their code suites. The paper summarizes the results obtained for the first two transient exercises defined for Phase II of the benchmark. The Parallel and Highly Innovative Simulation for INL Code System (PHISICS), coupled with the INL system code RELAP5-3D, was used to generate the results for the Depressurized Conduction Cooldown (DCC) (exercise II-1a) and Pressurized Conduction Cooldown (PCC) (exercise II-2) transients. These exercises require the time-dependent simulation of coupled neutronics and thermal-hydraulics phenomena, and utilize the steady-state solution previously obtained for exercise I-3 of Phase I. This paper also includes a comparison of the benchmark results obtained with a traditional system code “ring” model against a more detailed “block” model that include kinetics feedback on an individual block level and thermal feedbacks on a triangular sub-mesh. The higher spatial fidelity that can be obtained by the block model is illustrated with comparisons of the maximum fuel temperatures, especially in the case of natural convection conditions that dominate the DCC and PCC events. Differences up to 125 K (or 10%) were observed between the ring and block model predictions of the DCC transient, mostly due to the block model’s capability of tracking individual block decay powers and more detailed helium flow distributions. In general, the block model only required DCC and PCC calculation times twice as long as the ring models, and it therefore seems that the additional development and calculation time required for the block model could be worth the gain that can be

  12. Issues in Benchmark Metric Selection

    Science.gov (United States)

    Crolotte, Alain

    It is true that a metric can influence a benchmark but will esoteric metrics create more problems than they will solve? We answer this question affirmatively by examining the case of the TPC-D metric which used the much debated geometric mean for the single-stream test. We will show how a simple choice influenced the benchmark and its conduct and, to some extent, DBMS development. After examining other alternatives our conclusion is that the “real” measure for a decision-support benchmark is the arithmetic mean.

  13. California commercial building energy benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Kinney, Satkartar; Piette, Mary Ann

    2003-07-01

    Building energy benchmarking is the comparison of whole-building energy use relative to a set of similar buildings. It provides a useful starting point for individual energy audits and for targeting buildings for energy-saving measures in multiple-site audits. Benchmarking is of interest and practical use to a number of groups. Energy service companies and performance contractors communicate energy savings potential with ''typical'' and ''best-practice'' benchmarks while control companies and utilities can provide direct tracking of energy use and combine data from multiple buildings. Benchmarking is also useful in the design stage of a new building or retrofit to determine if a design is relatively efficient. Energy managers and building owners have an ongoing interest in comparing energy performance to others. Large corporations, schools, and government agencies with numerous facilities also use benchmarking methods to compare their buildings to each other. The primary goal of Task 2.1.1 Web-based Benchmarking was the development of a web-based benchmarking tool, dubbed Cal-Arch, for benchmarking energy use in California commercial buildings. While there were several other benchmarking tools available to California consumers prior to the development of Cal-Arch, there were none that were based solely on California data. Most available benchmarking information, including the Energy Star performance rating, were developed using DOE's Commercial Building Energy Consumption Survey (CBECS), which does not provide state-level data. Each database and tool has advantages as well as limitations, such as the number of buildings and the coverage by type, climate regions and end uses. There is considerable commercial interest in benchmarking because it provides an inexpensive method of screening buildings for tune-ups and retrofits. However, private companies who collect and manage consumption data are concerned that the

  14. A Heterogeneous Medium Analytical Benchmark

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    1999-01-01

    A benchmark, called benchmark BLUE, has been developed for one-group neutral particle (neutron or photon) transport in a one-dimensional sub-critical heterogeneous plane parallel medium with surface illumination. General anisotropic scattering is accommodated through the Green's Function Method (GFM). Numerical Fourier transform inversion is used to generate the required Green's functions which are kernels to coupled integral equations that give the exiting angular fluxes. The interior scalar flux is then obtained through quadrature. A compound iterative procedure for quadrature order and slab surface source convergence provides highly accurate benchmark qualities (4- to 5- places of accuracy) results

  15. MEETING: Lattice 88

    International Nuclear Information System (INIS)

    Mackenzie, Paul

    1989-01-01

    The forty-year dream of understanding the properties of the strongly interacting particles from first principles is now approaching reality. Quantum chromodynamics (QCD - the field theory of the quark and gluon constituents of strongly interacting particles) was initially handicapped by the severe limitations of the conventional (perturbation) approach in this picture, but Ken Wilson's inventions of lattice gauge theory and renormalization group methods opened new doors, making calculations of masses and other particle properties possible. Lattice gauge theory became a major industry around 1980, when Monte Carlo methods were introduced, and the first prototype calculations yielded qualitatively reasonable results. The promising developments over the past year were highlighted at the 1988 Symposium on Lattice Field Theory - Lattice 88 - held at Fermilab

  16. Angles in hyperbolic lattices

    DEFF Research Database (Denmark)

    Risager, Morten S.; Södergren, Carl Anders

    2017-01-01

    It is well known that the angles in a lattice acting on hyperbolic n -space become equidistributed. In this paper we determine a formula for the pair correlation density for angles in such hyperbolic lattices. Using this formula we determine, among other things, the asymptotic behavior of the den......It is well known that the angles in a lattice acting on hyperbolic n -space become equidistributed. In this paper we determine a formula for the pair correlation density for angles in such hyperbolic lattices. Using this formula we determine, among other things, the asymptotic behavior...... of the density function in both the small and large variable limits. This extends earlier results by Boca, Pasol, Popa and Zaharescu and Kelmer and Kontorovich in dimension 2 to general dimension n . Our proofs use the decay of matrix coefficients together with a number of careful estimates, and lead...

  17. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1999-01-01

    The present lecture has a main goal to show how the transport lattice calculations are realised in a standard computer code. This is illustrated on the example of the WIMSD code, belonging to the most popular tools for reactor calculations. Most of the approaches discussed here can be easily modified to any other lattice code. The description of the code assumes the basic knowledge of reactor lattice, on the level given in the lecture on 'Reactor lattice transport calculations'. For more advanced explanation of the WIMSD code the reader is directed to the detailed descriptions of the code cited in References. The discussion of the methods and models included in the code is followed by the generally used homogenisation procedure and several numerical examples of discrepancies in calculated multiplication factors based on different sources of library data. (author)

  18. MEETING: Lattice 88

    Energy Technology Data Exchange (ETDEWEB)

    Mackenzie, Paul

    1989-03-15

    The forty-year dream of understanding the properties of the strongly interacting particles from first principles is now approaching reality. Quantum chromodynamics (QCD - the field theory of the quark and gluon constituents of strongly interacting particles) was initially handicapped by the severe limitations of the conventional (perturbation) approach in this picture, but Ken Wilson's inventions of lattice gauge theory and renormalization group methods opened new doors, making calculations of masses and other particle properties possible. Lattice gauge theory became a major industry around 1980, when Monte Carlo methods were introduced, and the first prototype calculations yielded qualitatively reasonable results. The promising developments over the past year were highlighted at the 1988 Symposium on Lattice Field Theory - Lattice 88 - held at Fermilab.

  19. Computers for Lattice QCD

    International Nuclear Information System (INIS)

    Christ, Norman H

    2000-01-01

    The architecture and capabilities of the computers currently in use for large-scale lattice QCD calculations are described and compared. Based on this present experience, possible future directions are discussed

  20. A Global Vision over Benchmarking Process: Benchmarking Based Enterprises

    OpenAIRE

    Sitnikov, Catalina; Giurca Vasilescu, Laura

    2008-01-01

    Benchmarking uses the knowledge and the experience of others to improve the enterprise. Starting from the analysis of the performance and underlying the strengths and weaknesses of the enterprise it should be assessed what must be done in order to improve its activity. Using benchmarking techniques, an enterprise looks at how processes in the value chain are performed. The approach based on the vision “from the whole towards the parts” (a fragmented image of the enterprise’s value chain) redu...

  1. Benchmarking and Learning in Public Healthcare

    DEFF Research Database (Denmark)

    Buckmaster, Natalie; Mouritsen, Jan

    2017-01-01

    This research investigates the effects of learning-oriented benchmarking in public healthcare settings. Benchmarking is a widely adopted yet little explored accounting practice that is part of the paradigm of New Public Management. Extant studies are directed towards mandated coercive benchmarking...... applications. The present study analyses voluntary benchmarking in a public setting that is oriented towards learning. The study contributes by showing how benchmarking can be mobilised for learning and offers evidence of the effects of such benchmarking for performance outcomes. It concludes that benchmarking...... can enable learning in public settings but that this requires actors to invest in ensuring that benchmark data are directed towards improvement....

  2. C5 Benchmark Problem with Discrete Ordinate Radiation Transport Code DENOVO

    Energy Technology Data Exchange (ETDEWEB)

    Yesilyurt, Gokhan [ORNL; Clarno, Kevin T [ORNL; Evans, Thomas M [ORNL; Davidson, Gregory G [ORNL; Fox, Patricia B [ORNL

    2011-01-01

    The C5 benchmark problem proposed by the Organisation for Economic Co-operation and Development/Nuclear Energy Agency was modeled to examine the capabilities of Denovo, a three-dimensional (3-D) parallel discrete ordinates (S{sub N}) radiation transport code, for problems with no spatial homogenization. Denovo uses state-of-the-art numerical methods to obtain accurate solutions to the Boltzmann transport equation. Problems were run in parallel on Jaguar, a high-performance supercomputer located at Oak Ridge National Laboratory. Both the two-dimensional (2-D) and 3-D configurations were analyzed, and the results were compared with the reference MCNP Monte Carlo calculations. For an additional comparison, SCALE/KENO-V.a Monte Carlo solutions were also included. In addition, a sensitivity analysis was performed for the optimal angular quadrature and mesh resolution for both the 2-D and 3-D infinite lattices of UO{sub 2} fuel pin cells. Denovo was verified with the C5 problem. The effective multiplication factors, pin powers, and assembly powers were found to be in good agreement with the reference MCNP and SCALE/KENO-V.a Monte Carlo calculations.

  3. Lattice gauge theories

    International Nuclear Information System (INIS)

    Petronzio, R.

    1992-01-01

    Lattice gauge theories are about fifteen years old and I will report on the present status of the field without making the elementary introduction that can be found in the proceedings of the last two conferences. The talk covers briefly the following subjects: the determination of α s , the status of spectroscopy, heavy quark physics and in particular the calculation of their hadronic weak matrix elements, high temperature QCD, non perturbative Higgs bounds, chiral theories on the lattice and induced theories

  4. Permutohedral Lattice CNNs

    OpenAIRE

    Kiefel, Martin; Jampani, Varun; Gehler, Peter V.

    2014-01-01

    This paper presents a convolutional layer that is able to process sparse input features. As an example, for image recognition problems this allows an efficient filtering of signals that do not lie on a dense grid (like pixel position), but of more general features (such as color values). The presented algorithm makes use of the permutohedral lattice data structure. The permutohedral lattice was introduced to efficiently implement a bilateral filter, a commonly used image processing operation....

  5. Additive lattice kirigami.

    Science.gov (United States)

    Castle, Toen; Sussman, Daniel M; Tanis, Michael; Kamien, Randall D

    2016-09-01

    Kirigami uses bending, folding, cutting, and pasting to create complex three-dimensional (3D) structures from a flat sheet. In the case of lattice kirigami, this cutting and rejoining introduces defects into an underlying 2D lattice in the form of points of nonzero Gaussian curvature. A set of simple rules was previously used to generate a wide variety of stepped structures; we now pare back these rules to their minimum. This allows us to describe a set of techniques that unify a wide variety of cut-and-paste actions under the rubric of lattice kirigami, including adding new material and rejoining material across arbitrary cuts in the sheet. We also explore the use of more complex lattices and the different structures that consequently arise. Regardless of the choice of lattice, creating complex structures may require multiple overlapping kirigami cuts, where subsequent cuts are not performed on a locally flat lattice. Our additive kirigami method describes such cuts, providing a simple methodology and a set of techniques to build a huge variety of complex 3D shapes.

  6. Classical Logic and Quantum Logic with Multiple and Common Lattice Models

    Directory of Open Access Journals (Sweden)

    Mladen Pavičić

    2016-01-01

    Full Text Available We consider a proper propositional quantum logic and show that it has multiple disjoint lattice models, only one of which is an orthomodular lattice (algebra underlying Hilbert (quantum space. We give an equivalent proof for the classical logic which turns out to have disjoint distributive and nondistributive ortholattices. In particular, we prove that both classical logic and quantum logic are sound and complete with respect to each of these lattices. We also show that there is one common nonorthomodular lattice that is a model of both quantum and classical logic. In technical terms, that enables us to run the same classical logic on both a digital (standard, two-subset, 0-1-bit computer and a nondigital (say, a six-subset computer (with appropriate chips and circuits. With quantum logic, the same six-element common lattice can serve us as a benchmark for an efficient evaluation of equations of bigger lattice models or theorems of the logic.

  7. Lattice regularized chiral perturbation theory

    International Nuclear Information System (INIS)

    Borasoy, Bugra; Lewis, Randy; Ouimet, Pierre-Philippe A.

    2004-01-01

    Chiral perturbation theory can be defined and regularized on a spacetime lattice. A few motivations are discussed here, and an explicit lattice Lagrangian is reviewed. A particular aspect of the connection between lattice chiral perturbation theory and lattice QCD is explored through a study of the Wess-Zumino-Witten term

  8. Vortex lattices in layered superconductors

    International Nuclear Information System (INIS)

    Prokic, V.; Davidovic, D.; Dobrosavljevic-Grujic, L.

    1995-01-01

    We study vortex lattices in a superconductor--normal-metal superlattice in a parallel magnetic field. Distorted lattices, resulting from the shear deformations along the layers, are found to be unstable. Under field variation, nonequilibrium configurations undergo an infinite sequence of continuous transitions, typical for soft lattices. The equilibrium vortex arrangement is always a lattice of isocell triangles, without shear

  9. Performance Targets and External Benchmarking

    DEFF Research Database (Denmark)

    Friis, Ivar; Hansen, Allan; Vámosi, Tamás S.

    Research on relative performance measures, transfer pricing, beyond budgeting initiatives, target costing, piece rates systems and value based management has for decades underlined the importance of external benchmarking in performance management. Research conceptualises external benchmarking...... as a market mechanism that can be brought inside the firm to provide incentives for continuous improvement and the development of competitive advances. However, whereas extant research primarily has focused on the importance and effects of using external benchmarks, less attention has been directed towards...... the conditions upon which the market mechanism is performing within organizations. This paper aims to contribute to research by providing more insight to the conditions for the use of external benchmarking as an element in performance management in organizations. Our study explores a particular type of external...

  10. Benchmarking and Sustainable Transport Policy

    DEFF Research Database (Denmark)

    Gudmundsson, Henrik; Wyatt, Andrew; Gordon, Lucy

    2004-01-01

    Order to learn from the best. In 2000 the European Commission initiated research to explore benchmarking as a tool to promote policies for ‘sustainable transport’. This paper reports findings and recommendations on how to address this challenge. The findings suggest that benchmarking is a valuable...... tool that may indeed help to move forward the transport policy agenda. However, there are major conditions and limitations. First of all it is not always so straightforward to delimit, measure and compare transport services in order to establish a clear benchmark. Secondly ‘sustainable transport......’ evokes a broad range of concerns that are hard to address fully at the level of specific practices. Thirdly policies are not directly comparable across space and context. For these reasons attempting to benchmark ‘sustainable transport policies’ against one another would be a highly complex task, which...

  11. Thermal reactor benchmark testing of 69 group library

    International Nuclear Information System (INIS)

    Liu Guisheng; Wang Yaoqing; Liu Ping; Zhang Baocheng

    1994-01-01

    Using a code system NSLINK, AMPX master library in WIMS 69 groups structure are made from nuclides relating to 4 newest evaluated nuclear data libraries. Some integrals of 10 thermal reactor benchmark assemblies recommended by the U.S. CSEWG are calculated using rectified PASC-1 code system and compared with foreign results, the authors results are in good agreement with others. 69 group libraries of evaluated data bases in TPFAP interface file are generated with NJOY code system. The k ∞ values of 6 cell lattice assemblies are calculated by the code CBM. The calculated results are analysed and compared

  12. Benchmarking: contexts and details matter.

    Science.gov (United States)

    Zheng, Siyuan

    2017-07-05

    Benchmarking is an essential step in the development of computational tools. We take this opportunity to pitch in our opinions on tool benchmarking, in light of two correspondence articles published in Genome Biology.Please see related Li et al. and Newman et al. correspondence articles: www.dx.doi.org/10.1186/s13059-017-1256-5 and www.dx.doi.org/10.1186/s13059-017-1257-4.

  13. Handbook of critical experiments benchmarks

    International Nuclear Information System (INIS)

    Durst, B.M.; Bierman, S.R.; Clayton, E.D.

    1978-03-01

    Data from critical experiments have been collected together for use as benchmarks in evaluating calculational techniques and nuclear data. These benchmarks have been selected from the numerous experiments performed on homogeneous plutonium systems. No attempt has been made to reproduce all of the data that exists. The primary objective in the collection of these data is to present representative experimental data defined in a concise, standardized format that can easily be translated into computer code input

  14. Analysis of Benchmark 2 results

    International Nuclear Information System (INIS)

    Bacha, F.; Lefievre, B.; Maillard, J.; Silva, J.

    1994-01-01

    The code GEANT315 has been compared to different codes in two benchmarks. We analyze its performances through our results, especially in the thick target case. In spite of gaps in nucleus-nucleus interaction theories at intermediate energies, benchmarks allow possible improvements of physical models used in our codes. Thereafter, a scheme of radioactive waste burning system is studied. (authors). 4 refs., 7 figs., 1 tab

  15. Benchmarks for GADRAS performance validation

    International Nuclear Information System (INIS)

    Mattingly, John K.; Mitchell, Dean James; Rhykerd, Charles L. Jr.

    2009-01-01

    The performance of the Gamma Detector Response and Analysis Software (GADRAS) was validated by comparing GADRAS model results to experimental measurements for a series of benchmark sources. Sources for the benchmark include a plutonium metal sphere, bare and shielded in polyethylene, plutonium oxide in cans, a highly enriched uranium sphere, bare and shielded in polyethylene, a depleted uranium shell and spheres, and a natural uranium sphere. The benchmark experimental data were previously acquired and consist of careful collection of background and calibration source spectra along with the source spectra. The calibration data were fit with GADRAS to determine response functions for the detector in each experiment. A one-dimensional model (pie chart) was constructed for each source based on the dimensions of the benchmark source. The GADRAS code made a forward calculation from each model to predict the radiation spectrum for the detector used in the benchmark experiment. The comparisons between the GADRAS calculation and the experimental measurements are excellent, validating that GADRAS can correctly predict the radiation spectra for these well-defined benchmark sources.

  16. Benchmarking in Czech Higher Education

    Directory of Open Access Journals (Sweden)

    Plaček Michal

    2015-12-01

    Full Text Available The first part of this article surveys the current experience with the use of benchmarking at Czech universities specializing in economics and management. The results indicate that collaborative benchmarking is not used on this level today, but most actors show some interest in its introduction. The expression of the need for it and the importance of benchmarking as a very suitable performance-management tool in less developed countries are the impetus for the second part of our article. Based on an analysis of the current situation and existing needs in the Czech Republic, as well as on a comparison with international experience, recommendations for public policy are made, which lie in the design of a model of a collaborative benchmarking for Czech economics and management in higher-education programs. Because the fully complex model cannot be implemented immediately – which is also confirmed by structured interviews with academics who have practical experience with benchmarking –, the final model is designed as a multi-stage model. This approach helps eliminate major barriers to the implementation of benchmarking.

  17. Parameters calculation of fuel assembly with complex geometry

    International Nuclear Information System (INIS)

    Wu Hongchun; Ju Haitao; Yao Dong

    2006-01-01

    The code DRAGON was developed for CANDU reactor by Ecole Polytechnique de Montreal of Canada. In order to validate the DRAGON code's applicability for complex geometry fuel assembly calculation, the rod shape fuel assembly of PWR benchmark problem and the plate shape fuel assembly of MTR benchmark problem were analyzed by DRAGON code. Some other shape fuel assemblies were also discussed simply. Calculation results show that the DRAGON code can be used to calculate variform fuel assembly and the precision is high. (authors)

  18. Dynamic benchmarking of simulation codes

    International Nuclear Information System (INIS)

    Henry, R.E.; Paik, C.Y.; Hauser, G.M.

    1996-01-01

    Computer simulation of nuclear power plant response can be a full-scope control room simulator, an engineering simulator to represent the general behavior of the plant under normal and abnormal conditions, or the modeling of the plant response to conditions that would eventually lead to core damage. In any of these, the underlying foundation for their use in analysing situations, training of vendor/utility personnel, etc. is how well they represent what has been known from industrial experience, large integral experiments and separate effects tests. Typically, simulation codes are benchmarked with some of these; the level of agreement necessary being dependent upon the ultimate use of the simulation tool. However, these analytical models are computer codes, and as a result, the capabilities are continually enhanced, errors are corrected, new situations are imposed on the code that are outside of the original design basis, etc. Consequently, there is a continual need to assure that the benchmarks with important transients are preserved as the computer code evolves. Retention of this benchmarking capability is essential to develop trust in the computer code. Given the evolving world of computer codes, how is this retention of benchmarking capabilities accomplished? For the MAAP4 codes this capability is accomplished through a 'dynamic benchmarking' feature embedded in the source code. In particular, a set of dynamic benchmarks are included in the source code and these are exercised every time the archive codes are upgraded and distributed to the MAAP users. Three different types of dynamic benchmarks are used: plant transients; large integral experiments; and separate effects tests. Each of these is performed in a different manner. The first is accomplished by developing a parameter file for the plant modeled and an input deck to describe the sequence; i.e. the entire MAAP4 code is exercised. The pertinent plant data is included in the source code and the computer

  19. DRAGON solutions to the 3D transport benchmark over a range in parameter space

    International Nuclear Information System (INIS)

    Martin, Nicolas; Hebert, Alain; Marleau, Guy

    2010-01-01

    DRAGON solutions to the 'NEA suite of benchmarks for 3D transport methods and codes over a range in parameter space' are discussed in this paper. A description of the benchmark is first provided, followed by a detailed review of the different computational models used in the lattice code DRAGON. Two numerical methods were selected for generating the required quantities for the 729 configurations of this benchmark. First, S N calculations were performed using fully symmetric angular quadratures and high-order diamond differencing for spatial discretization. To compare S N results with those of another deterministic method, the method of characteristics (MoC) was also considered for this benchmark. Comparisons between reference solutions, S N and MoC results illustrate the advantages and drawbacks of each methods for this 3-D transport problem.

  20. Present Status and Extensions of the Monte Carlo Performance Benchmark

    Science.gov (United States)

    Hoogenboom, J. Eduard; Petrovic, Bojan; Martin, William R.

    2014-06-01

    The NEA Monte Carlo Performance benchmark started in 2011 aiming to monitor over the years the abilities to perform a full-size Monte Carlo reactor core calculation with a detailed power production for each fuel pin with axial distribution. This paper gives an overview of the contributed results thus far. It shows that reaching a statistical accuracy of 1 % for most of the small fuel zones requires about 100 billion neutron histories. The efficiency of parallel execution of Monte Carlo codes on a large number of processor cores shows clear limitations for computer clusters with common type computer nodes. However, using true supercomputers the speedup of parallel calculations is increasing up to large numbers of processor cores. More experience is needed from calculations on true supercomputers using large numbers of processors in order to predict if the requested calculations can be done in a short time. As the specifications of the reactor geometry for this benchmark test are well suited for further investigations of full-core Monte Carlo calculations and a need is felt for testing other issues than its computational performance, proposals are presented for extending the benchmark to a suite of benchmark problems for evaluating fission source convergence for a system with a high dominance ratio, for coupling with thermal-hydraulics calculations to evaluate the use of different temperatures and coolant densities and to study the correctness and effectiveness of burnup calculations. Moreover, other contemporary proposals for a full-core calculation with realistic geometry and material composition will be discussed.

  1. H.B. Robinson-2 pressure vessel benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Remec, I.; Kam, F.B.K.

    1998-02-01

    The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Section 1 of this report describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity. In section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90 {+-} 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 {+-} 0.10, 0.91 {+-} 0.10, and 0.90 {+-} 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively. It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.

  2. Qinshan CANDU NPP outage performance improvement through benchmarking

    International Nuclear Information System (INIS)

    Jiang Fuming

    2005-01-01

    With the increasingly fierce competition in the deregulated Energy Market, the optimization of outage duration has become one of the focal points for the Nuclear Power Plant owners around the world. People are seeking various ways to shorten the outage duration of NPP. Great efforts have been made in the Light Water Reactor (LWR) family with the concept of benchmarking and evaluation, which great reduced the outage duration and improved outage performance. The average capacity factor of LWRs has been greatly improved over the last three decades, which now is close to 90%. CANDU (Pressurized Heavy Water Reactor) stations, with its unique feature of on power refueling, of nuclear fuel remaining in the reactor all through the planned outage, have given raise to more stringent safety requirements during planned outage. In addition, the above feature gives more variations to the critical path of planned outage in different station. In order to benchmarking again the best practices in the CANDU stations, Third Qinshan Nuclear Power Company (TQNPC) have initiated the benchmarking program among the CANDU stations aiming to standardize the outage maintenance windows and optimize the outage duration. The initial benchmarking has resulted the optimization of outage duration in Qinshan CANDU NPP and the formulation of its first long-term outage plan. This paper describes the benchmarking works that have been proven to be useful for optimizing outage duration in Qinshan CANDU NPP, and the vision of further optimize the duration with joint effort from the CANDU community. (authors)

  3. U.S. integral and benchmark experiments

    International Nuclear Information System (INIS)

    Maienschein, F.C.

    1978-01-01

    Verification of methods for analysis of radiation-transport (shielding) problems in Liquid-Metal Fast Breeder Reactors has required a series of experiments that can be classified as benchmark, parametric, or design-confirmation experiments. These experiments, performed at the Oak Ridge Tower Shielding Facility, have included measurements of neutron transport in bulk shields of sodium, steel, and inconel and in configurations that simulate lower axial shields, pipe chases, and top-head shields. They have also included measurements of the effects of fuel stored within the reactor vessel and of gamma-ray energy deposition (heating). The paper consists of brief comments on these experiments, and also on a recent experiment in which neutron streaming problems in a Gas-Cooled Fast Breeder Reactor were studied. The need for additional experiments for a few areas of LMFBR shielding is also cited

  4. Resonance self-shielding method using resonance interference factor library for practical lattice physics computations of LWRs

    International Nuclear Information System (INIS)

    Choi, Sooyoung; Khassenov, Azamat; Lee, Deokjung

    2016-01-01

    This paper presents a new method of resonance interference effect treatment using resonance interference factor for high fidelity analysis of light water reactors (LWRs). Although there have been significant improvements in the lattice physics calculations over the several decades, there exist still relatively large errors in the resonance interference treatment, in the order of ∼300 pcm in the reactivity prediction of LWRs. In the newly developed method, the impact of resonance interference to the multi-group cross-sections has been quantified and tabulated in a library which can be used in lattice physics calculation as adjustment factors of multi-group cross-sections. The verification of the new method has been performed with Mosteller benchmark, UO_2 and MOX pin-cell depletion problems, and a 17×17 fuel assembly loaded with gadolinia burnable poison, and significant improvements were demonstrated in the accuracy of reactivity and pin power predictions, with reactivity errors down to the order of ∼100 pcm. (author)

  5. PEBBLES Simulation of Static Friction and New Static Friction Benchmark

    International Nuclear Information System (INIS)

    Cogliati, Joshua J.; Ougouag, Abderrafi M.

    2010-01-01

    Pebble bed reactors contain large numbers of spherical fuel elements arranged randomly. Determining the motion and location of these fuel elements is required for calculating certain parameters of pebble bed reactor operation. This paper documents the PEBBLES static friction model. This model uses a three dimensional differential static friction approximation extended from the two dimensional Cundall and Strack model. The derivation of determining the rotational transformation of pebble to pebble static friction force is provided. A new implementation for a differential rotation method for pebble to container static friction force has been created. Previous published methods are insufficient for pebble bed reactor geometries. A new analytical static friction benchmark is documented that can be used to verify key static friction simulation parameters. This benchmark is based on determining the exact pebble to pebble and pebble to container static friction coefficients required to maintain a stable five sphere pyramid.

  6. Dynamical lattice theory

    International Nuclear Information System (INIS)

    Chodos, A.

    1978-01-01

    A version of lattice gauge theory is presented in which the shape of the lattice is not assumed at the outset but is a consequence of the dynamics. Other related features which are not specified a priori include the internal and space-time symmetry groups and the dimensionality of space-time. The theory possesses a much larger invariance group than the usual gauge group on a lattice, and has associated with it an integer k 0 analogous to the topological quantum numer of quantum chromodynamics. Families of semiclassical solutions are found which are labeled by k 0 and a second integer x, but the analysis is not carried far enough to determine which space-time and internal symmetry groups characterize the lowest-lying states of the theory

  7. Graphene antidot lattice waveguides

    DEFF Research Database (Denmark)

    Pedersen, Jesper Goor; Gunst, Tue; Markussen, Troels

    2012-01-01

    We introduce graphene antidot lattice waveguides: nanostructured graphene where a region of pristine graphene is sandwiched between regions of graphene antidot lattices. The band gaps in the surrounding antidot lattices enable localized states to emerge in the central waveguide region. We model...... the waveguides via a position-dependent mass term in the Dirac approximation of graphene and arrive at analytical results for the dispersion relation and spinor eigenstates of the localized waveguide modes. To include atomistic details we also use a tight-binding model, which is in excellent agreement...... with the analytical results. The waveguides resemble graphene nanoribbons, but without the particular properties of ribbons that emerge due to the details of the edge. We show that electrons can be guided through kinks without additional resistance and that transport through the waveguides is robust against...

  8. Regional Competitive Intelligence: Benchmarking and Policymaking

    OpenAIRE

    Huggins , Robert

    2010-01-01

    Benchmarking exercises have become increasingly popular within the sphere of regional policymaking in recent years. The aim of this paper is to analyse the concept of regional benchmarking and its links with regional policymaking processes. It develops a typology of regional benchmarking exercises and regional benchmarkers, and critically reviews the literature, both academic and policy oriented. It is argued that critics who suggest regional benchmarking is a flawed concept and technique fai...

  9. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    International Nuclear Information System (INIS)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR-06 are highlighted, and the future of the two projects is discussed

  10. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-09-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

  11. IAEA CRP on HTGR Uncertainties in Modeling: Assessment of Phase I Lattice to Core Model Uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Rouxelin, Pascal Nicolas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Best-estimate plus uncertainty analysis of reactors is replacing the traditional conservative (stacked uncertainty) method for safety and licensing analysis. To facilitate uncertainty analysis applications, a comprehensive approach and methodology must be developed and applied. High temperature gas cooled reactors (HTGRs) have several features that require techniques not used in light-water reactor analysis (e.g., coated-particle design and large graphite quantities at high temperatures). The International Atomic Energy Agency has therefore launched the Coordinated Research Project on HTGR Uncertainty Analysis in Modeling to study uncertainty propagation in the HTGR analysis chain. The benchmark problem defined for the prismatic design is represented by the General Atomics Modular HTGR 350. The main focus of this report is the compilation and discussion of the results obtained for various permutations of Exercise I 2c and the use of the cross section data in Exercise II 1a of the prismatic benchmark, which is defined as the last and first steps of the lattice and core simulation phases, respectively. The report summarizes the Idaho National Laboratory (INL) best estimate results obtained for Exercise I 2a (fresh single-fuel block), Exercise I 2b (depleted single-fuel block), and Exercise I 2c (super cell) in addition to the first results of an investigation into the cross section generation effects for the super-cell problem. The two dimensional deterministic code known as the New ESC based Weighting Transport (NEWT) included in the Standardized Computer Analyses for Licensing Evaluation (SCALE) 6.1.2 package was used for the cross section evaluation, and the results obtained were compared to the three dimensional stochastic SCALE module KENO VI. The NEWT cross section libraries were generated for several permutations of the current benchmark super-cell geometry and were then provided as input to the Phase II core calculation of the stand alone neutronics Exercise

  12. Ghera: A Repository of Android App Vulnerability Benchmarks

    OpenAIRE

    Mitra, Joydeep; Ranganath, Venkatesh-Prasad

    2017-01-01

    Security of mobile apps affects the security of their users. This has fueled the development of techniques to automatically detect vulnerabilities in mobile apps and help developers secure their apps; specifically, in the context of Android platform due to openness and ubiquitousness of the platform. Despite a slew of research efforts in this space, there is no comprehensive repository of up-to-date and lean benchmarks that contain most of the known Android app vulnerabilities and, consequent...

  13. Benchmarking of FA2D/PARCS Code Package

    International Nuclear Information System (INIS)

    Grgic, D.; Jecmenica, R.; Pevec, D.

    2006-01-01

    FA2D/PARCS code package is used at Faculty of Electrical Engineering and Computing (FER), University of Zagreb, for static and dynamic reactor core analyses. It consists of two codes: FA2D and PARCS. FA2D is a multigroup two dimensional transport theory code for burn-up calculations based on collision probability method, developed at FER. It generates homogenised cross sections both of single pins and entire fuel assemblies. PARCS is an advanced nodal code developed at Purdue University for US NRC and it is based on neutron diffusion theory for three dimensional whole core static and dynamic calculations. It is modified at FER to enable internal 3D depletion calculation and usage of neutron cross section data in a format produced by FA2D and interface codes. The FA2D/PARCS code system has been validated on NPP Krsko operational data (Cycles 1 and 21). As we intend to use this code package for development of IRIS reactor loading patterns the first logical step was to validate the FA2D/PARCS code package on a set of IRIS benchmarks, starting from simple unit fuel cell, via fuel assembly, to full core benchmark. The IRIS 17x17 fuel with erbium burnable absorber was used in last full core benchmark. The results of modelling the IRIS full core benchmark using FA2D/PARCS code package have been compared with reference data showing the adequacy of FA2D/PARCS code package model for IRIS reactor core design.(author)

  14. Exact Lattice Supersymmetry

    Energy Technology Data Exchange (ETDEWEB)

    Catterall, Simon; Kaplan, David B.; Unsal, Mithat

    2009-03-31

    We provide an introduction to recent lattice formulations of supersymmetric theories which are invariant under one or more real supersymmetries at nonzero lattice spacing. These include the especially interesting case of N = 4 SYM in four dimensions. We discuss approaches based both on twisted supersymmetry and orbifold-deconstruction techniques and show their equivalence in the case of gauge theories. The presence of an exact supersymmetry reduces and in some cases eliminates the need for fine tuning to achieve a continuum limit invariant under the full supersymmetry of the target theory. We discuss open problems.

  15. Benchmarking of human resources management

    Directory of Open Access Journals (Sweden)

    David M. Akinnusi

    2008-11-01

    Full Text Available This paper reviews the role of human resource management (HRM which, today, plays a strategic partnership role in management. The focus of the paper is on HRM in the public sector, where much hope rests on HRM as a means of transforming the public service and achieving much needed service delivery. However, a critical evaluation of HRM practices in the public sector reveals that these services leave much to be desired. The paper suggests the adoption of benchmarking as a process to revamp HRM in the public sector so that it is able to deliver on its promises. It describes the nature and process of benchmarking and highlights the inherent difficulties in applying benchmarking in HRM. It concludes with some suggestions for a plan of action. The process of identifying “best” practices in HRM requires the best collaborative efforts of HRM practitioners and academicians. If used creatively, benchmarking has the potential to bring about radical and positive changes in HRM in the public sector. The adoption of the benchmarking process is, in itself, a litmus test of the extent to which HRM in the public sector has grown professionally.

  16. Benchmark simulation models, quo vadis?

    Science.gov (United States)

    Jeppsson, U; Alex, J; Batstone, D J; Benedetti, L; Comas, J; Copp, J B; Corominas, L; Flores-Alsina, X; Gernaey, K V; Nopens, I; Pons, M-N; Rodríguez-Roda, I; Rosen, C; Steyer, J-P; Vanrolleghem, P A; Volcke, E I P; Vrecko, D

    2013-01-01

    As the work of the IWA Task Group on Benchmarking of Control Strategies for wastewater treatment plants (WWTPs) is coming to an end, it is essential to disseminate the knowledge gained. For this reason, all authors of the IWA Scientific and Technical Report on benchmarking have come together to provide their insights, highlighting areas where knowledge may still be deficient and where new opportunities are emerging, and to propose potential avenues for future development and application of the general benchmarking framework and its associated tools. The paper focuses on the topics of temporal and spatial extension, process modifications within the WWTP, the realism of models, control strategy extensions and the potential for new evaluation tools within the existing benchmark system. We find that there are major opportunities for application within all of these areas, either from existing work already being done within the context of the benchmarking simulation models (BSMs) or applicable work in the wider literature. Of key importance is increasing capability, usability and transparency of the BSM package while avoiding unnecessary complexity.

  17. Analysis of Differences in Void Coefficient Predictions for Mixed-Oxide-Fueled Tight-Pitch Light Water Reactor Cells

    International Nuclear Information System (INIS)

    Unesaki, Hironobu; Shiroya, Seiji; Kanda, Keiji; Cathalau, Stephane; Carre, Franck-Olivier; Aizawa, Otohiko; Takeda, Toshikazu

    2000-01-01

    Analysis of the benchmark problems on the void coefficient of mixed-oxide (MOX)-fueled tight-pitch cells has been performed using the Japanese SRAC code system with the JENDL-3.2 library and the French APOLLO-2 code with the CEA93 library based on JEF-2.2. The benchmark problems have been specified to investigate the physical phenomena occurring during the progressive voidage of MOX-fueled tight-pitch lattices, such as high conversion light water reactor lattices, and to evaluate the impact of nuclear data and calculational methods. Despite the most recently compiled nuclear data libraries and the sophisticated calculation schemes employed in both code systems, the k ∞ and void reactivity values obtained by the two code systems show considerable discrepancy especially in the highly voided state. The discrepancy of k ∞ values shows an obvious dependence on void fraction and also has been shown to be sensitive to the isotopic composition of plutonium. The observed discrepancies are analyzed by being decomposed into contributing isotopes and reactions and have been shown to be caused by a complicated balance of both negative and positive components, which are mainly attributable to differences in a limited number of isotopes including 239 Pu, 241 Pu, 16 O, and stainless steel

  18. 2nd RCM of the CRP on Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems (ADS) and Technical Meeting on Low Enriched Uranium (LEU) Fuel Utilization in Accelerator Driven Sub-critical Systems. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The overall objective of the CRP is contributing to the generic R&D efforts in various fields common to innovative fast neutron system development, i.e., heavy liquid metal thermal hydraulics, dedicated transmutation fuels and associated core designs, theoretical nuclear reaction models, measurement and evaluation of nuclear data for transmutation, and development and validation of calculational methods and codes. Ultimately, the CRP’s overall objective is to make contributions towards the realization of a transmutation demonstration facility

  19. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  20. Testing of cross section libraries for TRIGA criticality benchmark

    International Nuclear Information System (INIS)

    Snoj, L.; Trkov, A.; Ravnik, M.

    2007-01-01

    Influence of various up-to-date cross section libraries on the multiplication factor of TRIGA benchmark as well as the influence of fuel composition on the multiplication factor of the system composed of various types of TRIGA fuel elements was investigated. It was observed that keff calculated by using the ENDF/B VII cross section library is systematically higher than using the ENDF/B-VI cross section library. The main contributions (∼ 2 20 pcm) are from 235 U and Zr. (author)

  1. Radiation Detection Computational Benchmark Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Shaver, Mark W.; Casella, Andrew M.; Wittman, Richard S.; McDonald, Ben S.

    2013-09-24

    Modeling forms an important component of radiation detection development, allowing for testing of new detector designs, evaluation of existing equipment against a wide variety of potential threat sources, and assessing operation performance of radiation detection systems. This can, however, result in large and complex scenarios which are time consuming to model. A variety of approaches to radiation transport modeling exist with complementary strengths and weaknesses for different problems. This variety of approaches, and the development of promising new tools (such as ORNL’s ADVANTG) which combine benefits of multiple approaches, illustrates the need for a means of evaluating or comparing different techniques for radiation detection problems. This report presents a set of 9 benchmark problems for comparing different types of radiation transport calculations, identifying appropriate tools for classes of problems, and testing and guiding the development of new methods. The benchmarks were drawn primarily from existing or previous calculations with a preference for scenarios which include experimental data, or otherwise have results with a high level of confidence, are non-sensitive, and represent problem sets of interest to NA-22. From a technical perspective, the benchmarks were chosen to span a range of difficulty and to include gamma transport, neutron transport, or both and represent different important physical processes and a range of sensitivity to angular or energy fidelity. Following benchmark identification, existing information about geometry, measurements, and previous calculations were assembled. Monte Carlo results (MCNP decks) were reviewed or created and re-run in order to attain accurate computational times and to verify agreement with experimental data, when present. Benchmark information was then conveyed to ORNL in order to guide testing and development of hybrid calculations. The results of those ADVANTG calculations were then sent to PNNL for

  2. Quarks, gluons and lattices

    International Nuclear Information System (INIS)

    Krojts, M.

    1987-01-01

    The book by the known american physicist-theoretist M.Kreuts represents the first monography in world literature, where a new perspective direction in elementary particle physics and quantum field theory - lattice formulation of gauge theories is stated systematically. Practically all main ideas of this direction are given. Material is stated in systematic and understandable form

  3. Phenomenology Using Lattice QCD

    Science.gov (United States)

    Gupta, R.

    2005-08-01

    This talk provides a brief summary of the status of lattice QCD calculations of the light quark masses and the kaon bag parameter BK. Precise estimates of these four fundamental parameters of the standard model, i.e., mu, md, ms and the CP violating parameter η, help constrain grand unified models and could provide a window to new physics.

  4. Baryons on the lattice

    International Nuclear Information System (INIS)

    Bali, G.S.

    2005-01-01

    I comment on progress of lattice QCD techniques and calculations. Recent results on pentaquark masses as well as of the spectrum of excited baryons are summarized and interpreted. The present state of calculations of quantities related to the nucleon structure and of electromagnetic transition form factors is surveyed

  5. Finite lattice extrapolation algorithms

    International Nuclear Information System (INIS)

    Henkel, M.; Schuetz, G.

    1987-08-01

    Two algorithms for sequence extrapolation, due to von den Broeck and Schwartz and Bulirsch and Stoer are reviewed and critically compared. Applications to three states and six states quantum chains and to the (2+1)D Ising model show that the algorithm of Bulirsch and Stoer is superior, in particular if only very few finite lattice data are available. (orig.)

  6. Lattice Multiverse Models

    OpenAIRE

    Williamson, S. Gill

    2010-01-01

    Will the cosmological multiverse, when described mathematically, have easily stated properties that are impossible to prove or disprove using mathematical physics? We explore this question by constructing lattice multiverses which exhibit such behavior even though they are much simpler mathematically than any likely cosmological multiverse.

  7. Quantum lattice problems

    NARCIS (Netherlands)

    de Raedt, Hans; von der Linden, W.; Binder, K

    1995-01-01

    In this chapter we review methods currently used to perform Monte Carlo calculations for quantum lattice models. A detailed exposition is given of the formalism underlying the construction of the simulation algorithms. We discuss the fundamental and technical difficulties that are encountered and

  8. Convex Lattice Polygons

    Science.gov (United States)

    Scott, Paul

    2006-01-01

    A "convex" polygon is one with no re-entrant angles. Alternatively one can use the standard convexity definition, asserting that for any two points of the convex polygon, the line segment joining them is contained completely within the polygon. In this article, the author provides a solution to a problem involving convex lattice polygons.

  9. Lattices for antiproton rings

    International Nuclear Information System (INIS)

    Autin, B.

    1984-01-01

    After a description of the constraints imposed by the cooling of Antiprotons on the lattice of the rings, the reasons which motivate the shape and the structure of these machines are surveyed. Linear and non-linear beam optics properties are treated with a special amplification to the Antiproton Accumulator. (orig.)

  10. EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

    International Nuclear Information System (INIS)

    Jones, D.B.

    1986-01-01

    EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

  11. 3-D neutron transport benchmarks

    International Nuclear Information System (INIS)

    Takeda, T.; Ikeda, H.

    1991-03-01

    A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of K eff , control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes

  12. Strategic behaviour under regulatory benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Jamasb, T. [Cambridge Univ. (United Kingdom). Dept. of Applied Economics; Nillesen, P. [NUON NV (Netherlands); Pollitt, M. [Cambridge Univ. (United Kingdom). Judge Inst. of Management

    2004-09-01

    In order to improve the efficiency of electricity distribution networks, some regulators have adopted incentive regulation schemes that rely on performance benchmarking. Although regulation benchmarking can influence the ''regulation game,'' the subject has received limited attention. This paper discusses how strategic behaviour can result in inefficient behaviour by firms. We then use the Data Envelopment Analysis (DEA) method with US utility data to examine implications of illustrative cases of strategic behaviour reported by regulators. The results show that gaming can have significant effects on the measured performance and profitability of firms. (author)

  13. Atomic Energy Research benchmark activity

    International Nuclear Information System (INIS)

    Makai, M.

    1998-01-01

    The test problems utilized in the validation and verification process of computer programs in Atomic Energie Research are collected into one bunch. This is the first step towards issuing a volume in which tests for VVER are collected, along with reference solutions and a number of solutions. The benchmarks do not include the ZR-6 experiments because they have been published along with a number of comparisons in the Final reports of TIC. The present collection focuses on operational and mathematical benchmarks which cover almost the entire range of reaktor calculation. (Author)

  14. Protocol Fuel Mix reporting

    International Nuclear Information System (INIS)

    2002-07-01

    The protocol in this document describes a method for an Electricity Distribution Company (EDC) to account for the fuel mix of electricity that it delivers to its customers, based on the best available information. Own production, purchase and sale of electricity, and certificates trading are taken into account. In chapter 2 the actual protocol is outlined. In the appendixes additional (supporting) information is given: (A) Dutch Standard Fuel Mix, 2000; (B) Calculation of the Dutch Standard fuel mix; (C) Procedures to estimate and benchmark the fuel mix; (D) Quality management; (E) External verification; (F) Recommendation for further development of the protocol; (G) Reporting examples

  15. Unquenched lattice upsilon spectroscopy

    International Nuclear Information System (INIS)

    Marcantonio, L.M.

    2001-03-01

    A non-relativistic effective theory of QCD (NRQCD) is used in calculations of the upsilon spectrum. Simultaneous multi-correlation fitting routines are used to yield lattice channel energies and amplitudes. The lattice configurations used were both dynamical, with two flavours of sea quarks included in the action; and quenched, with no sea quarks. These configurations were generated by the UKQCD collaboration. The dynamical configurations used were ''matched'', having the same lattice spacing, but differing in the sea quark mass. Thus, it was possible to analyse trends of observables with sea quark mass, in the certainty that the trend isn't partially due to varying lattice spacing. The lattice spacing used for spectroscopy was derived from the lattice 1 1 P 1 - 1 3 S 1 splitting. On each set of configurations two lattice bare b quark masses were used, giving kinetic masses bracketing the physical Υ mass. The only quantity showing a strong dependence on these masses was the hyperfine splitting, so it was interpolated to the real Υ mass. The radial and orbital splittings gave good agreement with experiment. The hyperfine splitting results showed a clear signal for unquenching and the dynamical hyperfine splitting results were extrapolated to a physical sea quark mass. This result, combined with the quenched result yielded a value for the hyperfine splitting at n f = 3, predicting an η b mass of 9.517(4) GeV. The NRQCD technique for obtaining a value of the strong coupling constant in the M-barS-bar scheme was followed. Using quenched and dynamical results a value was extrapolated to n f = 3. Employing a three loop beta function to run the coupling, with suitable matching conditions at heavy quark thresholds, the final result was obtained for n f = 5 at a scale equal to the Z boson mass. This result was α(5)/MS(Mz)=0.110(4). Two methods for finding the mass of the b quark in the MS scheme were employed. The results of both methods agree within error but the

  16. Benchmark for evaluation and validation of reactor simulations (BEAVRS)

    Energy Technology Data Exchange (ETDEWEB)

    Horelik, N.; Herman, B.; Forget, B.; Smith, K. [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

    2013-07-01

    Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

  17. Developing and modeling of the 'Laguna Verde' BWR CRDA benchmark

    International Nuclear Information System (INIS)

    Solis-Rodarte, J.; Fu, H.; Ivanov, K.N.; Matsui, Y.; Hotta, A.

    2002-01-01

    Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant - unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The 'Laguna Verde' (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTREE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions

  18. APOLLO2 calculations of RBMK lattices

    International Nuclear Information System (INIS)

    Kalashnikov, D.

    1998-01-01

    The purpose of this study is to investigate the use of erbium as burnable poison in RBMK reactors. The neutronic code APOLLO2 has been used and a comparison with the Monte-Carlo code TRIPOLI2 has been made. The first chapter briefly presents the RBMK characteristics, the second chapter deals with the neutronic behaviour of a fuel assembly in an infinite lattice which is an important step in the modelling process. The third chapter presents the analysis of the use of erbium in typical elements of the RBMK lattice. A good agreement is obtained between the 2 codes except in the draining situations. Erbium appears to reduce the positive reactivity effect of the draining configuration. (A.C.)

  19. Sensitivity study applied to the CB4 VVER-440 benchmark on burnup credit

    International Nuclear Information System (INIS)

    Markova, Ludmila

    2003-01-01

    A brief overview of four completed portions (CB1, CB2, CB3, CB3+, CB4) of the international VVER-440 benchmark focused on burnup credit and a sensitivity study as one of the final views of the benchmark results are presented in the paper. Finally, the influence of real and conservative VVER-440 fuel assembly models taken for the isotopics calculation by SCALE sas2 on the system k eff is shown in the paper. (author)

  20. The programme PIP2 for lattice cell thermal calculations

    International Nuclear Information System (INIS)

    Clayton, A.J.

    1964-08-01

    The programme PIP2 solves the multigroup equations obtained by applying the method of collision probabilities to a fuel region (which may contain a cluster of fuel elements), and the SPECTROX flux assumption in a surrounding 'moderator'. The programme does not calculate collision probabilities for the fuel region and any geometry can be treated in the fuel region for which collision probabilities can be calculated. Lattice cell source problems may be treated and it is possible to include part of the physical moderator with the fuel region for treatment by the collision probability method. The programme is primarily intended for thermal fixed source problems, with the sources in the (physical moderator), but by including part of the moderator with the fuel it is possible to include fixed sources in the fuel for the study of fast effects. (author)

  1. Behaviour of defective CANDU fuel: fuel oxidation kinetic and thermodynamic modelling

    International Nuclear Information System (INIS)

    Higgs, J.

    2005-01-01

    The thermal performance of operating CANDU fuel under defect conditions is affected by the ingress of heavy water into the fuel element. A mechanistic model has been developed to predict the extent of fuel oxidation in defective fuel and its affect on fuel thermal performance. A thermodynamic treatment of such oxidized fuel has been performed as a basis for the boundary conditions in the kinetic model. Both the kinetic and thermodynamic models have been benchmarked against recent experimental work. (author)

  2. Superspace approach to lattice supersymmetry

    International Nuclear Information System (INIS)

    Kostelecky, V.A.; Rabin, J.M.

    1984-01-01

    We construct a cubic lattice of discrete points in superspace, as well as a discrete subgroup of the supersymmetry group which maps this ''superlattice'' into itself. We discuss the connection between this structure and previous versions of lattice supersymmetry. Our approach clarifies the mathematical problems of formulating supersymmetric lattice field theories and suggests new methods for attacking them

  3. Basis reduction for layered lattices

    NARCIS (Netherlands)

    Torreão Dassen, Erwin

    2011-01-01

    We develop the theory of layered Euclidean spaces and layered lattices. We present algorithms to compute both Gram-Schmidt and reduced bases in this generalized setting. A layered lattice can be seen as lattices where certain directions have infinite weight. It can also be

  4. An overview of lattice QCD

    International Nuclear Information System (INIS)

    Woloshyn, R.M.

    1988-03-01

    The basic concepts of the Lagrangian formulation of lattice field theory are discussed. The Wilson and staggered schemes for dealing with fermions on the lattice are described. Some recent results for hadron masses and vector and axial vector current matrix elements in lattice QCD are reviewed. (Author) (118 refs., 16 figs.)

  5. Basis reduction for layered lattices

    NARCIS (Netherlands)

    E.L. Torreão Dassen (Erwin)

    2011-01-01

    htmlabstractWe develop the theory of layered Euclidean spaces and layered lattices. With this new theory certain problems that usually are solved by using classical lattices with a "weighting" gain a new, more natural form. Using the layered lattice basis reduction algorithms introduced here these

  6. Benchmarked Library Websites Comparative Study

    KAUST Repository

    Ramli, Rindra M.; Tyhurst, Janis

    2015-01-01

    This presentation provides an analysis of services provided by the benchmarked library websites. The exploratory study includes comparison of these websites against a list of criterion and presents a list of services that are most commonly deployed by the selected websites. In addition to that, the investigators proposed a list of services that could be provided via the KAUST library website.

  7. Prismatic Core Coupled Transient Benchmark

    International Nuclear Information System (INIS)

    Ortensi, J.; Pope, M.A.; Strydom, G.; Sen, R.S.; DeHart, M.D.; Gougar, H.D.; Ellis, C.; Baxter, A.; Seker, V.; Downar, T.J.; Vierow, K.; Ivanov, K.

    2011-01-01

    The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivated the development of more accurate and efficient tools for the design and safety evaluations of the PMR. In addition to the work invested in new methods, it is essential to develop appropriate benchmarks to verify and validate the new methods in computer codes. The purpose of this benchmark is to establish a well-defined problem, based on a common given set of data, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events. The benchmark-working group is currently seeking OECD/NEA sponsorship. This benchmark is being pursued and is heavily based on the success of the PBMR-400 exercise.

  8. Analysis of the impact of correlated benchmark experiments on the validation of codes for criticality safety analysis

    International Nuclear Information System (INIS)

    Bock, M.; Stuke, M.; Behler, M.

    2013-01-01

    The validation of a code for criticality safety analysis requires the recalculation of benchmark experiments. The selected benchmark experiments are chosen such that they have properties similar to the application case that has to be assessed. A common source of benchmark experiments is the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments' (ICSBEP Handbook) compiled by the 'International Criticality Safety Benchmark Evaluation Project' (ICSBEP). In order to take full advantage of the information provided by the individual benchmark descriptions for the application case, the recommended procedure is to perform an uncertainty analysis. The latter is based on the uncertainties of experimental results included in most of the benchmark descriptions. They can be performed by means of the Monte Carlo sampling technique. The consideration of uncertainties is also being introduced in the supplementary sheet of DIN 25478 'Application of computer codes in the assessment of criticality safety'. However, for a correct treatment of uncertainties taking into account the individual uncertainties of the benchmark experiments is insufficient. In addition, correlations between benchmark experiments have to be handled correctly. For example, these correlations can arise due to different cases of a benchmark experiment sharing the same components like fuel pins or fissile solutions. Thus, manufacturing tolerances of these components (e.g. diameter of the fuel pellets) have to be considered in a consistent manner in all cases of the benchmark experiment. At the 2012 meeting of the Expert Group on 'Uncertainty Analysis for Criticality Safety Assessment' (UACSA) of the OECD/NEA a benchmark proposal was outlined that aimed for the determination of the impact on benchmark correlations on the estimation of the computational bias of the neutron multiplication factor (k eff ). The analysis presented here is based on this proposal. (orig.)

  9. Determination of D2O - 2% enriched uranium lattice parameters by means of a critical system

    International Nuclear Information System (INIS)

    Raisic, N.; Takac, S.; Markovic, H.; Bosevski, T.

    1963-01-01

    In order to specify experimental procedures for few standard measurements sufficient to provide consistent set of lattice parameters, a series of experiments were performed at the RB reactor using 2% enriched tubular fuel elements. Obtained results were compared to standard two-group diffusion calculation indicating high degree of accuracy for a broad variety of reactor lattice configurations

  10. Lattice QCD calculations on commodity clusters at DESY

    International Nuclear Information System (INIS)

    Gellrich, A.; Pop, D.; Wegner, P.; Wittig, H.; Hasenbusch, M.; Jansen, K.

    2003-06-01

    Lattice Gauge Theory is an integral part of particle physics that requires high performance computing in the multi-Tflops regime. These requirements are motivated by the rich research program and the physics milestones to be reached by the lattice community. Over the last years the enormous gains in processor performance, memory bandwidth, and external I/O bandwidth for parallel applications have made commodity clusters exploiting PCs or workstations also suitable for large Lattice Gauge Theory applications. For more than one year two clusters have been operated at the two DESY sites in Hamburg and Zeuthen, consisting of 32 resp. 16 dual-CPU PCs, equipped with Intel Pentium 4 Xeon processors. Interconnection of the nodes is done by way of Myrinet. Linux was chosen as the operating system. In the course of the projects benchmark programs for architectural studies were developed. The performance of the Wilson-Dirac Operator (also in an even-odd preconditioned version) as the inner loop of the Lattice QCD (LQCD) algorithms plays the most important role in classifying the hardware basis to be used. Using the SIMD streaming extensions (SSE/SSE2) on Intel's Pentium 4 Xeon CPUs give promising results for both the single CPU and the parallel version. The parallel performance, in addition to the CPU power and the memory throughput, is nevertheless strongly influenced by the behavior of hardware components like the PC chip-set and the communication interfaces. The paper starts by giving a short explanation about the physics background and the motivation for using PC clusters for Lattice QCD. Subsequently, the concept, implementation, and operating experiences of the two clusters are discussed. Finally, the paper presents benchmark results and discusses comparisons to systems with different hardware components including Myrinet-, GigaBit-Ethernet-, and Infiniband-based interconnects. (orig.)

  11. Application of WIMSD-4 for ''MARIA'' reactor lattice calculations

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.

    1993-12-01

    A general description of the WIMSD-4 lattice code is given with the emphasis on available geometrical models. The difficulties encountered while modelling reactor lattices with the tubular type fuel elements are explained. Then the analysis of code options allowing to overcome these difficulties is carried out. Eventually, recommendations of options and input parameters for calculations of MARIA reactor lattice with satisfactory accuracy are given. During the work a set of modifications had to be introduced leading to a new code version called WIMS-S. Another version, under the name WIMS-T has been developed to allow for burnup calculations of the MARIA reactor lattice with improved resonance approach. (author). 14 refs, 6 figs, 10 tabs

  12. Datagrids for lattice QCD

    Energy Technology Data Exchange (ETDEWEB)

    Buechner, O. [Zentralinstitut fuer Angewandte Mathematik ZAM, 52425 Juelich (Germany); Ernst, M. [Deutsches Elektronen-Synchrotron DESY, 22603 Hamburg (Germany); Jansen, K. [John von Neumann-Institut fuer Computing NIC/DESY, 15738 Zeuthen (Germany); Lippert, Th. [Zentralinstitut fuer Angewandte Mathematik ZAM, 52425 Juelich (Germany); Melkumyan, D. [Deutsches Elektronen-Synchrotron DESY, 15738 Zeuthen (Germany); Orth, B. [Zentralinstitut fuer Angewandte Mathematik ZAM, 52425 Juelich (Germany); Pleiter, D. [John von Neumann-Institut fuer Computing NIC/DESY, 15738 Zeuthen (Germany)]. E-mail: dirk.pleiter@desy.de; Stueben, H. [Konrad-Zuse-Institut fuer Informationstechnik ZIB, 14195 Berlin (Germany); Wegner, P. [Deutsches Elektronen-Synchrotron DESY, 15738 Zeuthen (Germany); Wollny, S. [Konrad-Zuse-Institut fuer Informationstechnik ZIB, 14195 Berlin (Germany)

    2006-04-01

    As the need for computing resources to carry out numerical simulations of Quantum Chromodynamics (QCD) formulated on a lattice has increased significantly, efficient use of the generated data has become a major concern. To improve on this, groups plan to share their configurations on a worldwide level within the International Lattice DataGrid (ILDG). Doing so requires standardized description of the configurations, standards on binary file formats and common middleware interfaces. We describe the requirements and problems, and discuss solutions. Furthermore, an overview is given on the implementation of the LatFor DataGrid [http://www-zeuthen.desy.de/latfor/ldg], a France/German/Italian grid that will be one of the regional grids within the ILDG grid-of-grids concept.

  13. Lattice QCD for cosmology

    International Nuclear Information System (INIS)

    Borsanyi, Sz.; Kampert, K.H.; Fodor, Z.; Forschungszentrum Juelich; Eoetvoes Univ., Budapest

    2016-06-01

    We present a full result for the equation of state (EoS) in 2+1+1 (up/down, strange and charm quarks are present) flavour lattice QCD. We extend this analysis and give the equation of state in 2+1+1+1 flavour QCD. In order to describe the evolution of the universe from temperatures several hundreds of GeV to the MeV scale we also include the known effects of the electroweak theory and give the effective degree of freedoms. As another application of lattice QCD we calculate the topological susceptibility (χ) up to the few GeV temperature region. These two results, EoS and χ, can be used to predict the dark matter axion's mass in the post-inflation scenario and/or give the relationship between the axion's mass and the universal axionic angle, which acts as a initial condition of our universe.

  14. Lattice vibration spectra. 16

    International Nuclear Information System (INIS)

    Lutz, H.D.; Willich, P.

    1977-01-01

    The FIR absorption spectra of pyrite type compounds RuS 2 , RuSsub(2-x)Sesub(x), RuSe 2 , RuTe 2 , OsS 2 , OsSe 2 , and PtP 2 as well as loellingite type phosphides FeP 2 , RuP 2 , and OsP 2 are reported. For RuS 2 , RuSe 2 , RuTe 2 , OsS 2 , and PtP 2 all of the five infrared allowed modes (k = 0) are observed. As a first result of a numerical normal coordinate treatment vibration forms of pyrite structure are communicated. The spectra show that lattice forces of corresponding sulfides, tellurides, and phosphides are about the same strength, but increase strongly by substitution of iron by ruthenium and especially of ruthenium by osmium. The lattice constants of the RuSsub(2-x)Sesub(x) solid solution obey Vegard's rule. (author)

  15. Lattice Wigner equation

    Science.gov (United States)

    Solórzano, S.; Mendoza, M.; Succi, S.; Herrmann, H. J.

    2018-01-01

    We present a numerical scheme to solve the Wigner equation, based on a lattice discretization of momentum space. The moments of the Wigner function are recovered exactly, up to the desired order given by the number of discrete momenta retained in the discretization, which also determines the accuracy of the method. The Wigner equation is equipped with an additional collision operator, designed in such a way as to ensure numerical stability without affecting the evolution of the relevant moments of the Wigner function. The lattice Wigner scheme is validated for the case of quantum harmonic and anharmonic potentials, showing good agreement with theoretical results. It is further applied to the study of the transport properties of one- and two-dimensional open quantum systems with potential barriers. Finally, the computational viability of the scheme for the case of three-dimensional open systems is also illustrated.

  16. Lattice Quantum Chromodynamics

    CERN Document Server

    Sachrajda, C T

    2016-01-01

    I review the the application of the lattice formulation of QCD and large-scale numerical simulations to the evaluation of non-perturbative hadronic effects in Standard Model Phenomenology. I present an introduction to the elements of the calculations and discuss the limitations both in the range of quantities which can be studied and in the precision of the results. I focus particularly on the extraction of the QCD parameters, i.e. the quark masses and the strong coupling constant, and on important quantities in flavour physics. Lattice QCD is playing a central role in quantifying the hadronic effects necessary for the development of precision flavour physics and its use in exploring the limits of the Standard Model and in searches for inconsistencies which would signal the presence of new physics.

  17. Lattices of dielectric resonators

    CERN Document Server

    Trubin, Alexander

    2016-01-01

    This book provides the analytical theory of complex systems composed of a large number of high-Q dielectric resonators. Spherical and cylindrical dielectric resonators with inferior and also whispering gallery oscillations allocated in various lattices are considered. A new approach to S-matrix parameter calculations based on perturbation theory of Maxwell equations, developed for a number of high-Q dielectric bodies, is introduced. All physical relationships are obtained in analytical form and are suitable for further computations. Essential attention is given to a new unified formalism of the description of scattering processes. The general scattering task for coupled eigen oscillations of the whole system of dielectric resonators is described. The equations for the  expansion coefficients are explained in an applicable way. The temporal Green functions for the dielectric resonator are presented. The scattering process of short pulses in dielectric filter structures, dielectric antennas  and lattices of d...

  18. Lattice gauge theories

    International Nuclear Information System (INIS)

    Hasenfratz, A.; Hasenfratz, P.

    1985-01-01

    This paper deals almost exclusively with applications in QCD. Presumably QCD will remain in the center of lattice calculations in the near future. The existing techniques and the available computer resources should be able to produce trustworthy results in pure SU(3) gauge theory and in quenched hadron spectroscopy. Going beyond the quenched approximation might require some technical breakthrough or exceptional computer resources, or both. Computational physics has entered high-energy physics. From this point of view, lattice QCD is only one (although the most important, at present) of the research fields. Increasing attention is devoted to the study of other QFTs. It is certain that the investigation of nonasymptotically free theories, the Higgs phenomenon, or field theories that are not perturbatively renormalizable will be important research areas in the future

  19. Research on Elemental Technology of Advanced Nuclear Fuel Performance Verification

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Dong Uk; Jean, Sang Hwan; Koo, Min

    2003-04-01

    Most of current properties models and fuel performance models used in the performance evaluation codes are based on the in-pile data up to 33,000 MWd/MtU. Therefore, international experts are investigating the properties changes and developing advanced prediction models for high burn-up application. Current research is to develop high burn-up fission gas release model for the code and to support the code development activities by collecting data and models, reviewing/assessing the data and models together, and benchmarking the selected models against the appropriate in-pile data. For high burn-up applications, two stage two step fission gas release model is developed based on the real two diffusion process in the grain lattice and grain boundaries of the fission gases and the observation of accelerated release rate in the high burn-up. It is found that the prediction of this model is in excellent agreement with the in-pile measurement results, not only in the low burn-up but also in the high burn-up. This research is found that the importance of thermal conductivity of oxide fuel, especially in the high burn-up, is focused again. It is found that even the temperature dependent models differ from one to another and most of them overestimate the conductivity in the high burn-up. An in-pile data benchmarking of high LHGR fuel rod shows that the difference can reach 30%∼40%, which predicts 400 .deg. C lower than the real fuel centerline temperature. Recent models on the thermal expansion and heat capacity of oxide fuel are found to be well-defined. Irradiation swelling of the oxide fuel are now well-understood that in most cases in LWRs solid fission product swelling is dominant. Thus, the accumulation of in-pile data can enhance the accuracy of the model prediction, rather than theoretical modeling works. Thermo-physical properties of Zircaloy cladding are also well-defined and well-understood except the thermal expansion. However, it turns out that even the

  20. Lattice degeneracies of geometric fermions

    International Nuclear Information System (INIS)

    Raszillier, H.

    1983-05-01

    We give the minimal numbers of degrees of freedom carried by geometric fermions on all lattices of maximal symmetries in d = 2, 3, and 4 dimensions. These numbers are lattice dependent, but in the (free) continuum limit, part of the degrees of freedom have to escape to infinity by a Wilson mechanism built in, and 2sup(d) survive for any lattice. On self-reciprocal lattices we compare the minimal numbers of degrees of freedom of geometric fermions with the minimal numbers of naive fermions on these lattices and argue that these numbers are equal. (orig.)