Analysis of the European results on the HTTR's core physics benchmarks

International Nuclear Information System (INIS)

Raepsaet, X.; Damian, F.; Ohlig, U.A.; Brockmann, H.J.; Haas, J.B.M. de; Wallerboss, E.M.

2002-01-01

Within the frame of the European contract HTR-N1 calculations are performed on the benchmark problems of the HTTR's start-up core physics experiments initially proposed by the IAEA in a Co-ordinated Research Programme. Three European partners, the FZJ in Germany, NRG and IRI in the Netherlands, and CEA in France, have joined this work package with the aim to validate their calculational methods. Pre-test and post-test calculational results, obtained by the partners, are compared with each other and with the experiment. Parts of the discrepancies between experiment and pre-test predictions are analysed and tackled by different treatments. In the case of the Monte Carlo code TRIPOLI4, used by CEA, the discrepancy between measurement and calculation at the first criticality is reduced to Δk/k∼0.85%, when considering the revised data of the HTTR benchmark. In the case of the diffusion codes, this discrepancy is reduced to: Δk/k∼0.8% (FZJ) and 2.7 or 1.8% (CEA). (author)

Power-Energy Simulation for Multi-Core Processors in Bench-marking

Directory of Open Access Journals (Sweden)

Mona A. Abou-Of

2017-01-01

Full Text Available At Microarchitectural level, multi-core processor, as a complex System on Chip, has sophisticated on-chip components including cores, shared caches, interconnects and system controllers such as memory and ethernet controllers. At technological level, architects should consider the device types forecast in the International Technology Roadmap for Semiconductors (ITRS. Energy simulation enables architects to study two important metrics simultaneously. Timing is a key element of the CPU performance that imposes constraints on the CPU target clock frequency. Power and the resulting heat impose more severe design constraints, such as core clustering, while semiconductor industry is providing more transistors in the die area in pace with Moore’s law. Energy simulators provide a solution for such serious challenge. Energy is modelled either by combining performance benchmarking tool with a power simulator or by an integrated framework of both performance simulator and power profiling system. This article presents and asses trade-offs between different architectures using four cores battery-powered mobile systems by running a custom-made and a standard benchmark tools. The experimental results assure the Energy/ Frequency convexity rule over a range of frequency settings on different number of enabled cores. The reported results show that increasing the number of cores has a great effect on increasing the power consumption. However, a minimum energy dissipation will occur at a lower frequency which reduces the power consumption. Despite that, increasing the number of cores will also increase the effective cores value which will reflect a better processor performance.

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

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

Energy Technology Data Exchange (ETDEWEB)

Ellis, RJ

2001-02-02

The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.

Defining core elements and outstanding practice in Nutritional Science through collaborative benchmarking.

Science.gov (United States)

Samman, Samir; McCarthur, Jennifer O; Peat, Mary

2006-01-01

Benchmarking has been adopted by educational institutions as a potentially sensitive tool for improving learning and teaching. To date there has been limited application of benchmarking methodology in the Discipline of Nutritional Science. The aim of this survey was to define core elements and outstanding practice in Nutritional Science through collaborative benchmarking. Questionnaires that aimed to establish proposed core elements for Nutritional Science, and inquired about definitions of " good" and " outstanding" practice were posted to named representatives at eight Australian universities. Seven respondents identified core elements that included knowledge of nutrient metabolism and requirement, food production and processing, modern biomedical techniques that could be applied to understanding nutrition, and social and environmental issues as related to Nutritional Science. Four of the eight institutions who agreed to participate in the present survey identified the integration of teaching with research as an indicator of outstanding practice. Nutritional Science is a rapidly evolving discipline. Further and more comprehensive surveys are required to consolidate and update the definition of the discipline, and to identify the optimal way of teaching it. Global ideas and specific regional requirements also need to be considered.

Criticality safety benchmark experiment on 10% enriched uranyl nitrate solution using a 28-cm-thickness slab core

International Nuclear Information System (INIS)

Yamamoto, Toshihiro; Miyoshi, Yoshinori; Kikuchi, Tsukasa; Watanabe, Shouichi

2002-01-01

The second series of critical experiments with 10% enriched uranyl nitrate solution using 28-cm-thick slab core have been performed with the Static Experiment Critical Facility of the Japan Atomic Energy Research Institute. Systematic critical data were obtained by changing the uranium concentration of the fuel solution from 464 to 300 gU/l under various reflector conditions. In this paper, the thirteen critical configurations for water-reflected cores and unreflected cores are identified and evaluated. The effects of uncertainties in the experimental data on k eff are quantified by sensitivity studies. Benchmark model specifications that are necessary to construct a calculational model are given. The uncertainties of k eff 's included in the benchmark model specifications are approximately 0.1%Δk eff . The thirteen critical configurations are judged to be acceptable benchmark data. Using the benchmark model specifications, sample calculation results are provided with several sets of standard codes and cross section data. (author)

Calculational benchmark comparisons for a low sodium void worth actinide burner core design

International Nuclear Information System (INIS)

Hill, R.N.; Kawashima, M.; Arie, K.; Suzuki, M.

1992-01-01

Recently, a number of low void worth core designs with non-conventional core geometries have been proposed. Since these designs lack a good experimental and computational database, benchmark calculations are useful for the identification of possible biases in performance characteristics predictions. In this paper, a simplified benchmark model of a metal fueled, low void worth actinide burner design is detailed; and two independent neutronic performance evaluations are compared. Calculated performance characteristics are evaluated for three spatially uniform compositions (fresh uranium/plutonium, batch-averaged uranium/transuranic, and batch-averaged uranium/transuranic with fission products) and a regional depleted distribution obtained from a benchmark depletion calculation. For each core composition, the flooded and voided multiplication factor, power peaking factor, sodium void worth (and its components), flooded Doppler coefficient and control rod worth predictions are compared. In addition, the burnup swing, average discharge burnup, peak linear power, and fresh fuel enrichment are calculated for the depletion case. In general, remarkably good agreement is observed between the evaluations. The most significant difference is predicted performance characteristics is a 0.3--0.5% Δk/(kk) bias in the sodium void worth. Significant differences in the transmutation rate of higher actinides are also observed; however, these differences do not cause discrepancies in the performing predictions

Utilizing benchmark data from the ANL-ZPR diagnostic cores program

International Nuclear Information System (INIS)

Schaefer, R. W.; McKnight, R. D.

2000-01-01

The support of the criticality safety community is allowing the production of benchmark descriptions of several assemblies from the ZPR Diagnostic Cores Program. The assemblies have high sensitivities to nuclear data for a few isotopes. This can highlight limitations in nuclear data for selected nuclides or in standard methods used to treat these data. The present work extends the use of the simplified model of the U9 benchmark assembly beyond the validation of k eff . Further simplifications have been made to produce a data testing benchmark in the style of the standard CSEWG benchmark specifications. Calculations for this data testing benchmark are compared to results obtained with more detailed models and methods to determine their biases. These biases or corrections factors can then be applied in the use of the less refined methods and models. Data testing results using Versions IV, V, and VI of the ENDF/B nuclear data are presented for k eff , f 28 /f 25 , c 28 /f 25 , and β eff . These limited results demonstrate the importance of studying other integral parameters in addition to k eff in trying to improve nuclear data and methods and the importance of accounting for methods and/or modeling biases when using data testing results to infer the quality of the nuclear data files

A benchmarking tool to evaluate computer tomography perfusion infarct core predictions against a DWI standard.

Science.gov (United States)

Cereda, Carlo W; Christensen, Søren; Campbell, Bruce Cv; Mishra, Nishant K; Mlynash, Michael; Levi, Christopher; Straka, Matus; Wintermark, Max; Bammer, Roland; Albers, Gregory W; Parsons, Mark W; Lansberg, Maarten G

2016-10-01

Differences in research methodology have hampered the optimization of Computer Tomography Perfusion (CTP) for identification of the ischemic core. We aim to optimize CTP core identification using a novel benchmarking tool. The benchmarking tool consists of an imaging library and a statistical analysis algorithm to evaluate the performance of CTP. The tool was used to optimize and evaluate an in-house developed CTP-software algorithm. Imaging data of 103 acute stroke patients were included in the benchmarking tool. Median time from stroke onset to CT was 185 min (IQR 180-238), and the median time between completion of CT and start of MRI was 36 min (IQR 25-79). Volumetric accuracy of the CTP-ROIs was optimal at an rCBF threshold of benchmarking tool can play an important role in optimizing CTP software as it provides investigators with a novel method to directly compare the performance of alternative CTP software packages. © The Author(s) 2015.

Benchmarking NWP Kernels on Multi- and Many-core Processors

Science.gov (United States)

Michalakes, J.; Vachharajani, M.

2008-12-01

Increased computing power for weather, climate, and atmospheric science has provided direct benefits for defense, agriculture, the economy, the environment, and public welfare and convenience. Today, very large clusters with many thousands of processors are allowing scientists to move forward with simulations of unprecedented size. But time-critical applications such as real-time forecasting or climate prediction need strong scaling: faster nodes and processors, not more of them. Moreover, the need for good cost- performance has never been greater, both in terms of performance per watt and per dollar. For these reasons, the new generations of multi- and many-core processors being mass produced for commercial IT and "graphical computing" (video games) are being scrutinized for their ability to exploit the abundant fine- grain parallelism in atmospheric models. We present results of our work to date identifying key computational kernels within the dynamics and physics of a large community NWP model, the Weather Research and Forecast (WRF) model. We benchmark and optimize these kernels on several different multi- and many-core processors. The goals are to (1) characterize and model performance of the kernels in terms of computational intensity, data parallelism, memory bandwidth pressure, memory footprint, etc. (2) enumerate and classify effective strategies for coding and optimizing for these new processors, (3) assess difficulties and opportunities for tool or higher-level language support, and (4) establish a continuing set of kernel benchmarks that can be used to measure and compare effectiveness of current and future designs of multi- and many-core processors for weather and climate applications.

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

Energy Technology Data Exchange (ETDEWEB)

Ellis, RJ

2001-06-01

The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

CFD-calculations to a core catcher benchmark

International Nuclear Information System (INIS)

Willschuetz, H.G.

1999-04-01

There are numerous experiments for the exploration of the corium spreading behaviour, but comparable data have not been available up to now in the field of the long term behaviour of a corium expanded in a core catcher. The difficulty consists in the experimental simulation of the decay heat that can be neglected for the short-run course of events like relocation and spreading, which must, however, be considered during investigation of the long time behaviour. Therefore the German GRS, defined together with Battelle Ingenieurtechnik a benchmark problem in order to determine particular problems and differences of CFD codes simulating an expanded corium and from this, requirements for a reasonable measurement of experiments, that will be performed later. First the finite-volume-codes Comet 1.023, CFX 4.2 and CFX-TASCflow were used. To be able to make comparisons to a finite-element-code, now calculations are performed at the Institute of Safety Research at the Forschungszentrum Rossendorf with the code ANSYS/FLOTRAN. For the benchmark calculations of stage 1 a pure and liquid melt with internal heat sources was assumed uniformly distributed over the area of the planned core catcher of a EPR plant. Using the Standard-k-ε-turbulence model and assuming an initial state of a motionless superheated melt several large convection rolls will establish within the melt pool. The temperatures at the surface do not sink to a solidification level due to the enhanced convection heat transfer. The temperature gradients at the surface are relatively flat while there are steep gradients at the ground where the no slip condition is applied. But even at the ground no solidification temperatures are observed. Although the problem in the ANSYS-calculations is handled two-dimensional and not three-dimensional like in the finite-volume-codes, there are no fundamental deviations to the results of the other codes. (orig.)

BN-600 MOX Core Benchmark Analysis. Results from Phases 4 and 6 of a Coordinated Research Project on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects

International Nuclear Information System (INIS)

2013-12-01

For those Member States that have or have had significant fast reactor development programmes, it is of utmost importance that they have validated up to date codes and methods for fast reactor physics analysis in support of R and D and core design activities in the area of actinide utilization and incineration. In particular, some Member States have recently focused on fast reactor systems for minor actinide transmutation and on cores optimized for consuming rather than breeding plutonium; the physics of the breeder reactor cycle having already been widely investigated. Plutonium burning systems may have an important role in managing plutonium stocks until the time when major programmes of self-sufficient fast breeder reactors are established. For assessing the safety of these systems, it is important to determine the prediction accuracy of transient simulations and their associated reactivity coefficients. In response to Member States' expressed interest, the IAEA sponsored a coordinated research project (CRP) on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects. The CRP started in November 1999 and, at the first meeting, the members of the CRP endorsed a benchmark on the BN-600 hybrid core for consideration in its first studies. Benchmark analyses of the BN-600 hybrid core were performed during the first three phases of the CRP, investigating different nuclear data and levels of approximation in the calculation of safety related reactivity effects and their influence on uncertainties in transient analysis prediction. In an additional phase of the benchmark studies, experimental data were used for the verification and validation of nuclear data libraries and methods in support of the previous three phases. The results of phases 1, 2, 3 and 5 of the CRP are reported in IAEA-TECDOC-1623, BN-600 Hybrid Core Benchmark Analyses, Results from a Coordinated Research Project on Updated Codes and Methods to Reduce the

Benchmark exercises on PWR level-1 PSA (step 3). Analyses of accident sequence and conclusions

International Nuclear Information System (INIS)

Niwa, Yuji; Takahashi, Hideaki.

1996-01-01

The results of level 1 PSA generate fluctuations due to the assumptions based on several engineering judgements set in the stages of PSA analysis. On the purpose of the investigation of uncertainties due to assumptions, three kinds of a standard problem, what we call benchmark exercise have been set. In this report, sensitivity studies (benchmark exercise) of sequence analyses are treated and conclusions are mentioned. The treatment of inter-system dependency would generate uncertainly of PSA. In addition, as a conclusion of the PSA benchmark exercise, several findings in the sequence analysis together with previous benchmark analyses in earlier INSS Journals are treated. (author)

Analysis on First Criticality Benchmark Calculation of HTR-10 Core

International Nuclear Information System (INIS)

Zuhair; Ferhat-Aziz; As-Natio-Lasman

2000-01-01

HTR-10 is a graphite-moderated and helium-gas cooled pebble bed reactor with an average helium outlet temperature of 700 o C and thermal power of 10 MW. The first criticality benchmark problem of HTR-10 in this paper includes the loading number calculation of nuclear fuel in the form of UO 2 ball with U-235 enrichment of 17% for the first criticality under the helium atmosphere and core temperature of 20 o C, and the effective multiplication factor (k eff ) calculation of full core (5 m 3 ) under the helium atmosphere and various core temperatures. The group constants of fuel mixture, moderator and reflector materials were generated with WlMS/D4 using spherical model and 4 neutron energy group. The critical core height of 150.1 cm obtained from CITATION in 2-D R-Z reactor geometry exists in the calculation range of INET China, JAERI Japan and BATAN Indonesia, and OKBM Russia. The k eff calculation result of full core at various temperatures shows that the HTR-10 has negative temperature coefficient of reactivity. (author)

Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

International Nuclear Information System (INIS)

Rimpault, G.

2004-01-01

Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO 2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

International Nuclear Information System (INIS)

Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M.; Parma, E.J.; Ball, R.M.; Hoovler, G.S.

1993-01-01

Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors

MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

Science.gov (United States)

Selcow, Elizabeth C.; Cerbone, Ralph J.; Ludewig, Hans; Mughabghab, Said F.; Schmidt, Eldon; Todosow, Michael; Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.

1993-01-01

Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors.

Reactor physics tests and benchmark analyses of STACY

International Nuclear Information System (INIS)

Miyoshi, Yoshinori; Umano, Takuya

1996-01-01

The Static Experiment Critical Facility, STACY in the Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF is a solution type critical facility to accumulate fundamental criticality data on uranyl nitrate solution, plutonium nitrate solution and their mixture. A series of critical experiments have been performed for 10 wt% enriched uranyl nitrate solution using a cylindrical core tank. In these experiments, systematic data of the critical height, differential reactivity of the fuel solution, kinetic parameter and reactor power were measured with changing the uranium concentration of the fuel solution from 313 gU/l to 225 gU/l. Critical data through the first series of experiments for the basic core are reported in this paper for evaluating the accuracy of the criticality safety calculation codes. Benchmark calculations of the neutron multiplication factor k eff for the critical condition were made using a neutron transport code TWOTRAN in the SRAC system and a continuous energy Monte Carlo code MCNP 4A with a Japanese evaluated nuclear data library, JENDL 3.2. (J.P.N.)

Preliminary analysis of the proposed BN-600 benchmark core

International Nuclear Information System (INIS)

John, T.M.

2000-01-01

The Indira Gandhi Centre for Atomic Research is actively involved in the design of Fast Power Reactors in India. The core physics calculations are performed by the computer codes that are developed in-house or by the codes obtained from other laboratories and suitably modified to meet the computational requirements. The basic philosophy of the core physics calculations is to use the diffusion theory codes with the 25 group nuclear cross sections. The parameters that are very sensitive is the core leakage, like the power distribution at the core blanket interface etc. are calculated using transport theory codes under the DSN approximations. All these codes use the finite difference approximation as the method to treat the spatial variation of the neutron flux. Criticality problems having geometries that are irregular to be represented by the conventional codes are solved using Monte Carlo methods. These codes and methods have been validated by the analysis of various critical assemblies and calculational benchmarks. Reactor core design procedure at IGCAR consists of: two and three dimensional diffusion theory calculations (codes ALCIALMI and 3DB); auxiliary calculations, (neutron balance, power distributions, etc. are done by codes that are developed in-house); transport theory corrections from two dimensional transport calculations (DOT); irregular geometry treated by Monte Carlo method (KENO); cross section data library used CV2M (25 group)

A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models

International Nuclear Information System (INIS)

Kliem, S.

1999-01-01

During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)

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

Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system

International Nuclear Information System (INIS)

Arigane, Kenji

1987-04-01

The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)

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)

Benchmarking von Krankenhausinformationssystemen – eine vergleichende Analyse deutschsprachiger Benchmarkingcluster

Directory of Open Access Journals (Sweden)

Jahn, Franziska

2015-08-01

Full Text Available Benchmarking is a method of strategic information management used by many hospitals today. During the last years, several benchmarking clusters have been established within the German-speaking countries. They support hospitals in comparing and positioning their information system’s and information management’s costs, performance and efficiency against other hospitals. In order to differentiate between these benchmarking clusters and to provide decision support in selecting an appropriate benchmarking cluster, a classification scheme is developed. The classification scheme observes both general conditions and examined contents of the benchmarking clusters. It is applied to seven benchmarking clusters which have been active in the German-speaking countries within the last years. Currently, performance benchmarking is the most frequent benchmarking type, whereas the observed benchmarking clusters differ in the number of benchmarking partners and their cooperation forms. The benchmarking clusters also deal with different benchmarking subjects. Assessing costs and quality application systems, physical data processing systems, organizational structures of information management and IT services processes are the most frequent benchmarking subjects. There is still potential for further activities within the benchmarking clusters to measure strategic and tactical information management, IT governance and quality of data and data-processing processes. Based on the classification scheme and the comparison of the benchmarking clusters, we derive general recommendations for benchmarking of hospital information systems.

Benchmark calculation for water reflected STACY cores containing low enriched uranyl nitrate solution

International Nuclear Information System (INIS)

Miyoshi, Yoshinori; Yamamoto, Toshihiro; Nakamura, Takemi

2001-01-01

In order to validate the availability of criticality calculation codes and related nuclear data library, a series of fundamental benchmark experiments on low enriched uranyl nitrate solution have been performed with a Static Experiment Criticality Facility, STACY in JAERI. The basic core composed of a single tank with water reflector was used for accumulating the systematic data with well-known experimental uncertainties. This paper presents the outline of the core configurations of STACY, the standard calculation model, and calculation results with a Monte Carlo code and JENDL 3.2 nuclear data library. (author)

Comparison of the results of the fifth dynamic AER benchmark-a benchmark for coupled thermohydraulic system/three-dimensional hexagonal kinetic core models

International Nuclear Information System (INIS)

Kliem, S.

1998-01-01

The fifth dynamic benchmark was defined at seventh AER-Symposium, held in Hoernitz, Germany in 1997. It is the first benchmark for coupled thermohydraulic system/three-dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one control rod group stucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Each participant used own best-estimate nuclear cross section data. Only the initial subcriticality at the beginning of the transient was given. Solutions were received from Kurchatov Institute Russia with the code BIPR8/ATHLET, VTT Energy Finland with HEXTRAN/SMABRE, NRI Rez Czech Republic with DYN3/ATHLET, KFKI Budapest Hungary with KIKO3D/ATHLET and from FZR Germany with the code DYN3D/ATHLET.In this paper the results are compared. Beside the comparison of global results, the behaviour of several thermohydraulic and neutron kinetic parameters is presented to discuss the revealed differences between the solutions.(Authors)

Solution of the 6th dynamic AER benchmark using the coupled core DYN3D/ATHLET

International Nuclear Information System (INIS)

Seidel, A.; Kliem, S.

2001-01-01

The 6 th dynamic benchmark is a logical continuation of the work to validate systematically coupled neutron kinetics/thermohydraulics code systems for the estimation of the transient behaviour of WWER type nuclear power plant which was started in the 5 th dynamic benchmark. This benchmark concerns a double ended break of the main steam line (asymmetrical MSLB) in a WWER plant. The core is at the end of first cycle in full power conditions. The asymmetric leak causes a different depressurization of all steam generators. New features in comparison to the 5 th dynamic benchmark were included: asymmetric operation of the feed water system, consideration of incomplete coolant mixing in the reactor vessel, and the definition of a fixed isothermal recriticality temperature for normalising the nuclear data (Authors)

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

Recriticality analyses for CAPRA cores

International Nuclear Information System (INIS)

Maschek, W.; Thiem, D.

1995-01-01

The first scoping calculation performed show that the energetics levels from recriticalities in CAPRA cores are in the same range as in conventional cores. However, considerable uncertainties exist and further analyses are necessary. Additional investigations are performed for the separation scenarios of fuel/steel/inert and matrix material as a large influence of these processes on possible ramp rates and kinetics parameters was detected in the calculations. (orig./HP)

Recriticality analyses for CAPRA cores

Energy Technology Data Exchange (ETDEWEB)

Maschek, W.; Thiem, D.

1995-08-01

The first scoping calculation performed show that the energetics levels from recriticalities in CAPRA cores are in the same range as in conventional cores. However, considerable uncertainties exist and further analyses are necessary. Additional investigations are performed for the separation scenarios of fuel/steel/inert and matrix material as a large influence of these processes on possible ramp rates and kinetics parameters was detected in the calculations. (orig./HP)

Benchmarking multi-dimensional large strain consolidation analyses

International Nuclear Information System (INIS)

Priestley, D.; Fredlund, M.D.; Van Zyl, D.

2010-01-01

Analyzing the consolidation of tailings slurries and dredged fills requires a more extensive formulation than is used for common (small strain) consolidation problems. Large strain consolidation theories have traditionally been limited to 1-D formulations. SoilVision Systems has developed the capacity to analyze large strain consolidation problems in 2 and 3-D. The benchmarking of such formulations is not a trivial task. This paper presents several examples of modeling large strain consolidation in the beta versions of the new software. These examples were taken from the literature and were used to benchmark the large strain formulation used by the new software. The benchmarks reported here are: a comparison to the consolidation software application CONDES0, Townsend's Scenario B and a multi-dimensional analysis of long-term column tests performed on oil sands tailings. All three of these benchmarks were attained using the SVOffice suite. (author)

An analytical model for the study of a small LFR core dynamics: development and benchmark

International Nuclear Information System (INIS)

Bortot, S.; Cammi, A.; Lorenzi, S.; Moisseytsev, A.

2011-01-01

An analytical model for the study of a small Lead-cooled Fast Reactor (LFR) control-oriented dynamics has been developed aimed at providing a useful, very flexible and straightforward, though accurate, tool allowing relatively quick transient design-basis and stability analyses. A simplified lumped-parameter approach has been adopted to couple neutronics and thermal-hydraulics: the point-kinetics approximation has been employed and an average-temperature heat-exchange model has been implemented. The reactor transient responses following postulated accident initiators such as Unprotected Control Rod Withdrawal (UTOP), Loss of Heat Sink (ULOHS) and Loss of Flow (ULOF) have been studied for a MOX and a metal-fuelled core at the Beginning of Cycle (BoC) and End of Cycle (EoC) configurations. A benchmark analysis has been then performed by means of the SAS4A/SASSYS-1 Liquid Metal Reactor Code System, in which a core model based on three representative channels has been built with the purpose of providing verification for the analytical outcomes and indicating how the latter relate to more realistic one-dimensional calculations. As a general result, responses concerning the main core characteristics (namely, power, reactivity, etc.) have turned out to be mutually consistent in terms of both steady-state absolute figures and transient developments, showing discrepancies of the order of only some percents, thus confirming a very satisfactory agreement. (author)

Coupled fast-thermal core 'HERBE', as the benchmark experiment at the RB reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-10-01

Validation of the well-known Monte Carlo code MCNP TM against measured criticality data for the coupled fast-thermal HERBE. System at the RB research reactor is shown in this paper. Experimental data are obtained for regular HERBE core and for the cases of controlled flooding of the neutron converter zone by heavy water. Earlier calculations of these criticality parameters, done by combination of transport and diffusion codes using 2D geometry model are also compared to new calculations carried out by the MCNP code in 3D geometry, applying new detailed 3D model of the HEU fuel slug, developed recently. Satisfactory agreements in comparison of the HERBE criticality calculation results with experimental data, in spite complex heterogeneous composition of the HERBE core, are obtained and confirmed that HERBE core could be used as a criticality benchmark for coupled fast-thermal core. (author)

PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

Energy Technology Data Exchange (ETDEWEB)

Kuijper, J.C.

1994-10-01

This report contains the results of PANTHER calculations for the ``NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power``. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.).

Final PANTHER solution to the NEA-NSC3-DPWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

International Nuclear Information System (INIS)

Kuijper, J.C.

1996-10-01

This report contains the final results of PANTHER calculations for the 'NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power'. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.)

PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

International Nuclear Information System (INIS)

Kuijper, J.C.

1994-10-01

This report contains the results of PANTHER calculations for the ''NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power''. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.)

Final PANTHER solution to the NEA-NSC3-DPWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

Energy Technology Data Exchange (ETDEWEB)

Kuijper, J.C.

1996-10-01

This report contains the final results of PANTHER calculations for the `NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power`. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.).

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)

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)

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

VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS

Directory of Open Access Journals (Sweden)

Tagor Malem Sembiring

2015-10-01

Full Text Available ABSTRACT VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS. The coupled neutronic and thermal-hydraulic (T/H code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR ejection at peripheral core using a full core geometry model, the C1 and C2 cases.  By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM and the improved quasistatic method (IQS. All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16% occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4% for C2 case.  All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. Keywords: nodal method, coupled neutronic and thermal-hydraulic code, PWR, transient case, control rod ejection.   ABSTRAK VALIDASI MODEL GEOMETRI TERAS PENUH PAKET PROGRAM NODAL3 DALAM PROBLEM BENCHMARK GAYUT WAKTU PWR. Paket program kopel neutronik dan termohidraulika (T/H, NODAL3, telah divalidasi dengan beberapa kasus benchmark statis PWR dan kasus benchmark gayut waktu PWR NEACRP.  Akan tetapi, paket program NODAL3 belum divalidasi dalam kasus benchmark gayut waktu akibat penarikan sebuah perangkat batang kendali (CR di tepi teras menggunakan model geometri teras penuh, yaitu kasus C1 dan C2. Dengan penelitian ini, akurasi paket program

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.

JNC results of BN-600 benchmark calculation (phase 4)

International Nuclear Information System (INIS)

Ishikawa, Makoto

2003-01-01

The present work is the results of JNC, Japan, for the Phase 4 of the BN-600 core benchmark problem (Hex-Z fully MOX fuelled core model) organized by IAEA. The benchmark specification is based on 1) the RCM report of IAEA CRP on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of LMFR Reactivity Effects, Action 3.12' (Calculations for BN-600 fully fuelled MOX core for subsequent transient analyses). JENDL-3.2 nuclear data library was used for calculating 70 group ABBN-type group constants. Cell models for fuel assembly and control rod calculations were applied: homogeneous and heterogeneous (cylindrical supercell) model. Basic diffusion calculation was three-dimensional Hex-Z model, 18 group (Citation code). Transport calculations were 18 group, three-dimensional (NSHEC code) based on Sn-transport nodal method developed at JNC. The generated thermal power per fission was based on Sher's data corrected on the basis of ENDF/B-IV data library. Calculation results are presented in Tables for intercomparison

Benchmarking and qualification of the nufreq-npw code for best estimate prediction of multi-channel core stability margins

International Nuclear Information System (INIS)

Taleyarkhan, R.; McFarlane, A.F.; Lahey, R.T. Jr.; Podowski, M.Z.

1988-01-01

The work described in this paper is focused on the development, verification and benchmarking of the NUFREQ-NPW code at Westinghouse, USA for best estimate prediction of multi-channel core stability margins in US BWRs. Various models incorporated into NUFREQ-NPW are systematically compared against the Westinghouse channel stability analysis code MAZDA, which the Mathematical Model was developed in an entirely different manner. The NUFREQ-NPW code is extensively benchmarked against experimental stability data with and without nuclear reactivity feedback. Detailed comparisons are next performed against nuclear-coupled core stability data. A physically based algorithm is developed to correct for the effect of flow development on subcooled boiling. Use of this algorithm (to be described in the full paper) captures the peak magnitude as well as the resonance frequency with good accuracy

An analysis of the CSNI/GREST core concrete interaction chemical thermodynamic benchmark exercise using the MPEC2 computer code

International Nuclear Information System (INIS)

Muramatsu, Ken; Kondo, Yasuhiko; Uchida, Masaaki; Soda, Kunihisa

1989-01-01

Fission product (EP) release during a core concrete interaction (CCI) is an important factor of the uncertainty associated with a source term estimation for an LWR severe accident. An analysis was made on the CCI Chemical Thermodynamic Benchmark Exercise organized by OECD/NEA/CSNI Group of Experts on Source Terms (GREST) for investigating the uncertainty in thermodynamic modeling for CCI. The benchmark exercise was to calculate the equilibrium FP vapor pressure for given system of temperature, pressure, and debris composition. The benchmark consisted of two parts, A and B. Part A was a simplified problem intended to test the numerical techniques. In part B, the participants were requested to use their own best estimate thermodynamic data base to examine the variability of the results due to the difference in thermodynamic data base. JAERI participated in this benchmark exercise with use of the MPEC2 code. Chemical thermodynamic data base needed for analysis of Part B was taken from the VENESA code. This report describes the computer code used, inputs to the code, and results from the calculation by JAERI. The present calculation indicates that the FP vapor pressure depends strongly on temperature and Oxygen potential in core debris and the pattern of dependency may be different for different FP elements. (author)

NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

Directory of Open Access Journals (Sweden)

Surian Pinem

2016-01-01

Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

Validation of full core geometry model of the NODAL3 code in the PWR transient Benchmark problems

International Nuclear Information System (INIS)

T-M Sembiring; S-Pinem; P-H Liem

2015-01-01

The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16 % occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4 % for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. (author)

48{sup th} Annual meeting on nuclear technology (AMNT 2017). Key topic / Enhanced safety and operation excellence. Focus session: Uncertainty analyses in reactor core simulations

Energy Technology Data Exchange (ETDEWEB)

Zwermann, Winfried [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Forschungszentrum

2017-12-15

The supplementation of reactor simulations by uncertainty analyses is becoming increasingly important internationally due to the fact that the reliability of simulation calculations can be significantly increased by the quantification of uncertainties in comparison to the use of so-called conservative methods (BEPU- ''Best-Estimate plus Uncertainties''). While systematic uncertainty analyses for thermo-hydraulic calculations have been performed routinely for a long time, methods for taking into account uncertainties in nuclear data, which are the basis for neutron transport calculations, are under development. The Focus Session Uncertainty Analyses in Reactor Core Simulations was intended to provide an overview of international research and development with respect to supplementing reactor core simulations with uncertainty and sensitivity analyses, in research institutes as well as within the nuclear industry. The presented analyses not only focused on light water reactors, but also on advanced reactor systems. Particular emphasis was put on international benchmarks in the field. The session was chaired by Winfried Zwermann (Gesellschaft fuer Anlagen- und Reaktorsicherheit).

MC21/CTF and VERA multiphysics solutions to VERA core physics benchmark progression problems 6 and 7

Directory of Open Access Journals (Sweden)

Daniel J. Kelly, III

2017-09-01

Full Text Available The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated good agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle hot full power equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics/subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of 0.99994 ± 6.8E-6 (95% confidence interval. Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.

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

Application of the SPH method in nodal diffusion analyses of SFR cores

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety; Mikityuk, K. [Paul Scherrer Institut, Villigen (Switzerland)

2016-07-01

The current study investigated the potential of the SPH method, applied to correct the few-group XS produced by Serpent, to further improve the accuracy of the nodal diffusion solutions. The procedure for the generation of SPH-corrected few-group XS is presented in the paper. The performance of the SPH method was tested on a large oxide SFR core from the OECD/NEA SFR benchmark. The reference SFR core was modeled with the DYN3D and PARCS nodal diffusion codes using the SPH-corrected few-group XS generated by Serpent. The nodal diffusion results obtained with and without SPH correction were compared to the reference full-core Serpent MC solution. It was demonstrated that the application of the SPH method improves the accuracy of the nodal diffusion solutions, particularly for the rodded core state.

Selection and benchmarking of computer codes for research reactor core conversions

International Nuclear Information System (INIS)

Yilmaz, E.; Jones, B.G.

1983-01-01

A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC 2 , COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of Illinois. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general

A Benchmark Study of a Seismic Analysis Program for a Single Column of a HTGR Core

Energy Technology Data Exchange (ETDEWEB)

Kang, Ji Ho [KAERI, Daejeon (Korea, Republic of)

2016-05-15

A seismic analysis program, SAPCOR (Seismic Analysis of Prismatic HTGR Core), was developed in Korea Atomic Energy Research Institute. The program is used for the evaluation of deformed shapes and forces on the graphite blocks which using point-mass rigid bodies with Kelvin-Voigt impact models. In the previous studies, the program was verified using theoretical solutions and benchmark problems. To validate the program for more complicated problems, a free vibration analysis of a single column of a HTGR core was selected and the calculation results of the SAPCOR and a commercial FEM code, Abaqus, were compared in this study.

Research coordination meeting of the coordinated research project on analytical and experimental benchmark analyses of accelerator driven systems. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-07-01

The Technical Meeting hosted at the Belarus National Academy of Sciences in Minsk by the Joint Institute of Power Engineering and Nuclear Research 'SOSNY' from 5-9 December 2005 was the kick-off Research Coordination Meeting (RCM) of the IAEA Coordinated Research Project (CRP) on 'Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems (ADS)'. The CRP had received proposals for research agreements and contracts from scientists representing the following 25 institutions: Centro Atomico Bariloche, SCK CEN Mol, Instituto de Pesquisas Energeticas e Nucleares Sao Paulo, Joint Institute of Power Engineering and Nuclear Research SOSNY Minsk, China Institute of Atomic Energy, CEA Cadarache, CNRS Paris, FZ Rossendorf, FZ Karlsruhe, Budapest University of Technology and Economics, Politecnico di Torino, Japan Atomic Energy Agency, Nuclear Research and Consultancy Group (NRG) Petten, Pakistan Institute of Nuclear Science and Technology, AGH-University of Science and Technology Krakow, Institute of Atomic Energy Otwock/Swierk, ITEP Moscow, MEPHI Moscow, Kurchatov Institute, JINR Dubna, Universidad Politecnica de Madrid, CIEMAT Madrid, Royal Institute of Technology Stockholm, National Science Center 'Kharkov Institute and Technology', and Argonne National Laboratory). These institutions represent 18 IAEA Member States (i.e., Argentina, Belarus, Belgium, Brazil, China, France, Germany, Hungary, Italy, Japan, Netherlands, Pakistan, Poland, Russia, Spain, Sweden, Ukraine, USA), and one International Organization (JINR Dubna). The overall objective of the CRP is contributing to the generic R and 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

OECD/NRC BWR Turbine Trip Benchmark: Simulation by POLCA-T Code

International Nuclear Information System (INIS)

Panayotov, Dobromir

2004-01-01

Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and three-dimensional (3-D) neutron kinetics core models. Participation in the OECD/NRC BWR Turbine Trip (TT) Benchmark is a part of our efforts toward the code's validation. The paper describes the objectives for TT analyses and gives a brief overview of the developed plant system input deck and 3-D core model.The results of exercise 1, system model without netronics, are presented. Sensitivity studies performed cover the maximal time step, turbine stop valve position and mass flow, feedwater temperature, and steam bypass mass flow. Results of exercise 2, 3-D core neutronic and thermal-hydraulic model with boundary conditions, are also presented. Sensitivity studies include the core inlet temperature, cladding properties, and direct heating to core coolant and bypass.The entire plant model was validated in the framework of the benchmark's phase 3. Sensitivity studies include the effect of SCRAM initialization and carry-under. The results obtained - transient fission power and its initial axial distribution and steam dome, core exit, lower and upper plenum, main steam line, and turbine inlet pressures - showed good agreement with measured data. Thus, the POLCA-T code capabilities for correct simulation of pressurizing transients with very fast power were proved

BWR core melt progression phenomena: Experimental analyses

International Nuclear Information System (INIS)

Ott, L.J.

1992-01-01

In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

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

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)

Benchmark test of JENDL-3T and -3T/Rev.1

International Nuclear Information System (INIS)

Takano, Hideki; Kaneko, Kunio.

1989-10-01

The fast reactor 70-group constant set JFS-3-J3T has been generated by using the JENDL-3T nuclear data. One-dimensional 21-benchmark cores and the ZPPR-9 core were analysed with the JFS-3-J3T set. The results obtained are summarized as follows: (1) The values of keff are underestimated by 0.6% for Pu-fueled cores and overestimated by 2% for U-fueled cores. (2) The central reaction rate ratio 239 σ f φ/ 235 σ f φ is in a good agreement with the experimental value, though 238 σ c φ/ 239 σ f φ and 238 σ f φ/ 235 σ f φ are overestimated. (3) Doppler and Na-void reactivities are in a good agreement with the measured data. (4) The prediction accuracy of radial reaction rate distributions are improved in the comparison of the results obtained with the JENDL-2 data. Furthermore, the benchmark test of JENDL-3T/Rev. 1 which was revised from JENDL-3T for several important nuclides has been again performed. It was shown that JENDL-3T/Rev. 1 would predict nuclear characteristics more satisfactorily than JENDL-3T. (author)

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

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.

Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

Energy Technology Data Exchange (ETDEWEB)

Zwermann, W.; Aures, A.; Bernnat, W.; and others

2013-06-15

This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

Selection and benchmarking of computer codes for research reactor core conversions

Energy Technology Data Exchange (ETDEWEB)

Yilmaz, Emin [School of Aerospace, Mechanical and Nuclear Engineering, University of Oklahoma, Norman, OK (United States); Jones, Barclay G [Nuclear Engineering Program, University of IL at Urbana-Champaign, Urbana, IL (United States)

1983-09-01

A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC{sup 2}, COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of IL. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general. Deviation of k{sub eff} is about 0.5% for both enrichments at the beginning of life (BOL) and at the end of life (EOL). Flux ratios deviate only about 1.5% from IAEA contributor's mean value. (author)

Selection and benchmarking of computer codes for research reactor core conversions

International Nuclear Information System (INIS)

Yilmaz, Emin; Jones, Barclay G.

1983-01-01

A group of computer codes have been selected and obtained from the Nuclear Energy Agency (NEA) Data Bank in France for the core conversion study of highly enriched research reactors. ANISN, WIMSD-4, MC 2 , COBRA-3M, FEVER, THERMOS, GAM-2, CINDER and EXTERMINATOR were selected for the study. For the final work THERMOS, GAM-2, CINDER and EXTERMINATOR have been selected and used. A one dimensional thermal hydraulics code also has been used to calculate temperature distributions in the core. THERMOS and CINDER have been modified to serve the purpose. Minor modifications have been made to GAM-2 and EXTERMINATOR to improve their utilization. All of the codes have been debugged on both CDC and IBM computers at the University of IL. IAEA 10 MW Benchmark problem has been solved. Results of this work has been compared with the IAEA contributor's results. Agreement is very good for highly enriched fuel (HEU). Deviations from IAEA contributor's mean value for low enriched fuel (LEU) exist but they are small enough in general. Deviation of k eff is about 0.5% for both enrichments at the beginning of life (BOL) and at the end of life (EOL). Flux ratios deviate only about 1.5% from IAEA contributor's mean value. (author)

HS06 Benchmark for an ARM Server

Science.gov (United States)

Kluth, Stefan

2014-06-01

We benchmarked an ARM cortex-A9 based server system with a four-core CPU running at 1.1 GHz. The system used Ubuntu 12.04 as operating system and the HEPSPEC 2006 (HS06) benchmarking suite was compiled natively with gcc-4.4 on the system. The benchmark was run for various settings of the relevant gcc compiler options. We did not find significant influence from the compiler options on the benchmark result. The final HS06 benchmark result is 10.4.

HS06 benchmark for an ARM server

International Nuclear Information System (INIS)

Kluth, Stefan

2014-01-01

We benchmarked an ARM cortex-A9 based server system with a four-core CPU running at 1.1 GHz. The system used Ubuntu 12.04 as operating system and the HEPSPEC 2006 (HS06) benchmarking suite was compiled natively with gcc-4.4 on the system. The benchmark was run for various settings of the relevant gcc compiler options. We did not find significant influence from the compiler options on the benchmark result. The final HS06 benchmark result is 10.4.

HTR core physics analysis at NRG

International Nuclear Information System (INIS)

Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

2002-01-01

Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

Benchmark analyses for EBR-II shutdown heat removal tests SHRT-17 and SHRT-45R

Energy Technology Data Exchange (ETDEWEB)

Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui (Japan); Muranaka, Kohmei; Asai, Takayuki [Graduate School of Engineering, University of Fukui (Japan); Rooijen, W.F.G. van, E-mail: rooijen@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui (Japan)

2014-08-15

Highlights: • The IAEA EBR-II benchmarks SHRT-17 and SHRT-45R are analyzed with a 1D system code. • The calculated result of SHRT-17 corresponds well to the measured results. • For SHRT-45R ERANOS is used for various core parameters and reactivity coefficients. • SHRT-45R peak temperature is overestimated with the ERANOS feedback coefficients. • The peak temperature is well predicted when the feedback coefficient is reduced. - Abstract: Benchmark problems of several experiments in EBR-II, proposed by ANL and coordinated by the IAEA, are analyzed using the plant system code NETFLOW++ and the neutronics code ERANOS. The SHRT-17 test conducted as a loss-of-flow test is calculated using only the NETFLOW++ code because it is a purely thermal–hydraulic problem. The measured data were made available to the benchmark participants after the results of the blind benchmark calculations were submitted. Our work shows that major parameters of the plant are predicted with good accuracy. The SHRT-45R test, an unprotected loss of flow test is calculated using the NETFLOW++ code with the aid of delayed neutron data and reactivity coefficients calculated by the ERANOS code. These parameters are used in the NETFLOW++ code to perform a semi-coupled analysis of the neutronics – thermal–hydraulic problem. The measured data are compared with our calculated results. In our work, the peak temperature is underestimated, indicating that the reactivity feedback coefficients are too strong. When the reactivity feedback coefficient for thermal expansion is adjusted, good agreement is obtained in general for the calculated plant parameters, with a few exceptions.

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

MC21 Monte Carlo analysis of the Hoogenboom-Martin full-core PWR benchmark problem - 301

International Nuclear Information System (INIS)

Kelly, D.J.; Sutton, Th.M.; Trumbull, T.H.; Dobreff, P.S.

2010-01-01

At the 2009 American Nuclear Society Mathematics and Computation conference, Hoogenboom and Martin proposed a full-core PWR model to monitor the improvement of Monte Carlo codes to compute detailed power density distributions. This paper describes the application of the MC21 Monte Carlo code to the analysis of this benchmark model. With the MC21 code, we obtained detailed power distributions over the entire core. The model consisted of 214 assemblies, each made up of a 17x17 array of pins. Each pin was subdivided into 100 axial nodes, thus resulting in over seven million tally regions. Various cases were run to assess the statistical convergence of the model. This included runs of 10 billion and 40 billion neutron histories, as well as ten independent runs of 4 billion neutron histories each. The 40 billion neutron-history calculation resulted in 43% of all regions having a 95% confidence level of 2% or less implying a relative standard deviation of 1%. Furthermore, 99.7% of regions having a relative power density of 1.0 or greater have a similar confidence level. We present timing results that assess the MC21 performance relative to the number of tallies requested. Source convergence was monitored by analyzing plots of the Shannon entropy and eigenvalue versus active cycle. We also obtained an estimate of the dominance ratio. Additionally, we performed an analysis of the error in an attempt to ascertain the validity of the confidence intervals predicted by MC21. Finally, we look forward to the prospect of full core 3-D Monte Carlo depletion by scoping out the required problem size. This study provides an initial data point for the Hoogenboom-Martin benchmark model using a state-of-the-art Monte Carlo code. (authors)

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

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

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)

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)

Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 5. Analyses of the OECD MSLB Benchmark with the Codes DYN3D and DYN3D/ATHLET

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.

2001-01-01

The code DYN3D coupled with ATHLET was used for the analysis of the OECD Main-Steam-Line-Break (MSLB) Benchmark, which is based on real plant design and operational data of the TMI-1 pressurized water reactor (PWR). Like the codes RELAP or TRAC,ATHLET is a thermal-hydraulic system code with point or one-dimensional neutron kinetic models. ATHLET, developed by the Gesellschaft for Anlagen- und Reaktorsicherheit, is widely used in Germany for safety analyses of nuclear power plants. DYN3D consists of three-dimensional nodal kinetic models and a thermal-hydraulic part with parallel coolant channels of the reactor core. DYN3D was coupled with ATHLET for analyzing more complex transients with interactions between coolant flow conditions and core behavior. It can be applied to the whole spectrum of operational transients and accidents, from small and intermediate leaks to large breaks of coolant loops or steam lines at PWRs and boiling water reactors. The so-called external coupling is used for the benchmark, where the thermal hydraulics is split into two parts: DYN3D describes the thermal hydraulics of the core, while ATHLET models the coolant system. Three exercises of the benchmark were simulated: Exercise 1: point kinetics plant simulation (ATHLET) Exercise 2: coupled three-dimensional neutronics/core thermal-hydraulics evaluation of the core response for given core thermal-hydraulic boundary conditions (DYN3D) Exercise 3: best-estimate coupled core-plant transient analysis (DYN3D/ATHLET). Considering the best-estimate cases (scenarios 1 of exercises 2 and 3), the reactor does not reach criticality after the reactor trip. Defining more serious tests for the codes, the efficiency of the control rods was decreased (scenarios 2 of exercises 2 and 3) to obtain recriticality during the transient. Besides the standard simulation given by the specification, modifications are introduced for sensitivity studies. The results presented here show (a) the influence of a reduced

Spectrum integrated (n,He) cross section comparisons and least squares analyses for 6Li and 10B in benchmark fields

International Nuclear Information System (INIS)

Schenter, R.E.; Oliver, B.M.; Farrar, H. IV.

1986-06-01

Spectrum integrated cross sections for 6 Li and 10 B from five benchmark fast reactor neutron fields are compared with calculated values obtained using the ENDF/B-V Cross Section Files. The benchmark fields include the Coupled Fast Reactivity Measurements Facility (CFRMF) at the Idaho National Engineering Laboratory, the 10% Enriched U-235 Critical Assembly (BIG-10) at Los Alamos National Laboratory, the Sigma-Sigma and Fission Cavity fields of the BR-1 reactor at CEN/SCK, and the Intermediate Energy Standard Neutron Field (ISNF) at the National Bureau of Standards. Results from least square analyses using the FERRET computer code to obtain adjusted cross section values and their uncertainties are presented. Input to these calculations include the above five benchmark data sets. These analyses indicate a need for revision in the ENDF/B-V files for the 10 B and 6 Li cross sections for energies above 50 keV

Verification of MVP-II and SRAC2006 code to the core physics vera benchmark problem

International Nuclear Information System (INIS)

Jati Susilo

2014-01-01

In this research, verification calculation for VERA core physics benchmark on the Zero Power Physical Test (ZPPT) of the nuclear reactor Watts Bar 1. The reactor is a 1000 MWe class of PWR designed by. Westinghouse, arranged from 193 unit of 17 x 17 fuel assembly consisting 3 type enrichment of UO2 that are 2.1wt%, 2.619wt% and 3.1wt%. Core power factor distribution and k-eff calculation has been done for the first cycle operation of the core at beginning of cycle (BOC) and hot zero power (HZP). In this calculation, MVP-II and CITATION module of SRAC2006 computer code has been used with ENDF/B-VII.0. cross section data library. Calculation result showed that differences value of k-eff for the core at controlled and uncontrolled condition between reference with MVP-II (-0,07% and -0,014%) and SRAC2006 (0,92% and 0,99%) are very small or below 1%. Differences value of radial power peaking factor at controlled and uncontrolled of the core between reference value with MVP-II are 0,38% and 1,53%, even though with SRAC2006 are 1,13% and -2,45%. It can be said that the calculation result by both computer code showing suitability with reference value. In order to determinate of criticality of the core, the calculation result using MVP-II code is more conservative compare with SRAC2006 code. (author)

3-D core modelling of RIA transient: the TMI-1 benchmark

International Nuclear Information System (INIS)

Ferraresi, P.; Studer, E.; Avvakumov, A.; Malofeev, V.; Diamond, D.; Bromley, B.

2001-01-01

The increase of fuel burn up in core management poses actually the problem of the evaluation of the deposited energy during Reactivity Insertion Accidents (RIA). In order to precisely evaluate this energy, 3-D approaches are used more and more frequently in core calculations. This 'best-estimate' approach requires the evaluation of code uncertainties. To contribute to this evaluation, a code benchmark has been launched. A 3-D modelling for the TMI-1 central Ejected Rod Accident with zero and intermediate initial powers was carried out with three different methods of calculation for an inserted reactivity respectively fixed at 1.2 $ and 1.26 $. The studies implemented by the neutronics codes PARCS (BNL) and CRONOS (IPSN/CEA) describe an homogeneous assembly, whereas the BARS (KI) code allows a pin-by-pin representation (CRONOS has both possibilities). All the calculations are consistent, the variation in figures resulting mainly from the method used to build cross sections and reflectors constants. The maximum rise in enthalpy for the intermediate initial power (33 % P N ) calculation is, for this academic calculation, about 30 cal/g. This work will be completed in a next step by an evaluation of the uncertainty induced by the uncertainty on model parameters, and a sensitivity study of the key parameters for a peripheral Rod Ejection Accident. (authors)

3-D core modelling of RIA transient: the TMI-1 benchmark

Energy Technology Data Exchange (ETDEWEB)

Ferraresi, P. [CEA Cadarache, Institut de Protection et de Surete Nucleaire, Dept. de Recherches en Securite, 13 - Saint Paul Lez Durance (France); Studer, E. [CEA Saclay, Dept. Modelisation de Systemes et Structures, 91 - Gif sur Yvette (France); Avvakumov, A.; Malofeev, V. [Nuclear Safety Institute of Russian Research Center, Kurchatov Institute, Moscow (Russian Federation); Diamond, D.; Bromley, B. [Nuclear Energy and Infrastructure Systems Div., Brookhaven National Lab., BNL, Upton, NY (United States)

2001-07-01

The increase of fuel burn up in core management poses actually the problem of the evaluation of the deposited energy during Reactivity Insertion Accidents (RIA). In order to precisely evaluate this energy, 3-D approaches are used more and more frequently in core calculations. This 'best-estimate' approach requires the evaluation of code uncertainties. To contribute to this evaluation, a code benchmark has been launched. A 3-D modelling for the TMI-1 central Ejected Rod Accident with zero and intermediate initial powers was carried out with three different methods of calculation for an inserted reactivity respectively fixed at 1.2 $ and 1.26 $. The studies implemented by the neutronics codes PARCS (BNL) and CRONOS (IPSN/CEA) describe an homogeneous assembly, whereas the BARS (KI) code allows a pin-by-pin representation (CRONOS has both possibilities). All the calculations are consistent, the variation in figures resulting mainly from the method used to build cross sections and reflectors constants. The maximum rise in enthalpy for the intermediate initial power (33 % P{sub N}) calculation is, for this academic calculation, about 30 cal/g. This work will be completed in a next step by an evaluation of the uncertainty induced by the uncertainty on model parameters, and a sensitivity study of the key parameters for a peripheral Rod Ejection Accident. (authors)

Gas cooled fast reactor benchmarks for JNC and Cea neutronic tools assessment

International Nuclear Information System (INIS)

Rimpault, G.; Sugino, K.; Hayashi, H.

2005-01-01

In order to verify the adequacy of JNC and Cea computational tools for the definition of GCFR (gas cooled fast reactor) core characteristics, GCFR neutronic benchmarks have been performed. The benchmarks have been carried out on two different cores: 1) a conventional Gas-Cooled fast Reactor (EGCR) core with pin-type fuel, and 2) an innovative He-cooled Coated-Particle Fuel (CPF) core. Core characteristics being studied include: -) Criticality (Effective multiplication factor or K-effective), -) Instantaneous breeding gain (BG), -) Core Doppler effect, and -) Coolant depressurization reactivity. K-effective and coolant depressurization reactivity at EOEC (End Of Equilibrium Cycle) state were calculated since these values are the most critical characteristics in the core design. In order to check the influence due to the difference of depletion calculation systems, a simple depletion calculation benchmark was performed. Values such as: -) burnup reactivity loss, -) mass balance of heavy metals and fission products (FP) were calculated. Results of the core design characteristics calculated by both JNC and Cea sides agree quite satisfactorily in terms of core conceptual design study. Potential features for improving the GCFR computational tools have been discovered during the course of this benchmark such as the way to calculate accurately the breeding gain. Different ways to improve the accuracy of the calculations have also been identified. In particular, investigation on nuclear data for steel is important for EGCR and for lumped fission products in both cores. The outcome of this benchmark is already satisfactory and will help to design more precisely GCFR cores. (authors)

Analysis and sensitivity studies with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

Energy Technology Data Exchange (ETDEWEB)

Ferroukhi, H.; Coddington, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

2001-07-01

The OECD/NEA PWR rod ejection benchmark has been analysed using the 3-D nodal spatial-kinetic codes CORETRAN and RETRAN-3D. The following results were obtained. A) The agreement in 3-D solution between CORETRAN and RETRAN-3D was found to be very good both during steady-state and transient conditions. In particular at HZP (hot zero power), an excellent agreement in the initial steady-state 3-D power distribution and with regard to the core power excursion during the super-prompt critical phase of the transient (i.e. when the negative reactivity feedback is still very weak) was found. This illustrates the consistency in the neutronic solution between both codes. B) At both HZP and FP (full power) conditions, the CORETRAN and RETRAN-3D results lie well within the range of the previous benchmark solutions. In particular at HZP, both codes predict a power excursion and an increase in maximum pellet temperature that are among the closest results to those obtained with the benchmark reference solution. It must here be emphasised that these analyses are by no means a validation of the codes. However, the good agreement of both CORETRAN and RETRAN-3D with other 3-D solutions provides confidence in the ability of these codes to analyse LWR (light water reactor) core transients. In addition, it was found appropriate to perform, for this well-defined international benchmark problem, some sensitivity studies in order to assess the impact of modelling options on the CORETRAN and RETRAN-3D results. (authors)

Analysis and sensitivity studies with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

International Nuclear Information System (INIS)

Ferroukhi, H.; Coddington, P.

2001-01-01

The OECD/NEA PWR rod ejection benchmark has been analysed using the 3-D nodal spatial-kinetic codes CORETRAN and RETRAN-3D. The following results were obtained. A) The agreement in 3-D solution between CORETRAN and RETRAN-3D was found to be very good both during steady-state and transient conditions. In particular at HZP (hot zero power), an excellent agreement in the initial steady-state 3-D power distribution and with regard to the core power excursion during the super-prompt critical phase of the transient (i.e. when the negative reactivity feedback is still very weak) was found. This illustrates the consistency in the neutronic solution between both codes. B) At both HZP and FP (full power) conditions, the CORETRAN and RETRAN-3D results lie well within the range of the previous benchmark solutions. In particular at HZP, both codes predict a power excursion and an increase in maximum pellet temperature that are among the closest results to those obtained with the benchmark reference solution. It must here be emphasised that these analyses are by no means a validation of the codes. However, the good agreement of both CORETRAN and RETRAN-3D with other 3-D solutions provides confidence in the ability of these codes to analyse LWR (light water reactor) core transients. In addition, it was found appropriate to perform, for this well-defined international benchmark problem, some sensitivity studies in order to assess the impact of modelling options on the CORETRAN and RETRAN-3D results. (authors)

Boiling water reactor turbine trip (TT) benchmark

International Nuclear Information System (INIS)

2001-06-01

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current nuclear applications Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for the purpose. The present volume describes the specification of such a benchmark. The transient addressed is a turbine trip (TT) in a BWR involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the plant make the present benchmark very valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4). (authors)

Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

Energy Technology Data Exchange (ETDEWEB)

Parish, T.A.

1995-03-02

This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

Boiling water reactor turbine trip (TT) benchmark

International Nuclear Information System (INIS)

2005-01-01

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present report is the second in a series of four and summarises the results of the first benchmark exercise, which identifies the key parameters and important issues concerning the thermalhydraulic system modelling of the transient, with specified core average axial power distribution and fission power time transient history. The transient addressed is a turbine trip in a boiling water reactor, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable. (author)

Development of the evaluation methods in reactor safety analyses and core characteristics

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

In order to support the safety reviews by NRA on reactor safety design including the phenomena with multiple failures, the computer codes are developed and the safety evaluations with analyses are performed in the areas of thermal hydraulics and core characteristics evaluation. In the code preparation of safety analyses, the TRACE and RELAP5 code were prepared to conduct the safety analyses of LOCA and beyond design basis accidents with multiple failures. In the core physics code preparation, the functions of sensitivity and uncertainty analysis were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

Verification of NUREC Code Transient Calculation Capability Using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem

International Nuclear Information System (INIS)

Joo, Hyung Kook; Noh, Jae Man; Lee, Hyung Chul; Yoo, Jae Woon

2006-01-01

In this report, we verified the NUREC code transient calculation capability using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem. The benchmark problem consists of Part 1, a 2-D problem with given T/H conditions, Part 2, a 3-D problem at HFP condition, Part 3, a 3-D problem at HZP condition, and Part 4, a transient state initiated by a control rod ejection at HZP condition in Part 3. In Part 1, the results of NUREC code agreed well with the reference solution obtained from DeCART calculation except for the pin power distributions at the rodded assemblies. In Part 2, the results of NUREC code agreed well with the reference DeCART solutions. In Part 3, some results of NUREC code such as critical boron concentration and core averaged delayed neutron fraction agreed well with the reference PARCS 2G solutions. But the error of the assembly power at the core center was quite large. The pin power errors of NUREC code at the rodded assemblies was much smaller the those of PARCS code. The axial power distribution also agreed well with the reference solution. In Part 4, the results of NUREC code agreed well with those of PARCS 2G code which was taken as the reference solution. From the above results we can conclude that the results of NUREC code for steady states and transient states of the MOX loaded LWR core agree well with those of the other codes

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

Validation and applicability of the 3D core kinetics and thermal hydraulics coupled code SPARKLE

International Nuclear Information System (INIS)

Miyata, Manabu; Maruyama, Manabu; Ogawa, Junto; Otake, Yukihiko; Miyake, Shuhei; Tabuse, Shigehiko; Tanaka, Hirohisa

2009-01-01

The SPARKLE code is a coupled code system based on three individual codes whose physical models have already been verified and validated. Mitsubishi Heavy Industries (MHI) confirmed the coupling calculation, including data transfer and the total reactor coolant system (RCS) behavior of the SPARKLE code. The confirmation uses the OECD/NEA MSLB benchmark problem, which is based on Three Mile Island Unit 1 (TMI-1) nuclear power plant data. This benchmark problem has been used to verify coupled codes developed and used by many organizations. Objectives of the benchmark program are as follows. Phase 1 is to compare the results of the system transient code using point kinetics. Phase 2 is to compare the results of the coupled three-dimensional (3D) core kinetics code and 3D core thermal-hydraulics (T/H) code, and Phase 3 is to compare the results of the combined coupled system transient code, 3D core kinetics code, and 3D core T/H code as a total validation of the coupled calculation. The calculation results of the SPARKLE code indicate good agreement with other benchmark participants' results. Therefore, the SPARKLE code is validated through these benchmark problems. In anticipation of applying the SPARKLE code to licensing analyses, MHI and Japanese PWR utilities have established a safety analysis method regarding the calculation conditions such as power distributions, reactivity coefficients, and event-specific features. (author)

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...

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....

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

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.

HTR core physics and transient analyses by the Panthermix code system

International Nuclear Information System (INIS)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

2005-01-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

Full sphere hydrodynamic and dynamo benchmarks

KAUST Repository

Marti, P.

2014-01-26

Convection in planetary cores can generate fluid flow and magnetic fields, and a number of sophisticated codes exist to simulate the dynamic behaviour of such systems. We report on the first community activity to compare numerical results of computer codes designed to calculate fluid flow within a whole sphere. The flows are incompressible and rapidly rotating and the forcing of the flow is either due to thermal convection or due to moving boundaries. All problems defined have solutions that alloweasy comparison, since they are either steady, slowly drifting or perfectly periodic. The first two benchmarks are defined based on uniform internal heating within the sphere under the Boussinesq approximation with boundary conditions that are uniform in temperature and stress-free for the flow. Benchmark 1 is purely hydrodynamic, and has a drifting solution. Benchmark 2 is a magnetohydrodynamic benchmark that can generate oscillatory, purely periodic, flows and magnetic fields. In contrast, Benchmark 3 is a hydrodynamic rotating bubble benchmark using no slip boundary conditions that has a stationary solution. Results from a variety of types of code are reported, including codes that are fully spectral (based on spherical harmonic expansions in angular coordinates and polynomial expansions in radius), mixed spectral and finite difference, finite volume, finite element and also a mixed Fourier-finite element code. There is good agreement between codes. It is found that in Benchmarks 1 and 2, the approximation of a whole sphere problem by a domain that is a spherical shell (a sphere possessing an inner core) does not represent an adequate approximation to the system, since the results differ from whole sphere results. © The Authors 2014. Published by Oxford University Press on behalf of The Royal Astronomical Society.

Development of concept and neutronic calculation method for large LMFBR core

International Nuclear Information System (INIS)

Shirakata, K.; Ishikawa, M.; Ikegami, T.; Sanda, T.; Kaneto, K.; Kawashima, M.; Kaise, Y.; Shirakawa, M.; Hibi, K.

1991-01-01

Presented in this paper is the state of the art of reactor physics R and Ds for the development of concept and neutronic calculation method for large Liquid Metal Fast Breeder Reactor (LMFBR) core. Physics characteristics of concepts for mixed oxide (MOX) fueled large FBR core were investigated by a series of benchmark critical experiments. Next, an adequacy and accuracy of the current neutronic calculation method was assessed by the experiments analyses, and then neutronic prediction accuracies by the method were evaluated for physics characteristics of the large core. Concerns on core development were discussed in terms of neutronics. (author)

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

The fifth AER dynamic benchmark calculation with hextran-smabre

International Nuclear Information System (INIS)

Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

1998-01-01

The first AER benchmark for coupling of the thermohydraulic codes and three-dimensional reactordynamic core models is discussed. HEXTRAN 2.7 is used for the core dynamics and SMABRE 4.6 as a thermohydraulic model for the primary and secondary loops. The plant model for SMABRE is based mainly on two input models, the Loviisa model and standard VVER-440/213 plant model. The primary circuit includes six separate loops, totally 505 nodes and 652 junctions. The reactor pressure vessel is divided into six parallel channels. In HEXTRAN calculation 1/6 symmetry is used in the core. In the calculations nuclear data is based on the ENDF/B-IV library and it has been evaluated with the CASMO-HEX code. The importance of the nuclear data was illustrated by repeating the benchmark calculation with using three different data sets. Optimal extensive data valid from hot to cold conditions were not available for all types of fuel enrichments needed in this benchmark. (author)

HTR core physics and transient analyses by the Panthermix code system

Energy Technology Data Exchange (ETDEWEB)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

2005-07-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

Benchmark calculations of power distribution within assemblies

International Nuclear Information System (INIS)

Cavarec, C.; Perron, J.F.; Verwaerde, D.; West, J.P.

1994-09-01

The main objective of this Benchmark is to compare different techniques for fine flux prediction based upon coarse mesh diffusion or transport calculations. We proposed 5 ''core'' configurations including different assembly types (17 x 17 pins, ''uranium'', ''absorber'' or ''MOX'' assemblies), with different boundary conditions. The specification required results in terms of reactivity, pin by pin fluxes and production rate distributions. The proposal for these Benchmark calculations was made by J.C. LEFEBVRE, J. MONDOT, J.P. WEST and the specification (with nuclear data, assembly types, core configurations for 2D geometry and results presentation) was distributed to correspondents of the OECD Nuclear Energy Agency. 11 countries and 19 companies answered the exercise proposed by this Benchmark. Heterogeneous calculations and homogeneous calculations were made. Various methods were used to produce the results: diffusion (finite differences, nodal...), transport (P ij , S n , Monte Carlo). This report presents an analysis and intercomparisons of all the results received

Benchmark tests for fast and thermal reactor applications

International Nuclear Information System (INIS)

Seki, Yuji

1984-01-01

Integral tests of JENDL-2 library for fast and thermal reactor applications are reviewed including relevant analyses of JUPITER experiments. Criticality and core center characteristics were tested with one-dimensional models for a total of 27 fast critical assemblies. More sofisticated problems such as reaction rate distributions, control rod worths and sodium void reactivities were tested using two-dimensional models for MOZART and ZPPR-3 assemblies. Main observations from the fast core benchmark tests are as follows. 1) The criticality is well predicted; the average C/E value is 0.999+-0.008 for uranium cores and 0.997+-0.005 for plutonium cores. 2) The calculation underpredicts the reaction rate ratio 239 Pusub(fis)/ 235 Usub(fis) by 3% and overpredicts 238 Usub(cap)/ 239 Pusub(fis) by 6%. The results are consistent with those of JUPITER analyses. 3) The reaction rate distributions in the cores of prototype size are well predicted within +-3%. In larger JUPITER cores, however, the C/E value increases with the radial distance from the core center up to 6% at the outer core edge. 4) The prediction of control rod worths is satisfactory; C/E values are within the range from 0.92 to 0.97 with no apparent dependence on 10 B enrichment and the number of control rods inserted. Spatial dependence of C/E is also observed in the JUPITER cores. 5) The sodium void reactivity is overpredicted by 30% to 50% to the positive side. 1) The criticality is well predicted, as is the same in the fast core tests; the average C/E is 0.997+-0.003. 2) The calculation overpredicts 238 Usub(fis)/ 235 Usub(fis) by 3% to 6%, which shows the same tendency as in the small and medium size fast assemblies. The 238 Usub(cap)/ 235 Usub(fis) ratio is well predicted in the thermal cores. The calculated reaction rate ratios of 232 Th deviate from the measurements by 10% to 15%. (author)

European benchmark on the ASTRID-like low-void-effect core characterization: neutronic parameters and safety coefficients - 15361

International Nuclear Information System (INIS)

Bortot, S.; Mikityuk, K.; Panadero, A.L.; Pelloni, S.; Alvarez-Velarde, F.; Lopez, D.; Fridman, E.; Cruzado, I.G.; Herranz, N.G.; Ponomarev, A.; Sciora, P.; Vasile, A.; Seubert, A.; Tsige-Tamirat, H.

2015-01-01

A neutronic benchmark was launched with the participation of 8 European institutions using 10 codes and 4 data libraries, in order to study the main characteristics of a low-void-effect sodium-cooled fast spectrum core similar to the one of ASTRID at End-Of-Cycle conditions. The first results of this exercise are presented in this paper. As a major outcome of the study, the negative reactivity effect ensuing from the total voiding of the core was unanimously confirmed. Moreover, the code-to-code comparison allowed identifying a number of issues that require further clarifications and improvements. Some of them are mentioned here. The power generation in the non-fuel regions of the core was calculated by only 2 codes and the resulting result discrepancies reach 100%. Unexpected large discrepancies (up to 100 pcm) were observed in the Doppler constants predictions. The deviation of the Doppler effect's temperature dependence from a logarithmic law is also worth additional analysis. A discrepancy between nuclear data libraries (particularly between JEFF 3.1 and ENDF/B-VII.0) was observed in particular for the prediction of the CR worth

CEA-IPSN Participation in the MSLB Benchmark

International Nuclear Information System (INIS)

Royer, E.; Raimond, E.; Caruge, D.

2001-01-01

The OECD/NEA Main Steam Line Break (MSLB) Benchmark allows the comparison of state-of-the-art and best-estimate models used to compute reactivity accidents. The three exercises of the MSLB benchmark are defined with the aim of analyzing the space and time effects in the core and their modeling with computational tools. Point kinetics (exercise 1) simulation results in a return to power (RTP) after scram, whereas 3-D kinetics (exercises 2 and 3) does not display any RTP. The objective is to understand the reasons for the conservative solution of point kinetics and to assess the benefits of best-estimate models. First, the core vessel mixing model is analyzed; second, sensitivity studies on point kinetics are compared to 3-D kinetics; third, the core thermal hydraulics model and coupling with neutronics is presented; finally, RTP and a suitable model for MSLB are discussed

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.

Present status and extensions of the Monte Carlo performance benchmark

International Nuclear Information System (INIS)

Hoogenboom, J.E.; Petrovic, B.; Martin, W.R.

2013-01-01

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. (authors)

EBR-II Reactor Physics Benchmark Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

Pope, Chad L. [Idaho State Univ., Pocatello, ID (United States); Lum, Edward S [Idaho State Univ., Pocatello, ID (United States); Stewart, Ryan [Idaho State Univ., Pocatello, ID (United States); Byambadorj, Bilguun [Idaho State Univ., Pocatello, ID (United States); Beaulieu, Quinton [Idaho State Univ., Pocatello, ID (United States)

2017-12-28

This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

An improved benchmark model for the Big Ten critical assembly - 021

International Nuclear Information System (INIS)

Mosteller, R.D.

2010-01-01

A new benchmark specification is developed for the BIG TEN uranium critical assembly. The assembly has a fast spectrum, and its core contains approximately 10 wt.% enriched uranium. Detailed specifications for the benchmark are provided, and results from the MCNP5 Monte Carlo code using a variety of nuclear-data libraries are given for this benchmark and two others. (authors)

Radiocarbon analyses along the EDML ice core in Antarctica

NARCIS (Netherlands)

van de Wal, R.S.W.; Meijer, H.A.J.; van Rooij, M.; van der Veen, C.

2007-01-01

Samples, 17 in total, from the EDML core drilled at Kohnen station Antarctica are analysed for 14CO and 14CO2 with a dry-extraction technique in combination with accelerator mass spectrometry. Results of the in situ produced 14CO fraction show a very low concentration of in situ produced 14CO.

Radiocarbon analyses along the EDML ice core in Antarctica

NARCIS (Netherlands)

Van de Wal, R. S. W.; Meijer, H. A. J.; De Rooij, M.; Van der Veen, C.

Samples, 17 in total, from the EDML core drilled at Kohnen station Antarctica are analysed for (CO)-C-14 and (CO2)-C-14 with a dry-extraction technique in combination with accelerator mass spectrometry. Results of the in situ produced (CO)-C-14 fraction show a very low concentration of in situ

VENUS-2 Benchmark Problem Analysis with HELIOS-1.9

International Nuclear Information System (INIS)

Jeong, Hyeon-Jun; Choe, Jiwon; Lee, Deokjung

2014-01-01

Since there are reliable results of benchmark data from the OECD/NEA report of the VENUS-2 MOX benchmark problem, by comparing benchmark results users can identify the credibility of code. In this paper, the solution of the VENUS-2 benchmark problem from HELIOS 1.9 using the ENDF/B-VI library(NJOY91.13) is compared with the result from HELIOS 1.7 with consideration of the MCNP-4B result as reference data. The comparison contains the results of pin cell calculation, assembly calculation, and core calculation. The eigenvalues from those are considered by comparing the results from other codes. In the case of UOX and MOX assemblies, the differences from the MCNP-4B results are about 10 pcm. However, there is some inaccuracy in baffle-reflector condition, and relatively large differences were found in the MOX-reflector assembly and core calculation. Although HELIOS 1.9 utilizes an inflow transport correction, it seems that it has a limited effect on the error in baffle-reflector condition

The code DYN3DR for steady-state and transient analyses of light water reactor cores with Cartesian geometry

International Nuclear Information System (INIS)

Grundmann, U.

1995-11-01

The code DYN3D/M2 was developed for 3-dimensional steady-state and transient analyses of reactor cores with hexagonal fuel assemblies. The neutron kinetics of the new version DYN3DR is based on a nodal method for the solution of the 3-dimensional 2-group neutron diffusion equation for Cartesian geometry. The thermal-hydraulic model FLOCAL simulating the two phase flow of coolant and the fuel rod behaviour is used in the two versions. The fundamentals for the solution of the neutron diffusion equations in DYN3DR are described. The 3-dimensional NEACRP benchmarks for rod ejections in LWR with quadratic fuel assemblies were calculated and the results were compared with the published solutions. The developed algorithm for neutron kinetics are suitable for using parallel processing. The behaviour of speed-up versus the number of processors is demonstrated for calculations of a static neutron flux distribution using a workstation with 4 processors. (orig.) [de

Adaption of the PARCS Code for Core Design Audit Analyses

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyong Chol; Lee, Young Jin; Uhm, Jae Beop; Kim, Hyunjik [Nuclear Safety Evaluation, Daejeon (Korea, Republic of); Jeong, Hun Young; Ahn, Seunghoon; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

The eigenvalue calculation also includes quasi-static core depletion analyses. PARCS has implemented variety of features and has been qualified as a regulatory audit code in conjunction with other NRC thermal-hydraulic codes such as TRACE or RELAP5. In this study, as an adaptation effort for audit applications, PARCS is applied for an audit analysis of a reload core design. The lattice physics code HELIOS is used for cross section generation. PARCS-HELIOS code system has been established as a core analysis tool. Calculation results have been compared on a wide spectrum of calculations such as power distribution, critical soluble boron concentration, and rod worth. A reasonable agreement between the audit calculation and the reference results has been found.

Summary of the First Workshop on OECD/NRC boiling water reactor turbine trip benchmark

International Nuclear Information System (INIS)

2000-11-01

The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. This first workshop was focused on technical issues connected with the first draft of

A GFR benchmark comparison of transient analysis codes based on the ETDR concept

International Nuclear Information System (INIS)

Bubelis, E.; Coddington, P.; Castelliti, D.; Dor, I.; Fouillet, C.; Geus, E. de; Marshall, T.D.; Van Rooijen, W.; Schikorr, M.; Stainsby, R.

2007-01-01

A GFR (Gas-cooled Fast Reactor) transient benchmark study was performed to investigate the ability of different code systems to calculate the transition in the core heat removal from the main circuit forced flow to natural circulation cooling using the Decay Heat Removal (DHR) system. This benchmark is based on a main blower failure in the Experimental Technology Demonstration Reactor (ETDR) with reactor scram. The codes taking part into the benchmark are: RELAP5, TRAC/AAA, CATHARE, SIM-ADS, MANTA and SPECTRA. For comparison purposes the benchmark was divided into several stages: the initial steady-state solution, the main blower flow run-down, the opening of the DHR loop and the transition to natural circulation and finally the 'quasi' steady heat removal from the core by the DHR system. The results submitted by the participants showed that all the codes gave consistent results for all four stages of the benchmark. In the steady-state the calculations revealed some differences in the clad and fuel temperatures, the core and main loop pressure drops and in the total Helium mass inventory. Also some disagreements were observed in the Helium and water flow rates in the DHR loop during the final natural circulation stage. Good agreement was observed for the total main blower flow rate and Helium temperature rise in the core, as well as for the Helium inlet temperature into the core. In order to understand the reason for the differences in the initial 'blind' calculations a second round of calculations was performed using a more precise set of boundary conditions

Storage-Intensive Supercomputing Benchmark Study

Energy Technology Data Exchange (ETDEWEB)

Cohen, J; Dossa, D; Gokhale, M; Hysom, D; May, J; Pearce, R; Yoo, A

2007-10-30

Critical data science applications requiring frequent access to storage perform poorly on today's computing architectures. This project addresses efficient computation of data-intensive problems in national security and basic science by exploring, advancing, and applying a new form of computing called storage-intensive supercomputing (SISC). Our goal is to enable applications that simply cannot run on current systems, and, for a broad range of data-intensive problems, to deliver an order of magnitude improvement in price/performance over today's data-intensive architectures. This technical report documents much of the work done under LDRD 07-ERD-063 Storage Intensive Supercomputing during the period 05/07-09/07. The following chapters describe: (1) a new file I/O monitoring tool iotrace developed to capture the dynamic I/O profiles of Linux processes; (2) an out-of-core graph benchmark for level-set expansion of scale-free graphs; (3) an entity extraction benchmark consisting of a pipeline of eight components; and (4) an image resampling benchmark drawn from the SWarp program in the LSST data processing pipeline. The performance of the graph and entity extraction benchmarks was measured in three different scenarios: data sets residing on the NFS file server and accessed over the network; data sets stored on local disk; and data sets stored on the Fusion I/O parallel NAND Flash array. The image resampling benchmark compared performance of software-only to GPU-accelerated. In addition to the work reported here, an additional text processing application was developed that used an FPGA to accelerate n-gram profiling for language classification. The n-gram application will be presented at SC07 at the High Performance Reconfigurable Computing Technologies and Applications Workshop. The graph and entity extraction benchmarks were run on a Supermicro server housing the NAND Flash 40GB parallel disk array, the Fusion-io. The Fusion system specs are as follows

Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fifth Workshop (BWR-TT5)

International Nuclear Information System (INIS)

2003-01-01

The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fifth workshop was to discuss the results from Phase III (best

Development of core design and analyses technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

1999-03-01

Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

Development of core design and analyses technology for integral reactor

International Nuclear Information System (INIS)

Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

1999-03-01

Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

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)

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...

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)

Experimental benchmark for piping system dynamic-response analyses

International Nuclear Information System (INIS)

1981-01-01

This paper describes the scope and status of a piping system dynamics test program. A 0.20 m(8 in.) nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. Particular attention is given to representing piping restraints. Applied loadings consider component-induced vibration as well as seismic excitation. The principal objective of the program is to provide a benchmark for verification of piping design methods by correlation of predicted and measured responses. Pre-test analysis results and correlation methods are discussed

Experimental benchmark for piping system dynamic response analyses

International Nuclear Information System (INIS)

Schott, G.A.; Mallett, R.H.

1981-01-01

The scope and status of a piping system dynamics test program are described. A 0.20-m nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. Attention is given to representing piping restraints. Applied loadings consider component-induced vibration as well as seismic excitation. The principal objective of the program is to provide a benchmark for verification of piping design methods by correlation of predicted and measured responses. Pre-test analysis results and correlation methods are discussed. 3 refs

Benchmarking in Identifying Priority Directions of Development of Telecommunication Operators

Directory of Open Access Journals (Sweden)

Zaharchenko Lolita A.

2013-12-01

Full Text Available The article analyses evolution of development and possibilities of application of benchmarking in the telecommunication sphere. It studies essence of benchmarking on the basis of generalisation of approaches of different scientists to definition of this notion. In order to improve activity of telecommunication operators, the article identifies the benchmarking technology and main factors, that determine success of the operator in the modern market economy, and the mechanism of benchmarking and component stages of carrying out benchmarking by a telecommunication operator. It analyses the telecommunication market and identifies dynamics of its development and tendencies of change of the composition of telecommunication operators and providers. Having generalised the existing experience of benchmarking application, the article identifies main types of benchmarking of telecommunication operators by the following features: by the level of conduct of (branch, inter-branch and international benchmarking; by relation to participation in the conduct (competitive and joint; and with respect to the enterprise environment (internal and external.

Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 2. TRAB-3D/SMABRE Calculation of the OECD/ NRC PWR MSLB Benchmark

International Nuclear Information System (INIS)

Daavittila, A.; Haemaelaeinen, A.; Kyrki-Rajamaki, R.

2001-01-01

All three exercises of the OECD/NRC Pressurized Water Reactor (PWR) Main-Steam-Line-Break (MSLB) Benchmark were calculated at VTT Energy. The SMABRE thermal-hydraulics code was used for the first exercise, the plant simulation with point-kinetics neutronics. The second exercise was calculated with the TRAB-3D three-dimensional reactor dynamics code. The third exercise was calculated with the combination TRAB-3D/SMABRE. Both codes have been developed at VTT Energy. The results of all the exercises agree reasonably well with those of the other participants; thus, instead of reporting the results, this paper concentrates on describing the computational aspects of the calculation with the foregoing codes and on some observations of the sensitivity of the results. In the TRAB-3D neutron kinetics, the two-group diffusion equations are solved in homogenized fuel assembly geometry with an efficient two-level nodal method. The point of the two-level iteration scheme is that only one unknown variable per node, the average neutron flux, is calculated during the inner iteration. The nodal flux shapes and cross sections are recalculated only once in the outer iteration loop. The TRAB-3D core model includes also parallel one-dimensional channel hydraulics with detailed fuel models. Advanced implicit time discretization methods are used in all submodels. SMABRE is a fast-running five-equation model completed by a drift-flux model, with a time discretization based on a non-iterative semi-implicit algorithm. For the third exercise of the benchmark, the TMI-1 models of TRAB-3D and SMABRE were coupled. This was the first time these codes were coupled together. However, similar coupling of the HEXTRAN and SMABRE codes has been shown to be stable and efficient, when used in safety analyses of Finnish and foreign VVER-type reactors. The coupling used between the two codes is called a parallel coupling. SMABRE solves the thermal hydraulics both in the cooling circuit and in the core

VVER-1000 coolant transient benchmark. Phase 1 (V1000CT-1). Vol. 3: summary results of exercise 2 on coupled 3-D kinetics/core thermal-hydraulics

International Nuclear Information System (INIS)

2007-01-01

In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as current applications. (authors) Recently developed best-estimate computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present volume is a follow-up to the first two volumes. While the first described the specification of the benchmark, the second presented the results of the first exercise that identified the key parameters and important issues concerning the thermal-hydraulic system modelling of the simulated transient caused by the switching on of a main coolant pump when the other three were in operation. Volume 3 summarises the results for Exercise 2 of the benchmark that identifies the key parameters and important issues concerning the 3-D neutron kinetics modelling of the simulated transient. These studies are based on an experiment that was conducted by Bulgarian and Russian engineers during the plant-commissioning phase at the VVER-1000 Kozloduy Unit 6. The final volume will soon be published, completing Phase 1 of this study. (authors)

Stationary PWR-calculations by means of LWRSIM at the NEACRP 3D-LWRCT benchmark

International Nuclear Information System (INIS)

Van de Wetering, T.F.H.

1993-01-01

Within the framework of participation in an international benchmark, calculations were executed by means of an adjusted version of the computer code Light Water Reactor SIMulation (LWRSIM) for three-dimensional reactor core calculations of pressurized water reactors. The 3-D LWR Core Transient Benchmark was set up aimed at the comparison of 3-D computer codes for transient calculations in LWRs. Participation in the benchmark provided more insight in the accuracy of the code when applied for other pressurized water reactors than applied for the nuclear power plant Borssele in the Netherlands, for which the code has been developed and used originally

The fifth Atomic Energy Research dynamic benchmark calculation with HEXTRAN-SMABRE

International Nuclear Information System (INIS)

Haenaelaeinen, Anitta

1998-01-01

The fifth Atomic Energy Research dynamic benchmark is the first Atomic Energy Research benchmark for coupling of the thermohydraulic codes and three-dimensional reactor dynamic core models. In VTT HEXTRAN 2.7 is used for the core dynamics and SMABRE 4.6 as a thermohydraulic model for the primary and secondary loops. The plant model for SMABRE is based mainly on two input models. the Loviisa model and standard WWER-440/213 plant model. The primary circuit includes six separate loops, totally 505 nodes and 652 junctions. The reactor pressure vessel is divided into six parallel channels. In HEXTRAN calculation 176 symmetry is used in the core. In the sequence of main steam header break at the hot standby state, the liquid temperature is decreased symmetrically in the core inlet which leads to return to power. In the benchmark, no isolations of the steam generators are assumed and the maximum core power is about 38 % of the nominal power at four minutes after the break opening in the HEXTRAN-SMABRE calculation. Due to boric acid in the high pressure safety injection water, the power finally starts to decrease. The break flow is pure steam in the HEXTRAN-SMABRE calculation during the whole transient even in the swell levels in the steam generators are very high due to flashing. Because of sudden peaks in the preliminary results of the steam generator heat transfer, the SMABRE drift-flux model was modified. The new model is a simplified version of the EPRI correlation based on test data. The modified correlation behaves smoothly. In the calculations nuclear data is based on the ENDF/B-IV library and it has been evaluated with the CASMO-HEX code. The importance of the nuclear data was illustrated by repeating the benchmark calculation with using three different data sets. Optimal extensive data valid from hot to cold conditions were not available for all types of fuel enrichments needed in this benchmark.(Author)

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

Common Nearest Neighbor Clustering—A Benchmark

Directory of Open Access Journals (Sweden)

Oliver Lemke

2018-02-01

Full Text Available Cluster analyses are often conducted with the goal to characterize an underlying probability density, for which the data-point density serves as an estimate for this probability density. We here test and benchmark the common nearest neighbor (CNN cluster algorithm. This algorithm assigns a spherical neighborhood R to each data point and estimates the data-point density between two data points as the number of data points N in the overlapping region of their neighborhoods (step 1. The main principle in the CNN cluster algorithm is cluster growing. This grows the clusters by sequentially adding data points and thereby effectively positions the border of the clusters along an iso-surface of the underlying probability density. This yields a strict partitioning with outliers, for which the cluster represents peaks in the underlying probability density—termed core sets (step 2. The removal of the outliers on the basis of a threshold criterion is optional (step 3. The benchmark datasets address a series of typical challenges, including datasets with a very high dimensional state space and datasets in which the cluster centroids are aligned along an underlying structure (Birch sets. The performance of the CNN algorithm is evaluated with respect to these challenges. The results indicate that the CNN cluster algorithm can be useful in a wide range of settings. Cluster algorithms are particularly important for the analysis of molecular dynamics (MD simulations. We demonstrate how the CNN cluster results can be used as a discretization of the molecular state space for the construction of a core-set model of the MD improving the accuracy compared to conventional full-partitioning models. The software for the CNN clustering is available on GitHub.

How to Advance TPC Benchmarks with Dependability Aspects

Science.gov (United States)

Almeida, Raquel; Poess, Meikel; Nambiar, Raghunath; Patil, Indira; Vieira, Marco

Transactional systems are the core of the information systems of most organizations. Although there is general acknowledgement that failures in these systems often entail significant impact both on the proceeds and reputation of companies, the benchmarks developed and managed by the Transaction Processing Performance Council (TPC) still maintain their focus on reporting bare performance. Each TPC benchmark has to pass a list of dependability-related tests (to verify ACID properties), but not all benchmarks require measuring their performances. While TPC-E measures the recovery time of some system failures, TPC-H and TPC-C only require functional correctness of such recovery. Consequently, systems used in TPC benchmarks are tuned mostly for performance. In this paper we argue that nowadays systems should be tuned for a more comprehensive suite of dependability tests, and that a dependability metric should be part of TPC benchmark publications. The paper discusses WHY and HOW this can be achieved. Two approaches are introduced and discussed: augmenting each TPC benchmark in a customized way, by extending each specification individually; and pursuing a more unified approach, defining a generic specification that could be adjoined to any TPC benchmark.

Benchmarking infrastructure for mutation text mining.

Science.gov (United States)

Klein, Artjom; Riazanov, Alexandre; Hindle, Matthew M; Baker, Christopher Jo

2014-02-25

Experimental research on the automatic extraction of information about mutations from texts is greatly hindered by the lack of consensus evaluation infrastructure for the testing and benchmarking of mutation text mining systems. We propose a community-oriented annotation and benchmarking infrastructure to support development, testing, benchmarking, and comparison of mutation text mining systems. The design is based on semantic standards, where RDF is used to represent annotations, an OWL ontology provides an extensible schema for the data and SPARQL is used to compute various performance metrics, so that in many cases no programming is needed to analyze results from a text mining system. While large benchmark corpora for biological entity and relation extraction are focused mostly on genes, proteins, diseases, and species, our benchmarking infrastructure fills the gap for mutation information. The core infrastructure comprises (1) an ontology for modelling annotations, (2) SPARQL queries for computing performance metrics, and (3) a sizeable collection of manually curated documents, that can support mutation grounding and mutation impact extraction experiments. We have developed the principal infrastructure for the benchmarking of mutation text mining tasks. The use of RDF and OWL as the representation for corpora ensures extensibility. The infrastructure is suitable for out-of-the-box use in several important scenarios and is ready, in its current state, for initial community adoption.

Benchmarking infrastructure for mutation text mining

Science.gov (United States)

2014-01-01

Background Experimental research on the automatic extraction of information about mutations from texts is greatly hindered by the lack of consensus evaluation infrastructure for the testing and benchmarking of mutation text mining systems. Results We propose a community-oriented annotation and benchmarking infrastructure to support development, testing, benchmarking, and comparison of mutation text mining systems. The design is based on semantic standards, where RDF is used to represent annotations, an OWL ontology provides an extensible schema for the data and SPARQL is used to compute various performance metrics, so that in many cases no programming is needed to analyze results from a text mining system. While large benchmark corpora for biological entity and relation extraction are focused mostly on genes, proteins, diseases, and species, our benchmarking infrastructure fills the gap for mutation information. The core infrastructure comprises (1) an ontology for modelling annotations, (2) SPARQL queries for computing performance metrics, and (3) a sizeable collection of manually curated documents, that can support mutation grounding and mutation impact extraction experiments. Conclusion We have developed the principal infrastructure for the benchmarking of mutation text mining tasks. The use of RDF and OWL as the representation for corpora ensures extensibility. The infrastructure is suitable for out-of-the-box use in several important scenarios and is ready, in its current state, for initial community adoption. PMID:24568600

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)

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)

POLCA-T simulation of OECD/NRC BWR turbine trip benchmark exercise 3 best estimate scenario TT2 test and four extreme scenarios

International Nuclear Information System (INIS)

Panayotov, D.

2004-01-01

Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and the 3D neutron kinetics core model. Code validation plan includes the calculations of Peach Bottom end of cycle 2 turbine trip transients and low-flow stability tests. The paper describes the objectives, method, and results of analyses performed in the final phase of OECD/NRC Peach Bottom 2 Boiling Water Reactor Turbine Trip Benchmark. Brief overview of the code features, the method of simulation, the developed 3D core model and system input deck for Peach Bottom 2 are given. The paper presents the results of benchmark exercise 3 best estimate scenario: coupled 3D core neutron kinetics with system thermal-hydraulics analyses. Performed sensitivity studies cover the SCRAM initiation, carry-under, and decay power. Obtained results including total power, steam dome, core exit, lower and upper plenum, main steam line and turbine inlet pressures showed good agreement with measured plant data Thus the POLCA-T code capabilities for correct simulation of turbine trip transients were proved The performed calculations and obtained results for extreme cases demonstrate the POLCA-T code wide range capabilities to simulate transients when scram, steam bypass, and safety and relief valves are not activated. The code is able to handle such transients even when the reactor power and pressure reach values higher than 600 % of rated power, and 10.8 MPa. (authors)

Benchmarking strategies for measuring the quality of healthcare: problems and prospects.

Science.gov (United States)

Lovaglio, Pietro Giorgio

2012-01-01

Over the last few years, increasing attention has been directed toward the problems inherent to measuring the quality of healthcare and implementing benchmarking strategies. Besides offering accreditation and certification processes, recent approaches measure the performance of healthcare institutions in order to evaluate their effectiveness, defined as the capacity to provide treatment that modifies and improves the patient's state of health. This paper, dealing with hospital effectiveness, focuses on research methods for effectiveness analyses within a strategy comparing different healthcare institutions. The paper, after having introduced readers to the principle debates on benchmarking strategies, which depend on the perspective and type of indicators used, focuses on the methodological problems related to performing consistent benchmarking analyses. Particularly, statistical methods suitable for controlling case-mix, analyzing aggregate data, rare events, and continuous outcomes measured with error are examined. Specific challenges of benchmarking strategies, such as the risk of risk adjustment (case-mix fallacy, underreporting, risk of comparing noncomparable hospitals), selection bias, and possible strategies for the development of consistent benchmarking analyses, are discussed. Finally, to demonstrate the feasibility of the illustrated benchmarking strategies, an application focused on determining regional benchmarks for patient satisfaction (using 2009 Lombardy Region Patient Satisfaction Questionnaire) is proposed.

Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fourth Workshop (BWR-TT4)

International Nuclear Information System (INIS)

2002-01-01

The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fourth workshop was to present and discuss final results of

Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

International Nuclear Information System (INIS)

2013-12-01

Requests for severe accident investigations and assurance of mitigation measures have increased for operating nuclear power plants and the design of advanced nuclear power plants. Severe accident analysis investigations necessitate the analysis of the very complex physical phenomena that occur sequentially during various stages of accident progression. Computer codes are essential tools for understanding how the reactor and its containment might respond under severe accident conditions. The IAEA organizes coordinated research projects (CRPs) to facilitate technology development through international collaboration among Member States. The CRP on Benchmarking Severe Accident Computer Codes for HWR Applications was planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). This publication summarizes the results from the CRP participants. The CRP promoted international collaboration among Member States to improve the phenomenological understanding of severe core damage accidents and the capability to analyse them. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. Seven institutes from five countries with HWRs participated in this CRP

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

The OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark - Steady-state results and status

International Nuclear Information System (INIS)

Reitsma, F.; Han, J.; Ivanov, K.; Sartori, E.

2008-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated. Since only a few pebble-bed HTR experimental facilities or plant data are available the use of code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MW design and a representative set of transient cases is defined as an OECD benchmark. The scope of the benchmark is to establish a series of well-defined multi-dimensional computational benchmark problems with a common given set of cross-sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. The OECD benchmark includes steady-state and transients cases. Although the focus of the benchmark is on the modelling of the transient behaviour of the PBMR core, it was also necessary to define some steady-state cases to ensure consistency between the different approaches before results of transient cases could be compared. This paper describes the status of the benchmark project and shows the results for the three steady state exercises defined as a standalone neutronics calculation, a standalone thermal-hydraulic core calculation, and a coupled neutronics/thermal-hydraulic simulation. (authors)

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

Benchmarking Strategies for Measuring the Quality of Healthcare: Problems and Prospects

Science.gov (United States)

Lovaglio, Pietro Giorgio

2012-01-01

Over the last few years, increasing attention has been directed toward the problems inherent to measuring the quality of healthcare and implementing benchmarking strategies. Besides offering accreditation and certification processes, recent approaches measure the performance of healthcare institutions in order to evaluate their effectiveness, defined as the capacity to provide treatment that modifies and improves the patient's state of health. This paper, dealing with hospital effectiveness, focuses on research methods for effectiveness analyses within a strategy comparing different healthcare institutions. The paper, after having introduced readers to the principle debates on benchmarking strategies, which depend on the perspective and type of indicators used, focuses on the methodological problems related to performing consistent benchmarking analyses. Particularly, statistical methods suitable for controlling case-mix, analyzing aggregate data, rare events, and continuous outcomes measured with error are examined. Specific challenges of benchmarking strategies, such as the risk of risk adjustment (case-mix fallacy, underreporting, risk of comparing noncomparable hospitals), selection bias, and possible strategies for the development of consistent benchmarking analyses, are discussed. Finally, to demonstrate the feasibility of the illustrated benchmarking strategies, an application focused on determining regional benchmarks for patient satisfaction (using 2009 Lombardy Region Patient Satisfaction Questionnaire) is proposed. PMID:22666140

NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark. Volume II: uncertainty and sensitivity analyses of void distribution and critical power - Specification

International Nuclear Information System (INIS)

Aydogan, F.; Hochreiter, L.; Ivanov, K.; Martin, M.; Utsuno, H.; Sartori, E.

2010-01-01

experimental cases from the BFBT database for both steady-state void distribution and steady-state critical power uncertainty analyses. In order to study the basic thermal-hydraulics in a single channel, where the concern regarding the cross-flow effect modelling could be removed, an elemental task is proposed, consisting of two sub-tasks that are placed in each phase of the benchmark scope as follows: - Sub-task 1: Void fraction in elemental channel benchmark; - Sub-task 2: Critical power in elemental channel benchmark. The first task can also be utilised as an uncertainty analysis exercise for fine computational fluid dynamics (CFD) models for which the full bundle sensitivity or uncertainty analysis is more difficult. The task is added to the second volume of the specification as an optional exercise. Chapter 2 of this document provides the definition of UA/SA terms. Chapter 3 provides the selection and characterisation of the input uncertain parameters for the BFBT benchmark and the description of the elemental task. Chapter 4 describes the suggested approach for UA/SA of the BFBT benchmark. Chapter 5 provides the selection of data sets for the uncertainty analysis and the elemental task from the BFBT database. Chapter 6 specifies the requested output for void distribution and critical power uncertainty analyses (Exercises I-4 and II-3) as well as for the elemental task. Chapter 7 provides conclusions. Appendix 1 discusses the UA/SA methods. Appendix 2 presents the Phenomena Identification Ranking Tables (PIRT) developed at PSU for void distribution and critical power predictions in order to assist participants in selecting the most sensitive/uncertain code model parameters

VIPRE modeling of VVER-1000 reactor core for DNB analyses

Energy Technology Data Exchange (ETDEWEB)

Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

1995-09-01

Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

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

Boiling water reactor turbine trip (TT) benchmark. Volume II: Summary Results of Exercise 1

International Nuclear Information System (INIS)

Akdeniz, Bedirhan; Ivanov, Kostadin N.; Olson, Andy M.

2005-06-01

The OECD Nuclear Energy Agency (NEA) completed under US Nuclear Regulatory Commission (NRC) sponsorship a PWR main steam line break (MSLB) benchmark against coupled system three-dimensional (3-D) neutron kinetics and thermal-hydraulic codes. Another OECD/NRC coupled-code benchmark was recently completed for a BWR turbine trip (TT) transient and is the object of the present report. Turbine trip transients in a BWR are pressurisation events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. The data made available from actual experiments carried out at the Peach Bottom 2 plant make the present benchmark particularly valuable. While defining and coordinating the BWR TT benchmark, a systematic approach and level methodology not only allowed for a consistent and comprehensive validation process, but also contributed to the study of key parameters of pressurisation transients. The benchmark consists of three separate exercises, two initial states and five transient scenarios. The BWR TT Benchmark will be published in four volumes as NEA reports. CD-ROMs will also be prepared and will include the four reports and the transient boundary conditions, decay heat values as a function of time, cross-section libraries and supplementary tables and graphs not published in the paper version. BWR TT Benchmark - Volume I: Final Specifications was issued in 2001 [NEA/NSC/DOC(2001)]. The benchmark team [Pennsylvania State University (PSU) in co-operation with Exelon Nuclear and the NEA] has been responsible for coordinating benchmark activities, answering participant questions and assisting them in developing their models, as well as analysing submitted solutions and providing reports summarising the results for each phase. The benchmark team has also been involved in the technical aspects of the benchmark, including sensitivity studies for the different exercises. Volume II summarises the results for Exercise 1 of the

Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

International Nuclear Information System (INIS)

Bess, John D.; Fujimoto, Nozomu

2014-01-01

Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with 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

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)

CONSUL code package application for LMFR core calculations

Energy Technology Data Exchange (ETDEWEB)

Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)

2008-07-01

CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)

Pericles and Attila results for the C5G7 MOX benchmark problems

International Nuclear Information System (INIS)

Wareing, T.A.; McGhee, J.M.

2002-01-01

Recently the Nuclear Energy Agency has published a new benchmark entitled, 'C5G7 MOX Benchmark.' This benchmark is to test the ability of current transport codes to treat reactor core problems without spatial homogenization. The benchmark includes both a two- and three-dimensional problem. We have calculated results for these benchmark problems with our Pericles and Attila codes. Pericles is a one-,two-, and three-dimensional unstructured grid discrete-ordinates code and was used for the twodimensional benchmark problem. Attila is a three-dimensional unstructured tetrahedral mesh discrete-ordinate code and was used for the three-dimensional problem. Both codes use discontinuous finite element spatial differencing. Both codes use diffusion synthetic acceleration (DSA) for accelerating the inner iterations.

OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization

Energy Technology Data Exchange (ETDEWEB)

Hou, Jason, E-mail: jason.hou@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Ivanov, Kostadin N. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Boyarinov, Victor F.; Fomichenko, Peter A. [National Research Centre “Kurchatov Institute”, Kurchatov Sq. 1, Moscow (Russian Federation)

2017-06-15

Highlights: • A time-dependent homogenization-free neutron transport benchmark was created. • The first phase, known as the kinetics phase, was described in this work. • Preliminary results for selected 2-D transient exercises were presented. - Abstract: A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.

International benchmark tests of the FENDL-1 Nuclear Data Library

International Nuclear Information System (INIS)

Fischer, U.

1997-01-01

An international benchmark validation task has been conducted to validate the fusion evaluated nuclear data library FENDL-1 through data tests against integral 14 MeV neutron experiments. The main objective of this task was to qualify the FENDL-1 working libraries for fusion applications and to elaborate recommendations for further data improvements. Several laboratories and institutions from the European Union, Japan, the Russian Federation and US have contributed to the benchmark task. A large variety of existing integral 14 MeV benchmark experiments was analysed with the FENDL-1 working libraries for continuous energy Monte Carlo and multigroup discrete ordinate calculations. Results of the benchmark analyses have been collected, discussed and evaluated. The major findings, conclusions and recommendations are presented in this paper. With regard to the data quality, it is summarised that fusion nuclear data have reached a high confidence level with the available FENDL-1 data library. With few exceptions this holds for the materials of highest importance for fusion reactor applications. As a result of the performed benchmark analyses, some existing deficiencies and discrepancies have been identified that are recommended for removal in theforthcoming FENDL-2 data file. (orig.)

Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

2011-01-01

Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

Nuclear data uncertainties for local power densities in the Martin-Hoogenboom benchmark

International Nuclear Information System (INIS)

Van der Marck, S.C.; Rochman, D.A.

2013-01-01

The recently developed method of fast Total Monte Carlo to propagate nuclear data uncertainties was applied to the Martin-Hoogenboom benchmark. This Martin- Hoogenboom benchmark prescribes that one calculates local pin powers (of light water cooled reactor) with a statistical uncertainty lower than 1% everywhere. Here we report, for the first time, an estimate of the nuclear data uncertainties for these local pin powers. For each of the more than 6 million local power tallies, the uncertainty due to nuclear data uncertainties was calculated, based on random variation of data for 235 U, 238 U, 239 Pu and H in H 2 O thermal scattering. In the center of the core region, the nuclear data uncertainty is 0.9%. Towards the edges of the core, this uncertainty increases to roughly 3%. The nuclear data uncertainties have been shown to be larger than the statistical uncertainties that the benchmark prescribes

Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

Energy Technology Data Exchange (ETDEWEB)

Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-11-15

This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

TRX and UO2 criticality benchmarks with SAM-CE

International Nuclear Information System (INIS)

Beer, M.; Troubetzkoy, E.S.; Lichtenstein, H.; Rose, P.F.

1980-01-01

A set of thermal reactor benchmark calculations with SAM-CE which have been conducted at both MAGI and at BNL are described. Their purpose was both validation of the SAM-CE reactor eigenvalue capability developed by MAGI and a substantial contribution to the data testing of both ENDF/B-IV and ENDF/B-V libraries. This experience also resulted in increased calculational efficiency of the code and an example is given. The benchmark analysis included the TRX-1 infinite cell using both ENDF/B-IV and ENDF/B-V cross section sets and calculations using ENDF/B-IV of the TRX-1 full core and TRX-2 cell. BAPL-UO2-1 calculations were conducted for the cell using both ENDF/B-IV and ENDF/B-V and for the full core with ENDF/B-V

Pool critical assembly pressure vessel facility benchmark

International Nuclear Information System (INIS)

Remec, I.; Kam, F.B.K.

1997-07-01

This pool critical assembly (PCA) pressure vessel wall facility benchmark (PCA benchmark) is described and analyzed in this report. Analysis of the PCA benchmark can be used for partial fulfillment of the requirements for the qualification of the methodology for pressure vessel neutron fluence calculations, as required by the US Nuclear Regulatory Commission regulatory guide DG-1053. Section 1 of this report describes the PCA benchmark and provides all data necessary for the benchmark analysis. The measured quantities, to be compared with the calculated values, are the equivalent fission fluxes. In Section 2 the analysis of the PCA benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed for three ENDF/B-VI-based multigroup libraries: BUGLE-93, SAILOR-95, and BUGLE-96. An excellent agreement of the calculated (C) and measures (M) equivalent fission fluxes was obtained. The arithmetic average C/M for all the dosimeters (total of 31) was 0.93 ± 0.03 and 0.92 ± 0.03 for the SAILOR-95 and BUGLE-96 libraries, respectively. The average C/M ratio, obtained with the BUGLE-93 library, for the 28 measurements was 0.93 ± 0.03 (the neptunium measurements in the water and air regions were overpredicted and excluded from the average). No systematic decrease in the C/M ratios with increasing distance from the core was observed for any of the libraries used

Development of core design and analyses technology for integral reactor

International Nuclear Information System (INIS)

Zee, Sung Quun; Lee, C. C.; Kim, K. Y.

2002-03-01

In general, small and medium-sized integral reactors adopt new technology such as passive and inherent safety concepts to minimize the necessity of power source and operator actions, and to provide the automatic measures to cope with any accidents. Specifically, such reactors are often designed with a lower core power density and with soluble boron free concept for system simplification. Those reactors require ultra long cycle operation for higher economical efficiency. This cycle length requirement is one of the important factors in the design of burnable absorbers as well as assurance of shutdown margin. Hence, both computer code system and design methodology based on the today's design technology for the current commercial reactor cores require intensive improvement for the small and medium-sized soluble boron free reactors. New database is also required for the development of this type of reactor core. Under these technical requirements, conceptual design of small integral reactor SMART has been performed since July 1997, and recently completed under the long term nuclear R and D program. Thus, the final objectives of this work is design and development of an integral reactor core and development of necessary indigenous design technology. To reach the goal of the 2nd stage R and D program for basic design of SMART, design bases and requirements adequate for ultra long cycle and soluble boron free concept are established. These bases and requirements are satisfied by the core loading pattern. Based on the core loading pattern, nuclear, and thermal and hydraulic characteristics are analyzed. Also included are fuel performance analysis and development of a core protection and monitoring system that is adequate for the soluble boron free core of an integral reactor. Core shielding design analysis is accomplished, too. Moreover, full scope interface data are produced for reactor safety and performance analyses and other design activities. Nuclear, thermal and

Activities of the AZTLAN team on the OECD/Nea benchmark on fast reactors

Energy Technology Data Exchange (ETDEWEB)

Lopez S, R.; Gomez T, A.; Puente E, F. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E.; Arriaga R, L., E-mail: armando.gomez@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, 07738 Ciudad de Mexico (Mexico)

2017-09-15

In the present paper, the activities of the AZTLAN Platform Fast Reactor Group on the OECD/Nea Benchmark will be described. The main objective of these activities is to test the group staff and capabilities as well as the domestic code reliability by putting them into test in this exercise with different institutions from around the world. Six different core configurations were treated; these are described in two different versions of the Benchmark document. The main tools used by the group were the Finnish stochastic Monte Carlo code Serpent for full core calculations and macroscopic Cross Sections (X S) generation, and the domestic deterministic code AZNHEX for full core calculations. Different calculations were performed, such as full core calculations under nominal conditions, with control rods fully and partially inserted and with the sodium voided in the active zone as well as different reactivity shift values due to various conditions of radial and axial expansion of the fuel elements and structural material. The results obtained in the full core calculations and most of the reactivity shift calculations obtained by our group were indeed comparable to the ones obtained by different institutions when using similar methodologies. Given these favorable results it can be said that the main objective was met and the group showed their capabilities, as well as its possibility to collaborate with other institutes, placing Mexico in a good position in fast reactor analysis. Future work will continue with the calculations not yet treated and with the new core specifications on the new versions of the Benchmark document. (Author)

DORT-TD/THERMIX solutions for the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

International Nuclear Information System (INIS)

Tyobeka, Bismark; Pautz, Andreas; Ivanov, Kostadin

2008-01-01

In new reactor designs that are still under review such as the PBMR, not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400 MW OECD/NEA/NSC coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate the transient scenarios in the above-mentioned benchmark problem. Steady-state calculations results are compared with selected participants' results as well as transient models in which the diffusion and transport theory solutions of the same code system are directly compared. Several sensitivity studies are also shown in order to determine how much the change in certain parameters influences the overall behaviour of a given transient. It is shown in this paper that DORT-TD/THERMIX is a versatile tool which can be deployed for design and safety analyses of high temperature reactors of pebble-bed type. (authors)

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)

Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

International Nuclear Information System (INIS)

Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina

2017-01-01

The 7 th dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7 th AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7 th AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

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)

Development of the computer code system for the analyses of PWR core

International Nuclear Information System (INIS)

Tsujimoto, Iwao; Naito, Yoshitaka.

1992-11-01

This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

JENDL-4.0 benchmarking for fission reactor applications

International Nuclear Information System (INIS)

Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

2011-01-01

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0. (author)

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

Analyses on the BFS critical experiments. An analysis on the BFS-62-1 and 62-2 cores

International Nuclear Information System (INIS)

Sugino, Kazuteru; Shono, Akira

2002-04-01

In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 and BFS-62-2 cores. The BFS-62-1 core models the present BN-600, and contains the enriched UO 2 fuel surrounded by the UO 2 blanket. The BFS-62-2 core has the same layout as the BFS-62-1 but the blanket region was replaced with stainless steel shied. For core parameter analyses, the 3-D Hexagonal-Z or XYZ geometry model was applied by not only diffusion calculation but also transport calculation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality, the reaction rate ratio and reaction rate distribution in BFS-62-1. In the reaction rate distribution of BFS-62-2 calculation without cross-section adjustment produced big radial dependency of calculation over experiment value (C/E value) in the core region and overestimation in the shield region. Cross-section adjustment technique procedure improved those estimation, however alternation of cross-section of Iron, which was dominant in above improvement, compared to the cross-section error, and further investigation was required. Concerning the control rod worth of BFS-62-1, radial dependency of the C/E value was observed whether cross-section adjustment technique was applied or not, therefore comparison with results of other BFS-62 cores analyses is

Benchmarking high performance computing architectures with CMS’ skeleton framework

Science.gov (United States)

Sexton-Kennedy, E.; Gartung, P.; Jones, C. D.

2017-10-01

In 2012 CMS evaluated which underlying concurrency technology would be the best to use for its multi-threaded framework. The available technologies were evaluated on the high throughput computing systems dominating the resources in use at that time. A skeleton framework benchmarking suite that emulates the tasks performed within a CMSSW application was used to select Intel’s Thread Building Block library, based on the measured overheads in both memory and CPU on the different technologies benchmarked. In 2016 CMS will get access to high performance computing resources that use new many core architectures; machines such as Cori Phase 1&2, Theta, Mira. Because of this we have revived the 2012 benchmark to test it’s performance and conclusions on these new architectures. This talk will discuss the results of this exercise.

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

Energy Technology Data Exchange (ETDEWEB)

Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

2005-07-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

International Nuclear Information System (INIS)

Elina Syrjaelahti; Anitta Haemaelaeinen

2005-01-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Start-up of a cold loop in a VVER-440, the 7{sup th} AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

Energy Technology Data Exchange (ETDEWEB)

Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

2017-09-15

The 7{sup th} dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7{sup th} AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7{sup th} AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

Benchmark analyses for EFF-1, -3 and FENDL-1, -2 beryllium data

International Nuclear Information System (INIS)

Fischer, U.; Wu, Y.

1999-01-01

The present article is part of the summary report on the Consultants' Meeting on the transport sublibrary of the Fusion Evaluated Data Library version 2.0. It reports on the comparison between beryllium benchmark experiments and Monte Carlo calculations, using different versions of the FENDL and EFF libraries

Solution of the fifth dynamic Atomic Energy Research benchmark problem using the coupled code DIN3/ATHLET

International Nuclear Information System (INIS)

Kliem, S.

1998-01-01

The fifth dynamic benchmark is the first benchmark for coupled thermohydraulic system/three dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and the shutdown conditions with one control rod group s tucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Several aspects of the very complex and complicated benchmark problem are analyzed in detail. Sensitivity studies with different hydraulic parameters are made. The influence on the course of the transient and on the solution is discussed.(Author)

Present status and benchmark tests of JENDL-2

International Nuclear Information System (INIS)

Kikuchi, Y.

1983-01-01

The second version of Japanese Evaluated Nuclear Data Library (JENDL-2) consists of the evaluated data from 10 -5 eV to 20 MeV for 176 nuclides including 99 fission product nuclei. Complete reevaluation has been made to heavy actinide, fission product and main structural material nuclides. Benchmark tests have been made on JENDL-2 for fast reactor application. Various characteristics in core center have been tested with one-dimensional model for total of 27 assemblies, and more sophisticated problems have been examined for MOZART and ZPPR-3. Furthermore analyses of JUPITER project give useful information. Satisfactory results have been obtained as a whole. However, the spectrum is a little underestimated above a few hundred keV and below a few keV. The positive sodium void reactivity worth is much overestimated. As to the latter, the sensitivity analysis with the generalized perturbation method suggests that the fission cross section of 239 Pu below a few keV has an important role. (Auth.)

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.

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

Verification of HELIOS-MASTER system through benchmark of Halden boiling water reactor (HBWR)

International Nuclear Information System (INIS)

Kim, Ha Yong; Song, Jae Seung; Cho, Jin Young; Kim, Kang Seok; Lee, Chung Chan; Zee, Sung Quun

2004-01-01

To verify the HELIOS-MASTER computer code system for a nuclear design, we have been performed benchmark calculations for various reactor cores. The Halden reactor is a boiling, heavy water moderated reactor. At a full power of 18-20MWt, the moderator temperature is 240 .deg. C and the pressure is 33 bar. This study describes the verification of the HELIOS-MASTER computer code system for a nuclear design and the analysis of a hexagonal and D 2 O moderated core through a benchmark of the Halden reactor core. 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 the neutronics model with the TPEN (Triangle based Polynomial Expansion Nodal) method for a hexagonal geometry

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.

Benchmarking Linked Open Data Management Systems

NARCIS (Netherlands)

R. Angles Rojas (Renzo); M.-D. Pham (Minh-Duc); P.A. Boncz (Peter)

2014-01-01

htmlabstractWith inherent support for storing and analysing highly interconnected data, graph and RDF databases appear as natural solutions for developing Linked Open Data applications. However, current benchmarks for these database technologies do not fully attain the desirable characteristics

JENDL-4.0 benchmarking for effective delayed neutron fraction with a continuous-energy Monte Carlo code MVP

International Nuclear Information System (INIS)

Nagaya, Yasunobu

2013-01-01

Benchmark calculations with a continuous-energy Monte Carlo code have been performed for delayed neutron data of JENDL-4.0. JENDL-4.0 gives good prediction for the effective delayed neutron fraction in the present benchmarks but further detailed analysis is required for some cores. (author)

Neutron radiography (NRAD) reactor 64-element core upgrade

Energy Technology Data Exchange (ETDEWEB)

Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-03-01

The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately ±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

A framework for benchmarking land models

Directory of Open Access Journals (Sweden)

Y. Q. Luo

2012-10-01

Full Text Available Land models, which have been developed by the modeling community in the past few decades to predict future states of ecosystems and climate, have to be critically evaluated for their performance skills of simulating ecosystem responses and feedback to climate change. Benchmarking is an emerging procedure to measure performance of models against a set of defined standards. This paper proposes a benchmarking framework for evaluation of land model performances and, meanwhile, highlights major challenges at this infant stage of benchmark analysis. The framework includes (1 targeted aspects of model performance to be evaluated, (2 a set of benchmarks as defined references to test model performance, (3 metrics to measure and compare performance skills among models so as to identify model strengths and deficiencies, and (4 model improvement. Land models are required to simulate exchange of water, energy, carbon and sometimes other trace gases between the atmosphere and land surface, and should be evaluated for their simulations of biophysical processes, biogeochemical cycles, and vegetation dynamics in response to climate change across broad temporal and spatial scales. Thus, one major challenge is to select and define a limited number of benchmarks to effectively evaluate land model performance. The second challenge is to develop metrics of measuring mismatches between models and benchmarks. The metrics may include (1 a priori thresholds of acceptable model performance and (2 a scoring system to combine data–model mismatches for various processes at different temporal and spatial scales. The benchmark analyses should identify clues of weak model performance to guide future development, thus enabling improved predictions of future states of ecosystems and climate. The near-future research effort should be on development of a set of widely acceptable benchmarks that can be used to objectively, effectively, and reliably evaluate fundamental properties

Analyses of the stability and core taxonomic memberships of the human microbiome.

Directory of Open Access Journals (Sweden)

Kelvin Li

Full Text Available Analyses of the taxonomic diversity associated with the human microbiome continue to be an area of great importance. The study of the nature and extent of the commonly shared taxa ("core", versus those less prevalent, establishes a baseline for comparing healthy and diseased groups by quantifying the variation among people, across body habitats and over time. The National Institutes of Health (NIH sponsored Human Microbiome Project (HMP has provided an unprecedented opportunity to examine and better define what constitutes the taxonomic core within and across body habitats and individuals through pyrosequencing-based profiling of 16S rRNA gene sequences from oral, skin, distal gut (stool, and vaginal body habitats from over 200 healthy individuals. A two-parameter model is introduced to quantitatively identify the core taxonomic members of each body habitat's microbiota across the healthy cohort. Using only cutoffs for taxonomic ubiquity and abundance, core taxonomic members were identified for each of the 18 body habitats and also for the 4 higher-level body regions. Although many microbes were shared at low abundance, they exhibited a relatively continuous spread in both their abundance and ubiquity, as opposed to a more discretized separation. The numbers of core taxa members in the body regions are comparatively small and stable, reflecting the relatively high, but conserved, interpersonal variability within the cohort. Core sizes increased across the body regions in the order of: vagina, skin, stool, and oral cavity. A number of "minor" oral taxonomic core were also identified by their majority presence across the cohort, but with relatively low and stable abundances. A method for quantifying the difference between two cohorts was introduced and applied to samples collected on a second visit, revealing that over time, the oral, skin, and stool body regions tended to be more transient in their taxonomic structure than the vaginal body region.

The development of code benchmarks

International Nuclear Information System (INIS)

Glass, R.E.

1986-01-01

Sandia National Laboratories has undertaken a code benchmarking effort to define a series of cask-like problems having both numerical solutions and experimental data. The development of the benchmarks includes: (1) model problem definition, (2) code intercomparison, and (3) experimental verification. The first two steps are complete and a series of experiments are planned. The experiments will examine the elastic/plastic behavior of cylinders for both the end and side impacts resulting from a nine meter drop. The cylinders will be made from stainless steel and aluminum to give a range of plastic deformations. This paper presents the results of analyses simulating the model's behavior using materials properties for stainless steel and aluminum

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)

Oxygen and carbon isotope analyses of a Late Quaternary core in the Zaire (Congo) fan

International Nuclear Information System (INIS)

Olausson, E.

1984-01-01

Oxygen and carbon isotope analyses have been carried out on samples from a core of the Angola Basin (6 0 50'S, 10 0 45'E, depth 2100 m). The pelagic foraminifer Globigerinoides ruber, a species with a shallow water habitat, and two benthic species Uvigerina peregrina and Bulimina aculeata have been analysed. The data are given relative to PDB. (Auth.)

WIPP Benchmark calculations with the large strain SPECTROM codes

International Nuclear Information System (INIS)

Callahan, G.D.; DeVries, K.L.

1995-08-01

This report provides calculational results from the updated Lagrangian structural finite-element programs SPECTROM-32 and SPECTROM-333 for the purpose of qualifying these codes to perform analyses of structural situations in the Waste Isolation Pilot Plant (WIPP). Results are presented for the Second WIPP Benchmark (Benchmark II) Problems and for a simplified heated room problem used in a parallel design calculation study. The Benchmark II problems consist of an isothermal room problem and a heated room problem. The stratigraphy involves 27 distinct geologic layers including ten clay seams of which four are modeled as frictionless sliding interfaces. The analyses of the Benchmark II problems consider a 10-year simulation period. The evaluation of nine structural codes used in the Benchmark II problems shows that inclusion of finite-strain effects is not as significant as observed for the simplified heated room problem, and a variety of finite-strain and small-strain formulations produced similar results. The simplified heated room problem provides stratigraphic complexity equivalent to the Benchmark II problems but neglects sliding along the clay seams. The simplified heated problem does, however, provide a calculational check case where the small strain-formulation produced room closures about 20 percent greater than those obtained using finite-strain formulations. A discussion is given of each of the solved problems, and the computational results are compared with available published results. In general, the results of the two SPECTROM large strain codes compare favorably with results from other codes used to solve the problems

The OECD/NEA/NSC PBMR 400 MW coupled neutronics thermal hydraulics transient benchmark: transient results - 290

International Nuclear Information System (INIS)

Strydom, G.; Reitsma, F.; Ngeleka, P.T.; Ivanov, K.N.

2010-01-01

The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of well defined multi-dimensional computational benchmark problems with a common given set of cross sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants' results are compared using a statistical method and possible areas of future code improvement are identified. (authors)

Transient analyses for a molten salt fast reactor with optimized core geometry

Energy Technology Data Exchange (ETDEWEB)

Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

2015-10-15

Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

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.

Hospital benchmarking: are U.S. eye hospitals ready?

Science.gov (United States)

de Korne, Dirk F; van Wijngaarden, Jeroen D H; Sol, Kees J C A; Betz, Robert; Thomas, Richard C; Schein, Oliver D; Klazinga, Niek S

2012-01-01

Benchmarking is increasingly considered a useful management instrument to improve quality in health care, but little is known about its applicability in hospital settings. The aims of this study were to assess the applicability of a benchmarking project in U.S. eye hospitals and compare the results with an international initiative. We evaluated multiple cases by applying an evaluation frame abstracted from the literature to five U.S. eye hospitals that used a set of 10 indicators for efficiency benchmarking. Qualitative analysis entailed 46 semistructured face-to-face interviews with stakeholders, document analyses, and questionnaires. The case studies only partially met the conditions of the evaluation frame. Although learning and quality improvement were stated as overall purposes, the benchmarking initiative was at first focused on efficiency only. No ophthalmic outcomes were included, and clinicians were skeptical about their reporting relevance and disclosure. However, in contrast with earlier findings in international eye hospitals, all U.S. hospitals worked with internal indicators that were integrated in their performance management systems and supported benchmarking. Benchmarking can support performance management in individual hospitals. Having a certain number of comparable institutes provide similar services in a noncompetitive milieu seems to lay fertile ground for benchmarking. International benchmarking is useful only when these conditions are not met nationally. Although the literature focuses on static conditions for effective benchmarking, our case studies show that it is a highly iterative and learning process. The journey of benchmarking seems to be more important than the destination. Improving patient value (health outcomes per unit of cost) requires, however, an integrative perspective where clinicians and administrators closely cooperate on both quality and efficiency issues. If these worlds do not share such a relationship, the added

Advanced core-analyses for subsurface characterization

Science.gov (United States)

Pini, R.

2017-12-01

The heterogeneity of geological formations varies over a wide range of length scales and represents a major challenge for predicting the movement of fluids in the subsurface. Although they are inherently limited in the accessible length-scale, laboratory measurements on reservoir core samples still represent the only way to make direct observations on key transport properties. Yet, properties derived on these samples are of limited use and should be regarded as sample-specific (or `pseudos'), if the presence of sub-core scale heterogeneities is not accounted for in data processing and interpretation. The advent of imaging technology has significantly reshaped the landscape of so-called Special Core Analysis (SCAL) by providing unprecedented insight on rock structure and processes down to the scale of a single pore throat (i.e. the scale at which all reservoir processes operate). Accordingly, improved laboratory workflows are needed that make use of such wealth of information by e.g., referring to the internal structure of the sample and in-situ observations, to obtain accurate parameterisation of both rock- and flow-properties that can be used to populate numerical models. We report here on the development of such workflow for the study of solute mixing and dispersion during single- and multi-phase flows in heterogeneous porous systems through a unique combination of two complementary imaging techniques, namely X-ray Computed Tomography (CT) and Positron Emission Tomography (PET). The experimental protocol is applied to both synthetic and natural porous media, and it integrates (i) macroscopic observations (tracer effluent curves), (ii) sub-core scale parameterisation of rock heterogeneities (e.g., porosity, permeability and capillary pressure), and direct 3D observation of (iii) fluid saturation distribution and (iv) the dynamic spreading of the solute plumes. Suitable mathematical models are applied to reproduce experimental observations, including both 1D and 3D

Analyses and results of the OECD/NEA WPNCS EGUNF benchmark phase II. Technical report; Analysen und Ergebnisse zum OECD/NEA WPNCS EGUNF Benchmark Phase II. Technischer Bericht

Energy Technology Data Exchange (ETDEWEB)

Hannstein, Volker; Sommer, Fabian

2017-05-15

The report summarizes the performed studies and results in the frame of the phase II benchmarks of the expert group of used nuclear fuel (EGUNF) of the working party of nuclear criticality safety (WPNCS) of the nuclear energy agency (NEA) of the organization for economic co-operation and development (OECD). The studies specified within the benchmarks have been realized to the full extent. The scope of the benchmarks was the comparison of a generic BWR fuel element with gadolinium containing fuel rods with several computer codes and cross section libraries of different international working groups and institutions. The used computational model allows the evaluation of the accuracy of fuel rod and their influence of the inventory calculations and the respective influence on BWR burnout credit calculations.

Benchmarking local healthcare-associated infections: Available benchmarks and interpretation challenges

Directory of Open Access Journals (Sweden)

Aiman El-Saed

2013-10-01

Full Text Available Summary: Growing numbers of healthcare facilities are routinely collecting standardized data on healthcare-associated infection (HAI, which can be used not only to track internal performance but also to compare local data to national and international benchmarks. Benchmarking overall (crude HAI surveillance metrics without accounting or adjusting for potential confounders can result in misleading conclusions. Methods commonly used to provide risk-adjusted metrics include multivariate logistic regression analysis, stratification, indirect standardization, and restrictions. The characteristics of recognized benchmarks worldwide, including the advantages and limitations are described. The choice of the right benchmark for the data from the Gulf Cooperation Council (GCC states is challenging. The chosen benchmark should have similar data collection and presentation methods. Additionally, differences in surveillance environments including regulations should be taken into consideration when considering such a benchmark. The GCC center for infection control took some steps to unify HAI surveillance systems in the region. GCC hospitals still need to overcome legislative and logistic difficulties in sharing data to create their own benchmark. The availability of a regional GCC benchmark may better enable health care workers and researchers to obtain more accurate and realistic comparisons. Keywords: Benchmarking, Comparison, Surveillance, Healthcare-associated infections

Benchmark experiments of effective delayed neutron fraction βeff at FCA

International Nuclear Information System (INIS)

Sakurai, Takeshi; Okajima, Shigeaki

1999-01-01

Benchmark experiments of effective delayed neutron fraction β eff were performed at Fast Critical Assembly (FCA) in the Japan Atomic Energy Research Institute. The experiments were made in three cores providing systematic change of nuclide contribution to the β eff : XIX-1 core fueled with 93% enriched uranium, XIX-2 core fueled with plutonium and uranium (23% enrichment) and XIX-3 core fueled with plutonium (92% fissile Pu). Six organizations from five countries participated in these experiments and measured the β eff by using their own methods and instruments. Target accuracy in the β eff was achieved to be better than ±3% by averaging the β eff values measured using a wide variety of experimental methods. (author)

Benchmark Specification for an HTR Fuelled with Reactor-grade Plutonium (or Reactor-grade Pu/Th and U/Th). Proposal version 2

International Nuclear Information System (INIS)

Hosking, J.G.; Newton, T.D.; Morris, P.

2007-01-01

This benchmark proposal builds upon that specified in NEA/NSC/DOC(2003)22 report. In addition to the three phases described in that report, another two phases have now been defined. Additional items for calculation have also been added to the existing phases. It is intended that further items may be added to the benchmark after consultation with its participants. Although the benchmark is specifically designed to provide inter-comparisons for plutonium- and thorium-containing fuels, it is proposed that phases considering simple calculations for a uranium fuel cell and uranium core be included. The purpose of these is to identify any increased uncertainties, relative to uranium fuel, associated with the lesser-known fuels to be investigated in different phases of this benchmark. The first phase considers an infinite array of fuel pebbles fuelled with uranium fuel. Phase 2 considers a similar array of pebbles but for plutonium fuel. Phase 3 continues the plutonium fuel inter-comparisons within the context of whole core calculations. Calculations for Phase 4 are for a uranium-fuelled core. Phase 5 considers an infinite array of pebbles containing thorium. In setting the benchmark the requirements in the definition of the LEUPRO-12 PROTEUS benchmark have been considered. Participants were invited to submit both deterministic results as well as, where appropriate, results from Monte Carlo calculations. Fundamental nuclear data, Avogadro's number, natural abundance data and atomic weights have been taken from the references indicated in the document

Defining a methodology for benchmarking spectrum unfolding codes

International Nuclear Information System (INIS)

Meyer, W.; Kirmser, P.G.; Miller, W.H.; Hu, K.K.

1976-01-01

It has long been recognized that different neutron spectrum unfolding codes will produce significantly different results when unfolding the same measured data. In reviewing the results of such analyses it has been difficult to determine which result if any is the best representation of what was measured by the spectrometer detector. A proposal to develop a benchmarking procedure for spectrum unfolding codes is presented. The objective of the procedure will be to begin to develop a methodology and a set of data with a well established and documented result that could be used to benchmark and standardize the various unfolding methods and codes. It is further recognized that development of such a benchmark must involve a consensus of the technical community interested in neutron spectrum unfolding

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.

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)

Operating experience, measurements, and analysis of the LEU whole core demonstration at the FNR

International Nuclear Information System (INIS)

Weha, D.K.; Drumm, C.R.; King, J.S.; Martin, W.R.; Lee, J.C.

1984-01-01

The 2-MW Ford Nuclear Reactor at the University of Michigan is serving as the demonstration reactor for the MTR-type low enrichment (LEU) fuel for the Reduced Enrichment for Research and Test Reactor program. Operational experience gained through six months of LEU core operation and seven months of mixed HEU-LEU core operation is presented. Subcadmium flux measurements performed with rhodium self-powered neutron detectors and iron wire activations are compared with calculations. Measured reactivity parameters are compared for HEU and LEU cores. Finally, the benchmark calculations for several HEU, LEU, and mixed HEU-LEU FNR cores and the International Atomic Energy Agency (IAEA) benchmark problem are presented. (author)

Neutron Fluence, Dosimetry and Damage Response Determination in In-Core/Ex-Core Components of the VENUS CEN/SCK LWR Using 3-D Monte Carlo Simulations: NEA's VENUS-3 Benchmark

International Nuclear Information System (INIS)

Perlado, J. Manuel; Marian, Jaime; Sanz, Jesus Garcia

2000-01-01

Validating state-of-the-art methods used to predict fluence exposure to reactor pressure vessels (RPVs) has become an important issue in identifying the sources of uncertainty in the estimated RPV fluence for pressurized water reactors. This is a very important aspect in evaluating irradiation damage leading to the hardening and embrittlement of such structural components. One of the major benchmark experiments carried out to test three-dimensional methodologies is the VENUS-3 Benchmark Experiment in which three-dimensional Monte Carlo and S n codes have proved more efficient than synthesis methods. At the Instituto de Fusion Nuclear (DENIM) at the Universidad Politecnica de Madrid, a detailed full three-dimensional model of the Venus Critical Facility has been developed making use of the Monte Carlo transport code MCNP4B. The problem geometry and source modeling are described, and results, including calculated versus experimental (C/E) ratios as well as additional studies, are presented. Evidence was found that the great majority of C/E values fell within the 10% tolerance and most within 5%. Tolerance limits are discussed on the basis of evaluated data library and fission spectra sensitivity, where a value ranging between 10 to 15% should be accepted. Also, a calculation of the atomic displacement rate has been carried out in various locations throughout the reactor, finding that values of 0.0001 displacements per atom in external components such as the core barrel are representative of this type of reactor during a 30-yr time span

3-D extension C5G7 MOX benchmark calculation using threedant code

International Nuclear Information System (INIS)

Kim, H.Ch.; Han, Ch.Y.; Kim, J.K.; Na, B.Ch.

2005-01-01

It pursued the benchmark on deterministic 3-D MOX fuel assembly transport calculations without spatial homogenization (C5G7 MOX Benchmark Extension). The goal of this benchmark is to provide a more through test results for the abilities of current available 3-D methods to handle the spatial heterogeneities of reactor core. The benchmark requires solutions in the form of normalized pin powers as well as the eigenvalue for each of the control rod configurations; without rod, with A rods, and with B rods. In this work, the DANTSYS code package was applied to analyze the 3-D Extension C5G7 MOX Benchmark problems. The THREEDANT code within the DANTSYS code package, which solves the 3-D transport equation in x-y-z, and r-z-theta geometries, was employed to perform the benchmark calculations. To analyze the benchmark with the THREEDANT code, proper spatial and angular approximations were made. Several calculations were performed to investigate the effects of the different spatial approximations on the accuracy. The results from these sensitivity studies were analyzed and discussed. From the results, it is found that the 4*4 grid per pin cell is sufficiently refined so that very little benefit is obtained by increasing the mesh size. (authors)

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

Array analyses of SmKS waves and the stratification of Earth's outermost core

Science.gov (United States)

Kaneshima, Satoshi

2018-03-01

We perform array analyses of SmKS waves in order to investigate the Vp structure of the Earth's outermost core. For earthquakes recorded by broadband seismometer networks in the world, we measure differential travel times between S3KS and S2KS, between S4KS and S3KS, and between S5KS and S3KS by array techniques. The differential times are well fit by a Vp model of the Earth's outermost core, KHOMC (Kaneshima and Helffrich, 2013). Differential slownesses of S4KS and S2KS relative to S2KS are also measured for the highest quality data. The measured slownesses, with unique sensitivity to the outer core 200-400 km below the CMB, are matched by KHOMC. These observations consolidate the evidence for the presence at the top of the outer core of a layer that has a distinctively steeper Vp gradient than the bulk of the outer core. We invert new SmKS differential time data set by a tau-p method and attempt to refine the Vp profile of KHOMC. The essential features of KHOMC are preserved after the model refinement. However, the newly estimated layer thickness is nearly 450 km, which is thicker than that of KHOMC. The Vp anomalies relative to PREM for the depths 400-800 km below the CMB are less than 0.03 km/s, consistent with the degree of agreement between different Vp models for the depth range.

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)

Analysis of Homogeneous BFS-73-1 MA Benchmark Core

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

2007-06-15

Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty.

Analysis of Homogeneous BFS-73-1 MA Benchmark Core

International Nuclear Information System (INIS)

Kim, Yeong Il; Yoo, Jae Woon; Song, Hoon; Jang, Jin Wook; Kim, Yeong Il

2007-06-01

Analysis of BFS-73-1 critical assembly for MA transmutation has been carried out by using K-CORE system mainly, DIF3D code. All of measured data are compared with the results of analysis and sensitiveness of calculation conditions, for example, number of neutron energy groups, mesh size used, and analysis method, are assessed. Effective multiplication factor was in good agreement within experimental uncertainty in both transport and diffusion calculations. Fission rate distribution of U-235 and U-238 is also fairly good agreed with experimental results within maximum 5% in core region. But large discrepancy was seen in blanket region and it tends to increase as the location closes to core boundary. Largest error of relative reaction rate ratio was seen in Am-243 fission and U-238 capture. For the case of Am-243, the error lay on appropriate range considering the measurement uncertainty of that as 4.6%. Sample reactivity worths for scattering dominant isotope was greatly differ from the experimental results, which can be explained in terms of sample heterogeneity effect, sample self shielding and finally resonance bilinear correction effect. These effects will be evaluated as future study. C/E of effective delayed neutron fraction is within 4%, which is within the measurement uncertainty

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

The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

International Nuclear Information System (INIS)

Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

2006-01-01

The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

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)

Accelerator driven systems. ADS benchmark calculations. Results of stage 2. Radiotoxic waste transmutation

Energy Technology Data Exchange (ETDEWEB)

Freudenreich, W.E.; Gruppelaar, H

1998-12-01

This report contains the results of calculations made at ECN-Petten of a benchmark to study the neutronic potential of a modular fast spectrum ADS (Accelerator-Driven System) for radiotoxic waste transmutation. The study is focused on the incineration of TRans-Uranium elements (TRU), Minor Actinides (MA) and Long-Lived Fission Products (LLFP), in this case {sup 99}Tc. The benchmark exercise is made in the framework of an IAEA Co-ordinated Research Programme. A simplified description of an ADS, restricted to the reactor part, with TRU or MA fuel (k{sub eff}=0.96) has been analysed. All spectrum calculations have been performed with the Monte Carlo code MCNP-4A. The burnup calculations have been performed with the code FISPACT coupled to MCNP-4A by means of our OCTOPUS system. The cross sections are based upon JEF-2.2 for transport calculations and supplemented with EAF-4 data for inventory calculations. The determined quantities are: core dimensions, fuel inventories, system power, sensitivity on external source spectrum and waste transmutation rates. The main conclusions are: The MA-burner requires only a small accelerator current increase during burnup, in contrast to the TRU-burner. The {sup 99} Tc-burner has a large initial loading; a more effective design may be possible. 5 refs.

Accelerator driven systems. ADS benchmark calculations. Results of stage 2. Radiotoxic waste transmutation

International Nuclear Information System (INIS)

Freudenreich, W.E.; Gruppelaar, H.

1998-12-01

This report contains the results of calculations made at ECN-Petten of a benchmark to study the neutronic potential of a modular fast spectrum ADS (Accelerator-Driven System) for radiotoxic waste transmutation. The study is focused on the incineration of TRans-Uranium elements (TRU), Minor Actinides (MA) and Long-Lived Fission Products (LLFP), in this case 99 Tc. The benchmark exercise is made in the framework of an IAEA Co-ordinated Research Programme. A simplified description of an ADS, restricted to the reactor part, with TRU or MA fuel (k eff =0.96) has been analysed. All spectrum calculations have been performed with the Monte Carlo code MCNP-4A. The burnup calculations have been performed with the code FISPACT coupled to MCNP-4A by means of our OCTOPUS system. The cross sections are based upon JEF-2.2 for transport calculations and supplemented with EAF-4 data for inventory calculations. The determined quantities are: core dimensions, fuel inventories, system power, sensitivity on external source spectrum and waste transmutation rates. The main conclusions are: The MA-burner requires only a small accelerator current increase during burnup, in contrast to the TRU-burner. The 99 Tc-burner has a large initial loading; a more effective design may be possible. 5 refs

Definition of the seventh dynamic AER benchmark-WWER-440 pressure vessel coolant mixing by re-connection of an isolated loop

International Nuclear Information System (INIS)

Kotsarev, A.; Lizorkin, M.; Petrin, R.

2010-01-01

The seventh dynamic benchmark is a continuation of the efforts to validate systematically codes for the estimation of the transient behavior of VVER type nuclear power plants. This benchmark is a continuation of the work in the sixth dynamic benchmark. It is proposed to be simulated the transient - re-connection of an isolated circulating loop with low temperature or low boron concentration in a VVER-440 plant. It is supposed to expand the benchmark to other cases when a different number of loops are in operation leading to different symmetric and asymmetric core boundary conditions. The purposes of the proposed benchmark are: 1) Best-estimate simulations of an transient with a coolant flow mixing in the Reactor Pressure Vessel of WWER-440 plant by re-connection of one coolant loop to the several ones on operation, 2) Performing of code-to-code comparisons. The core is at the end of its first cycle with a power of 1196.25 MWt. The basic additional difference of the 7-seventh benchmark is in the detailed description of the downcomer and bottom part of the reactor vessel that allow describing the effects of coolant mixing in the Reactor Pressure Vessel without any additional conservative assumptions. The burn-up and the power distributions at this reactor state have to be calculated by the participants. The thermohydraulic conditions of the core in the beginning of the transient are specified. Participants self-generated best estimate nuclear data is to be used. The main geometrical parameters of the plant and the characteristics of the control and safety systems are also specified. Use generated input data decks developed for a WWER-440 plant and for the applied codes should be used. The behaviour of the plant should be studied applying coupled system codes, which combine a three-dimensional neutron kinetics description of the core with a pseudo or real 3D thermohydraulics system code. (Authors)

TCA UO2/MOX core analyses

International Nuclear Information System (INIS)

Tahara, Yoshihisa; Noda, Hideyuki

2000-01-01

In order to examine the adequacy of nuclear data, the TCA UO 2 and MOX core experiments were analyzed with MVP using the libraries based on ENDF/B-VI Mod.3 and JENDL-3.2. The ENDF/B-VI data underpredict k eff values. The replacement of 238 U data with the JENDL-3.2 data and the adjustment of 235 ν-value raise the k eff values by 0.3% for UO 2 cores, but still underpredict k eff values. On the other hand, the nuclear data of JENDL-3.2 for H, O, Al, 238 U and 235 U of ENDF/B-VI whose 235 ν-value in thermal energy region is adjusted to the average value of JENDL-3.2 give a good prediction of k eff . (author)

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.

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

Three-dimensional space-time kinetic analysis with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

Energy Technology Data Exchange (ETDEWEB)

Ferroukhi, H.; Coddington, P

2001-03-01

One of the activities within the STARS project, in the Laboratory for Reactor Physics and System Behaviour; is the development of a coupling methodology between the three-dimensional, space-time kinetics codes CORETRAN and RETRAN-3D in order to perform core and plant transient analyses of the Swiss LWRs. The CORETRAN code is a 3-D full-core simulator, intended to be used for core-related analyses, while RETRAN-3D is the three-dimensional kinetics version of the plant system code RETRAN, and can therefore be used for best-estimate analyses of a wide range of transients in both PWRs and BWRs. Because the neutronics solver in both codes is based on the same kinetics model, one important advantage is that the codes can be coupled so that the initial conditions for a RETRAN-3D plant analysis are generated by a detailed-core, steady-state calculation using CORETRAN. As a first step towards using CORETRAN and RETRAN-3D for kinetic applications, the NEACRP PWR rod ejection benchmark has been analyzed with both codes, and is presented in this paper. The first objective is to verify the consistency between the static and kinetic solutions of the two codes, and so gain confidence in the coupling methodology. The second objective is to assess the CORETRAN and RETRAN-3D solutions for a well-defined RIA transient, comparing with previously published results. In parallel, several sensitivity studies have been performed in an attempt to identify models and calculational options important for a correct analysis of an RIA event in a LWR using these two codes. (author)

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1 with MCNP6

International Nuclear Information System (INIS)

Marck, Steven C. van der

2012-01-01

Recent releases of three major world nuclear reaction data libraries, ENDF/B-VII.1, JENDL-4.0, and JEFF-3.1.1, have been tested extensively using benchmark calculations. The calculations were performed with the latest release of the continuous energy Monte Carlo neutronics code MCNP, i.e. MCNP6. 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 2000 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, D2O, H2O, concrete, polyethylene and teflon). The new functionality in MCNP6 to calculate the effective delayed neutron fraction was tested by comparison with more than thirty measurements in widely varying systems. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra. The performance of the three libraries, in combination with MCNP6, is shown to be good. The results for the LEU-COMP-THERM category are on average very close to the benchmark value. Also for most other categories the results are satisfactory. Deviations from the benchmark values do occur in certain benchmark series, or in isolated cases within benchmark series. Such

A highly simplified 3D BWR benchmark problem

International Nuclear Information System (INIS)

Douglass, Steven; Rahnema, Farzad

2010-01-01

The resurgent interest in reactor development associated with the nuclear renaissance has paralleled significant advancements in computer technology, and allowed for unprecedented computational power to be applied to the numerical solution of neutron transport problems. The current generation of core-level solvers relies on a variety of approximate methods (e.g. nodal diffusion theory, spatial homogenization) to efficiently solve reactor problems with limited computer power; however, in recent years, the increased availability of high-performance computer systems has created an interest in the development of new methods and codes (deterministic and Monte Carlo) to directly solve whole-core reactor problems with full heterogeneity (lattice and core level). This paper presents the development of a highly simplified heterogeneous 3D benchmark problem with physics characteristic of boiling water reactors. The aim of this work is to provide a problem for developers to use to validate new whole-core methods and codes which take advantage of the advanced computational capabilities that are now available. Additionally, eigenvalues and an overview of the pin fission density distribution are provided for the benefit of the reader. (author)

Benchmarking of the FENDL-3 Neutron Cross-Section Data Library for Fusion Applications

International Nuclear Information System (INIS)

Fischer, U.; Kondo, K.; Angelone, M.; Batistoni, P.; Villari, R.; Bohm, T.; Sawan, M.; Walker, B.; Konno, C.

2014-03-01

This report summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) with the objective to test and qualify the neutron induced general purpose FENDL-3.0 data library for fusion applications. The benchmark approach consisted of two major steps including the analysis of a simple ITER-like computational benchmark, and a series of analyses of benchmark experiments conducted previously at the 14 MeV neutron generator facilities at ENEA Frascati, Italy (FNG) and JAEA, Tokai-mura, Japan (FNS). The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses analysed. There is a slight trend, however, for an increase of the fast neutron flux in the shielding experiment and a decrease in the breeder mock-up experiments. The photon flux spectra measured in the bulk shield and the tungsten experiments are significantly better reproduced with FENDL-3.0 data. In general, FENDL-3, as compared to FENDL-2.1, shows an improved performance for fusion neutronics applications. It is thus recommended to ITER to replace FENDL-2.1 as reference data library for neutronics calculation by FENDL-3.0. (author)

Benchmark analysis of SPERT-IV reactor with Monte Carlo code MVP

International Nuclear Information System (INIS)

Motalab, M.A.; Mahmood, M.S.; Khan, M.J.H.; Badrun, N.H.; Lyric, Z.I.; Altaf, M.H.

2014-01-01

Highlights: • MVP was used for SPERT-IV core modeling. • Neutronics analysis of SPERT-IV reactor was performed. • Calculation performed to estimate critical rod height, excess reactivity. • Neutron flux, time integrated neutron flux and Cd-ratio also calculated. • Calculated values agree with experimental data. - Abstract: The benchmark experiment of the SPERT-IV D-12/25 reactor core has been analyzed with the Monte Carlo code MVP using the cross-section libraries based on JENDL-3.3. The MVP simulation was performed for the clean and cold core. The estimated values of K eff at the experimental critical rod height and the core excess reactivity were within 5% with the experimental data. Thermal neutron flux profiles at different vertical and horizontal positions of the core were also estimated. Cadmium Ratio at different point of the core was also estimated. All estimated results have been compared with the experimental results. Generally good agreement has been found between experimentally determined and the calculated results

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)

Benchmark Analysis Of The High Temperature Gas Cooled Reactors Using Monte Carlo Technique

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Huda, M.Q.

2008-01-01

Information about several past and present experimental and prototypical facilities based on High Temperature Gas-Cooled Reactor (HTGR) concepts have been examined to assess the potential of these facilities for use in this benchmarking effort. Both reactors and critical facilities applicable to pebble-bed type cores have been considered. Two facilities - HTR-PROTEUS of Switzerland and HTR-10 of China and one conceptual design from Germany - HTR-PAP20 - appear to have the greatest potential for use in benchmarking the codes. This study presents the benchmark analysis of these reactors technologies by using MCNP4C2 and MVP/GMVP Codes to support the evaluation and future development of HTGRs. The ultimate objective of this work is to identify and develop new capabilities needed to support Generation IV initiative. (author)

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.

A SAS2H/KENO-V Methodology for 3D Full Core depletion analysis

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J.; Petrovic, B.

2003-04-01

This paper describes the use of a SAS2H/KENO-V methodology for 3D full core depletion analysis and illustrates its capabilities by applying it to burnup analysis of the IRIS core benchmarks. This new SAS2H/KENO-V sequence combines a 3D Monte Carlo full core calculation of node power distribution and a 1D Wigner-Seitz equivalent cell transport method for independent depletion calculation of each of the nodes. This approach reduces by more than an order of magnitude the time required for getting comparable results using the MOCUP code system. The SAS2H/KENO-V results for the asymmetric IRIS core benchmark are in good agreement with the results of the ALPHA/PHOENIX/ANC code system. (author)

Analysis of the OECD main steam line break benchmark using ANC-K/MIDAC code

International Nuclear Information System (INIS)

Aoki, Shigeaki; Tahara, Yoshihisa; Suemura, Takayuki; Ogawa, Junto

2004-01-01

A three-dimensional (3D) neutronics and thermal-and-hydraulics (T/H) coupling code ANC-K/MIDAC has been developed. It is the combination of the 3D nodal kinetic code ANC-K and the 3D drift flux thermal hydraulic code MIDAC. In order to verify the adequacy of this code, we have performed several international benchmark problems. In this paper, we show the calculation results of ''OECD Main Steam Line Break Benchmark (MSLB benchmark)'', which gives the typical local power peaking problem. And we calculated the return-to-power scenario of the Phase II problem. The comparison of the results shows the very good agreement of important core parameters between the ANC-K/MIDAC and other participant codes. (author)

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.

Determination of Benchmarks Stability within Ahmadu Bello ...

African Journals Online (AJOL)

Heights of six geodetic benchmarks over a total distance of 8.6km at the Ahmadu Bello University (ABU), Zaria, Nigeria were recomputed and analysed using least squares adjustment technique. The network computations were tied to two fix primary reference pillars situated outside the campus. The two-tail Chi-square ...

Benchmark tests of JENDL-1

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki; Hasegawa, Akira; Takano, Hideki; Kamei, Takanobu; Hojuyama, Takeshi; Sasaki, Makoto; Seki, Yuji; Zukeran, Atsushi; Otake, Iwao.

1982-02-01

Various benchmark tests were made on JENDL-1. At the first stage, various core center characteristics were tested for many critical assemblies with one-dimensional model. At the second stage, applicability of JENDL-1 was further tested to more sophisticated problems for MOZART and ZPPR-3 assemblies with two-dimensional model. It was proved that JENDL-1 predicted various quantities of fast reactors satisfactorily as a whole. However, the following problems were pointed out: 1) There exists discrepancy of 0.9% in the k sub(eff)-values between the Pu- and U-cores. 2) The fission rate ratio of 239 Pu to 235 U is underestimated by 3%. 3) The Doppler reactivity coefficients are overestimated by about 10%. 4) The control rod worths are underestimated by 4%. 5) The fission rates of 235 U and 239 Pu are underestimated considerably in the outer core and radial blanket regions. 6) The negative sodium void reactivities are overestimated, when the sodium is removed from the outer core. As a whole, most of problems of JENDL-1 seem to be related with the neutron leakage and the neutron spectrum. It was found through the further study that most of these problems came from too small diffusion coefficients and too large elastic removal cross sections above 100 keV, which might be probably caused by overestimation of the total and elastic scattering cross sections for structural materials in the unresolved resonance region up to several MeV. (author)

FENDL neutronics benchmark: Specifications for the calculational neutronics and shielding benchmark

International Nuclear Information System (INIS)

Sawan, M.E.

1994-12-01

During the IAEA Advisory Group Meeting on ''Improved Evaluations and Integral Data Testing for FENDL'' held in Garching near Munich, Germany in the period 12-16 September 1994, the Working Group II on ''Experimental and Calculational Benchmarks on Fusion Neutronics for ITER'' recommended that a calculational benchmark representative of the ITER design should be developed. This report describes the neutronics and shielding calculational benchmark available for scientists interested in performing analysis for this benchmark. (author)

Moderator temperature effects on reactivity of HEU core of MNSR

International Nuclear Information System (INIS)

Ahmad, Siraj-ul-Islam; Sahibzada, Tasveer Muhammad

2012-01-01

Highlights: ► The MNSR core was analyzed to see the cross section effects on moderator temperature coefficient of reactivity. ► WIMS-D code was used for cell calculations. ► The 3D diffusion theory code PRIDE was first validated using IAEA benchmark problem and then used for analysis of MNSR. ► The differences among results for various libraries were discussed. -- Abstract: In this article we report on analyses that were performed to investigate the influence of cross section differences among libraries released by various centers on reactivity of Miniature Neutron Source Reactors. The 3D model of the core was developed with WIMS-D and PRIDE codes and six cross section libraries were used including JENDL-3.2, JEF-2.2, JEFF-3.3, ENDF/B-VI and ENDF/B-VII, and IAEA library. It was observed that all the libraries predict the reactivity within 10%, with IAEA library giving minimum reactivity worth, and JEF-2.2 data library resulted in highest worth.

Structural and magnetic properties of multi-core nanoparticles analysed using a generalised numerical inversion method

Science.gov (United States)

Bender, P.; Bogart, L. K.; Posth, O.; Szczerba, W.; Rogers, S. E.; Castro, A.; Nilsson, L.; Zeng, L. J.; Sugunan, A.; Sommertune, J.; Fornara, A.; González-Alonso, D.; Barquín, L. Fernández; Johansson, C.

2017-01-01

The structural and magnetic properties of magnetic multi-core particles were determined by numerical inversion of small angle scattering and isothermal magnetisation data. The investigated particles consist of iron oxide nanoparticle cores (9 nm) embedded in poly(styrene) spheres (160 nm). A thorough physical characterisation of the particles included transmission electron microscopy, X-ray diffraction and asymmetrical flow field-flow fractionation. Their structure was ultimately disclosed by an indirect Fourier transform of static light scattering, small angle X-ray scattering and small angle neutron scattering data of the colloidal dispersion. The extracted pair distance distribution functions clearly indicated that the cores were mostly accumulated in the outer surface layers of the poly(styrene) spheres. To investigate the magnetic properties, the isothermal magnetisation curves of the multi-core particles (immobilised and dispersed in water) were analysed. The study stands out by applying the same numerical approach to extract the apparent moment distributions of the particles as for the indirect Fourier transform. It could be shown that the main peak of the apparent moment distributions correlated to the expected intrinsic moment distribution of the cores. Additional peaks were observed which signaled deviations of the isothermal magnetisation behavior from the non-interacting case, indicating weak dipolar interactions. PMID:28397851

Numerical investigation on turbulent natural convection in partially connected cylindrical enclosures for analysing SFR safety under core meltdown scenario

International Nuclear Information System (INIS)

David, Dijo K.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

2015-01-01

Under the unlikely event of severe core meltdown accident in pool type SFR, the molten core materials may rupture the grid plate which supports the fuel subassemblies and it can get relocated in to the lower pool. These debris may eventually settle on the debris collector (i.e., core catcher) installed above the bottom wall of the lower pool. The bed thus formed generates heat due to radioactive decay which has to be passively removed for maintaining the structural integrity of main vessel. By means of natural convection, the heat generated in the debris bed will be transferred to the top pool where the heat sink (i.e., Decay heat exchanger (DHX)) is installed. Heat transfer to the DHX (which is a part of safety grade decay heat removal system) can take place through the opening created in the grid plate which connects the two liquid pools (i.e., the top pool and the lower pool). Heat transfer can also take place through the lateral wall of the lower cylindrical pool to the side pool and eventually to the top pool, and thus to the DHX. This study numerically investigates the effectiveness of heat transfer between lower pool and top pool during PARR by considering them as partially connected cylindrical enclosures. The governing equations have been numerically solved using finite volume method in cylindrical co-ordinates using SIMPLE algorithm. Turbulence has been modeled using k-ω model and the model is validated against benchmark problems of natural convection found in literature. The effect of parameters such as the heat generation rate in the bed and the size of the grid plate opening are evaluated. Also PAHR in SFR pool is modeled using an axi-symmetric model to fund out the influence of grid plate opening on heat removal from core catcher. The results obtained are useful for improving the cooling capability of in-vessel tray type core catcher for handling the whole core meltdown scenarios in SFR. (author)

ES-RBE Event sequence reliability Benchmark exercise

International Nuclear Information System (INIS)

Poucet, A.E.J.

1991-01-01

The event Sequence Reliability Benchmark Exercise (ES-RBE) can be considered as a logical extension of the other three Reliability Benchmark Exercices : the RBE on Systems Analysis, the RBE on Common Cause Failures and the RBE on Human Factors. The latter, constituting Activity No. 1, was concluded by the end of 1987. The ES-RBE covered the techniques that are currently used for analysing and quantifying sequences of events starting from an initiating event to various plant damage states, including analysis of various system failures and/or successes, human intervention failure and/or success and dependencies between systems. By this way, one of the scopes of the ES-RBE was to integrate the experiences gained in the previous exercises

[Benchmarking and other functions of ROM: back to basics].

Science.gov (United States)

Barendregt, M

2015-01-01

Since 2011 outcome data in the Dutch mental health care have been collected on a national scale. This has led to confusion about the position of benchmarking in the system known as routine outcome monitoring (rom). To provide insight into the various objectives and uses of aggregated outcome data. A qualitative review was performed and the findings were analysed. Benchmarking is a strategy for finding best practices and for improving efficacy and it belongs to the domain of quality management. Benchmarking involves comparing outcome data by means of instrumentation and is relatively tolerant with regard to the validity of the data. Although benchmarking is a function of rom, it must be differentiated form other functions from rom. Clinical management, public accountability, research, payment for performance and information for patients are all functions of rom which require different ways of data feedback and which make different demands on the validity of the underlying data. Benchmarking is often wrongly regarded as being simply a synonym for 'comparing institutions'. It is, however, a method which includes many more factors; it can be used to improve quality and has a more flexible approach to the validity of outcome data and is less concerned than other rom functions about funding and the amount of information given to patients. Benchmarking can make good use of currently available outcome data.

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…

Nonlinear Resonance Benchmarking Experiment at the CERN Proton Synchrotron

CERN Document Server

Hofmann, I; Giovannozzi, Massimo; Martini, M; Métral, Elias

2003-01-01

As a first step of a space charge - nonlinear resonance benchmarking experiment over a large number of turns, beam loss and emittance evolution were measured over 1 s on a 1.4 GeV kinetic energy flat-bottom in the presence of a single octupole. By lowering the working point towards the resonance a gradual transition from a loss-free core emittance blow-up to a regime dominated by continuous loss was found. Our 3D simulations with analytical space charge show that trapping on the resonance due to synchrotron oscillation causes the observed core emittance growth as well as halo formation, where the latter is explained as the source of the observed loss.

Review of Seismic Evaluation Methodologies for Nuclear Power Plants Based on a Benchmark Exercise

International Nuclear Information System (INIS)

2013-11-01

Niigataken-chuetsu-oki (NCO) earthquake (Mw = 6.6) occurred on 16 July 2007 and affected the Kashiwazaki-Kariwa (K-K) NPP in Japan. Although there was significant loss of main shock data due to transmission problems, a significant number of instruments were still able to measure the acceleration at different locations in soil (boreholes) and in structures at the K-K NPP during the main shock and the aftershocks. The availability of all these instrumental data provided an excellent background for initiating a benchmarking exercise known as the KAshiwazaki-Kariwa Research Initiative for Seismic Margin Assessment (KARISMA). The main objective of the KARISMA benchmark exercise is to study a comparison between analytical seismic response versus real response of selected structure, system and components (SSCs) of K-K NPP Unit 7. The KARISMA benchmark exercise includes benchmarking the analytical tools and numerical simulation techniques used for predicting seismic response of NPP structures (in linear and non-linear ranges), site response, soil-structure interaction phenomena, seismic response of piping systems, 'sloshing' in the spent fuel pool and buckling of tanks. The benchmark is primarily based on data provided by Tokyo Electric Power Company (TEPCO). It is not linked to the seismic re-evaluation of K-K NPP carried out by TEPCO. Twenty-one organizations, comprising researchers, operating organizations, regulatory authorities, vendors and technical support organizations from 14 countries, participated in the benchmarking exercises. This publication, including a CD-ROM, summarizes the analyses of the main results of the benchmarking exercise for the K-K NPP reactor building (including static and modal analyses of the fixed base model, soil column analyses, analyses of the soil-structure models and margin assessment of the K-K NPP reactor building), the analyses of the main results of the benchmarking exercise for the residual heat removal piping system (including

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

Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

International Nuclear Information System (INIS)

DeVault, G.P.; Bell, C.R.

1985-01-01

Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

Benchmarking health IT among OECD countries: better data for better policy.

Science.gov (United States)

Adler-Milstein, Julia; Ronchi, Elettra; Cohen, Genna R; Winn, Laura A Pannella; Jha, Ashish K

2014-01-01

To develop benchmark measures of health information and communication technology (ICT) use to facilitate cross-country comparisons and learning. The effort is led by the Organisation for Economic Co-operation and Development (OECD). Approaches to definition and measurement within four ICT domains were compared across seven OECD countries in order to identify functionalities in each domain. These informed a set of functionality-based benchmark measures, which were refined in collaboration with representatives from more than 20 OECD and non-OECD countries. We report on progress to date and remaining work to enable countries to begin to collect benchmark data. The four benchmarking domains include provider-centric electronic record, patient-centric electronic record, health information exchange, and tele-health. There was broad agreement on functionalities in the provider-centric electronic record domain (eg, entry of core patient data, decision support), and less agreement in the other three domains in which country representatives worked to select benchmark functionalities. Many countries are working to implement ICTs to improve healthcare system performance. Although many countries are looking to others as potential models, the lack of consistent terminology and approach has made cross-national comparisons and learning difficult. As countries develop and implement strategies to increase the use of ICTs to promote health goals, there is a historic opportunity to enable cross-country learning. To facilitate this learning and reduce the chances that individual countries flounder, a common understanding of health ICT adoption and use is needed. The OECD-led benchmarking process is a crucial step towards achieving this.

Benchmarking Analysis of Institutional University Autonomy in Denmark, Lithuania, Romania, Scotland, and Sweden

DEFF Research Database (Denmark)

This book presents a benchmark, comparative analysis of institutional university autonomy in Denmark, Lithuania, Romania, Scotland and Sweden. These countries are partners in a EU TEMPUS funded project 'Enhancing University Autonomy in Moldova' (EUniAM). This benchmark analysis was conducted...... by the EUniAM Lead Task Force team that collected and analysed secondary and primary data in each of these countries and produced four benchmark reports that are part of this book. For each dimension and interface of institutional university autonomy, the members of the Lead Task Force team identified...... respective evaluation criteria and searched for similarities and differences in approaches to higher education sectors and respective autonomy regimes in these countries. The consolidated report that precedes the benchmark reports summarises the process and key findings from the four benchmark reports...

Benchmarking the Multidimensional Stellar Implicit Code MUSIC

Science.gov (United States)

Goffrey, T.; Pratt, J.; Viallet, M.; Baraffe, I.; Popov, M. V.; Walder, R.; Folini, D.; Geroux, C.; Constantino, T.

2017-04-01

We present the results of a numerical benchmark study for the MUltidimensional Stellar Implicit Code (MUSIC) based on widely applicable two- and three-dimensional compressible hydrodynamics problems relevant to stellar interiors. MUSIC is an implicit large eddy simulation code that uses implicit time integration, implemented as a Jacobian-free Newton Krylov method. A physics based preconditioning technique which can be adjusted to target varying physics is used to improve the performance of the solver. The problems used for this benchmark study include the Rayleigh-Taylor and Kelvin-Helmholtz instabilities, and the decay of the Taylor-Green vortex. Additionally we show a test of hydrostatic equilibrium, in a stellar environment which is dominated by radiative effects. In this setting the flexibility of the preconditioning technique is demonstrated. This work aims to bridge the gap between the hydrodynamic test problems typically used during development of numerical methods and the complex flows of stellar interiors. A series of multidimensional tests were performed and analysed. Each of these test cases was analysed with a simple, scalar diagnostic, with the aim of enabling direct code comparisons. As the tests performed do not have analytic solutions, we verify MUSIC by comparing it to established codes including ATHENA and the PENCIL code. MUSIC is able to both reproduce behaviour from established and widely-used codes as well as results expected from theoretical predictions. This benchmarking study concludes a series of papers describing the development of the MUSIC code and provides confidence in future applications.

Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

Energy Technology Data Exchange (ETDEWEB)

Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)

2017-03-15

Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

International Nuclear Information System (INIS)

Badea, Aurelian F.; Cacuci, Dan G.

2017-01-01

Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

INL Results for Phases I and III of the OECD/NEA MHTGR-350 Benchmark

Energy Technology Data Exchange (ETDEWEB)

Gerhard Strydom; Javier Ortensi; Sonat Sen; Hans Hammer

2013-09-01

The Idaho National Laboratory (INL) Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Methods Core Simulation group led the construction of the Organization for Economic Cooperation and Development (OECD) Modular High Temperature Reactor (MHTGR) 350 MW benchmark for comparing and evaluating prismatic VHTR analysis codes. The benchmark is sponsored by the OECD's Nuclear Energy Agency (NEA), and the project will yield a set of reference steady-state, transient, and lattice depletion problems that can be used by the Department of Energy (DOE), the Nuclear Regulatory Commission (NRC), and vendors to assess their code suits. The Methods group is responsible for defining the benchmark specifications, leading the data collection and comparison activities, and chairing the annual technical workshops. This report summarizes the latest INL results for Phase I (steady state) and Phase III (lattice depletion) of the benchmark. The INSTANT, Pronghorn and RattleSnake codes were used for the standalone core neutronics modeling of Exercise 1, and the results obtained from these codes are compared in Section 4. Exercise 2 of Phase I requires the standalone steady-state thermal fluids modeling of the MHTGR-350 design, and the results for the systems code RELAP5-3D are discussed in Section 5. The coupled neutronics and thermal fluids steady-state solution for Exercise 3 are reported in Section 6, utilizing the newly developed Parallel and Highly Innovative Simulation for INL Code System (PHISICS)/RELAP5-3D code suit. Finally, the lattice depletion models and results obtained for Phase III are compared in Section 7. The MHTGR-350 benchmark proved to be a challenging simulation set of problems to model accurately, and even with the simplifications introduced in the benchmark specification this activity is an important step in the code-to-code verification of modern prismatic VHTR codes. A final OECD/NEA comparison report will compare the Phase I and III

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)

Beyond-CMOS Device Benchmarking for Boolean and Non-Boolean Logic Applications

OpenAIRE

Pan, Chenyun; Naeemi, Azad

2017-01-01

The latest results of benchmarking research are presented for a variety of beyond-CMOS charge- and spin-based devices. In addition to improving the device-level models, several new device proposals and a few majorly modified devices are investigated. Deep pipelining circuits are employed to boost the throughput of low-power devices. Furthermore, the benchmarking methodology is extended to interconnect-centric analyses and non-Boolean logic applications. In contrast to Boolean circuits, non-Bo...

Discrepancies in Communication Versus Documentation of Weight-Management Benchmarks

Directory of Open Access Journals (Sweden)

Christy B. Turer MD, MHS

2017-02-01

Full Text Available To examine gaps in communication versus documentation of weight-management clinical practices, communication was recorded during primary care visits with 6- to 12-year-old overweight/obese Latino children. Communication/documentation content was coded by 3 reviewers using communication transcripts and health-record documentation. Discrepancies in communication/documentation content codes were resolved through consensus. Bivariate/multivariable analyses examined factors associated with discrepancies in benchmark communication/documentation. Benchmarks were neither communicated nor documented in up to 42% of visits, and communicated but not documented or documented but not communicated in up to 20% of visits. Lowest benchmark performance rates were for laboratory studies (35% and nutrition/weight-management referrals (42%. In multivariable analysis, overweight (vs obesity was associated with 1.6 more discrepancies in communication versus documentation (P = .03. Many weight-management benchmarks are not met, not documented, or performed without being communicated. Enhanced communication with families and documentation in health records may promote lifestyle changes in overweight children and higher quality care for overweight children in primary care.

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.

Towards an Industrial Application of Statistical Uncertainty Analysis Methods to Multi-physical Modelling and Safety Analyses

International Nuclear Information System (INIS)

Zhang, Jinzhao; Segurado, Jacobo; Schneidesch, Christophe

2013-01-01

Since 1980's, Tractebel Engineering (TE) has being developed and applied a multi-physical modelling and safety analyses capability, based on a code package consisting of the best estimate 3D neutronic (PANTHER), system thermal hydraulic (RELAP5), core sub-channel thermal hydraulic (COBRA-3C), and fuel thermal mechanic (FRAPCON/FRAPTRAN) codes. A series of methodologies have been developed to perform and to license the reactor safety analysis and core reload design, based on the deterministic bounding approach. Following the recent trends in research and development as well as in industrial applications, TE has been working since 2010 towards the application of the statistical sensitivity and uncertainty analysis methods to the multi-physical modelling and licensing safety analyses. In this paper, the TE multi-physical modelling and safety analyses capability is first described, followed by the proposed TE best estimate plus statistical uncertainty analysis method (BESUAM). The chosen statistical sensitivity and uncertainty analysis methods (non-parametric order statistic method or bootstrap) and tool (DAKOTA) are then presented, followed by some preliminary results of their applications to FRAPCON/FRAPTRAN simulation of OECD RIA fuel rod codes benchmark and RELAP5/MOD3.3 simulation of THTF tests. (authors)

Performance of Artificial Intelligence Workloads on the Intel Core 2 Duo Series Desktop Processors

Directory of Open Access Journals (Sweden)

Abdul Kareem PARCHUR

2010-12-01

Full Text Available As the processor architecture becomes more advanced, Intel introduced its Intel Core 2 Duo series processors. Performance impact on Intel Core 2 Duo processors are analyzed using SPEC CPU INT 2006 performance numbers. This paper studied the behavior of Artificial Intelligence (AI benchmarks on Intel Core 2 Duo series processors. Moreover, we estimated the task completion time (TCT @1 GHz, @2 GHz and @3 GHz Intel Core 2 Duo series processors frequency. Our results show the performance scalability in Intel Core 2 Duo series processors. Even though AI benchmarks have similar execution time, they have dissimilar characteristics which are identified using principal component analysis and dendogram. As the processor frequency increased from 1.8 GHz to 3.167 GHz the execution time is decreased by ~370 sec for AI workloads. In the case of Physics/Quantum Computing programs it was ~940 sec.

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:

Calculations with ANSYS/FLOTRAN to a core catcher benchmark

International Nuclear Information System (INIS)

Willschuetz, H.G.

1999-01-01

There are numerous experiments for the exploration of the corium spreading behaviour, but comparable data have not been available up to now in the field of the long-term behaviour of a corium expanded in a core catcher. For the calculations a pure liquid oxidic melt with a homogeneous internal heat source was assumed. The melt was distributed uniformly over the spreading area of the EPR core catcher. All codes applied the well known k-ε-turbulence-model to simulate the turbulent flow regime of this melt configuration. While the FVM-code calculations were performed with three dimensional models using a simple symmetry, the problem was modelled two-dimensionally with ANSYS due to limited CPU performance. In addition, the 2D results of ANSYS should allow a comparison for the planned second stage of the calculations. In this second stage, the behaviour of a segregated metal oxide melt should be examined. However, first estimates and pre-calculations showed that a 3D simulation of the problem is not possible with any of the codes due to lacking computer performance. (orig.)

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...

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

Discussion of OECD LWR Uncertainty Analysis in Modelling Benchmark

International Nuclear Information System (INIS)

Ivanov, K.; Avramova, M.; Royer, E.; Gillford, J.

2013-01-01

The demand for best estimate calculations in nuclear reactor design and safety evaluations has increased in recent years. Uncertainty quantification has been highlighted as part of the best estimate calculations. The modelling aspects of uncertainty and sensitivity analysis are to be further developed and validated on scientific grounds in support of their performance and application to multi-physics reactor simulations. The Organization for Economic Co-operation and Development (OECD) / Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC) has endorsed the creation of an Expert Group on Uncertainty Analysis in Modelling (EGUAM). Within the framework of activities of EGUAM/NSC the OECD/NEA initiated the Benchmark for Uncertainty Analysis in Modelling for Design, Operation, and Safety Analysis of Light Water Reactor (OECD LWR UAM benchmark). The general objective of the benchmark is to propagate the predictive uncertainties of code results through complex coupled multi-physics and multi-scale simulations. The benchmark is divided into three phases with Phase I highlighting the uncertainty propagation in stand-alone neutronics calculations, while Phase II and III are focused on uncertainty analysis of reactor core and system respectively. This paper discusses the progress made in Phase I calculations, the Specifications for Phase II and the incoming challenges in defining Phase 3 exercises. The challenges of applying uncertainty quantification to complex code systems, in particular the time-dependent coupled physics models are the large computational burden and the utilization of non-linear models (expected due to the physics coupling). (authors)

Piping benchmark problems for the Westinghouse AP600 Standardized Plant

International Nuclear Information System (INIS)

Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

1997-01-01

To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the Westinghouse AP600 Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the AP600 standard design. It will be required that the combined license licensees demonstrate that their solutions to these problems are in agreement with the benchmark problem set

CMFD and GPU acceleration on method of characteristics for hexagonal cores

International Nuclear Information System (INIS)

Han, Yu; Jiang, Xiaofeng; Wang, Dezhong

2014-01-01

Highlights: • A merged hex-mesh CMFD method solved via tri-diagonal matrix inversion. • Alternative hardware acceleration of using inexpensive GPU. • A hex-core benchmark with solution to confirm two acceleration methods. - Abstract: Coarse Mesh Finite Difference (CMFD) has been widely adopted as an effective way to accelerate the source iteration of transport calculation. However in a core with hexagonal assemblies there are non-hexagonal meshes around the edges of assemblies, causing a problem for CMFD if the CMFD equations are still to be solved via tri-diagonal matrix inversion by simply scanning the whole core meshes in different directions. To solve this problem, we propose an unequal mesh CMFD formulation that combines the non-hexagonal cells on the boundary of neighboring assemblies into non-regular hexagonal cells. We also investigated the alternative hardware acceleration of using graphics processing units (GPU) with graphics card in a personal computer. The tool CUDA is employed, which is a parallel computing platform and programming model invented by the company NVIDIA for harnessing the power of GPU. To investigate and implement these two acceleration methods, a 2-D hexagonal core transport code using the method of characteristics (MOC) is developed. A hexagonal mini-core benchmark problem is established to confirm the accuracy of the MOC code and to assess the effectiveness of CMFD and GPU parallel acceleration. For this benchmark problem, the CMFD acceleration increases the speed 16 times while the GPU acceleration speeds it up 25 times. When used simultaneously, they provide a speed gain of 292 times

CMFD and GPU acceleration on method of characteristics for hexagonal cores

Energy Technology Data Exchange (ETDEWEB)

Han, Yu, E-mail: hanyu1203@gmail.com [School of Nuclear Science and Engineering, Shanghai Jiaotong University, Shanghai 200240 (China); Jiang, Xiaofeng [Shanghai NuStar Nuclear Power Technology Co., Ltd., No. 81 South Qinzhou Road, XuJiaHui District, Shanghai 200000 (China); Wang, Dezhong [School of Nuclear Science and Engineering, Shanghai Jiaotong University, Shanghai 200240 (China)

2014-12-15

Highlights: • A merged hex-mesh CMFD method solved via tri-diagonal matrix inversion. • Alternative hardware acceleration of using inexpensive GPU. • A hex-core benchmark with solution to confirm two acceleration methods. - Abstract: Coarse Mesh Finite Difference (CMFD) has been widely adopted as an effective way to accelerate the source iteration of transport calculation. However in a core with hexagonal assemblies there are non-hexagonal meshes around the edges of assemblies, causing a problem for CMFD if the CMFD equations are still to be solved via tri-diagonal matrix inversion by simply scanning the whole core meshes in different directions. To solve this problem, we propose an unequal mesh CMFD formulation that combines the non-hexagonal cells on the boundary of neighboring assemblies into non-regular hexagonal cells. We also investigated the alternative hardware acceleration of using graphics processing units (GPU) with graphics card in a personal computer. The tool CUDA is employed, which is a parallel computing platform and programming model invented by the company NVIDIA for harnessing the power of GPU. To investigate and implement these two acceleration methods, a 2-D hexagonal core transport code using the method of characteristics (MOC) is developed. A hexagonal mini-core benchmark problem is established to confirm the accuracy of the MOC code and to assess the effectiveness of CMFD and GPU parallel acceleration. For this benchmark problem, the CMFD acceleration increases the speed 16 times while the GPU acceleration speeds it up 25 times. When used simultaneously, they provide a speed gain of 292 times.

JENDL-3.2 performance in analyses of MISTRAL critical experiments for high-moderation MOX cores

International Nuclear Information System (INIS)

Takada, Naoyuki; Hibi, Koki; Ishii, Kazuya; Ando, Yoshihira; Yamamoto, Toru; Ueji, Masao; Iwata, Yutaka

2001-01-01

NUPEC and CEA have launched an extensive experimental program called MISTRAL to study highly moderated MOX cores for the advanced LWRs. The analyses using SRAC system and MVP code with JENDL-3.2 library are in progress on the experiments of the MISTRAL and the former EPICURE programs. Various comparisons have been made between calculation results and measurement values. (author)

Overview of cooperative international piping benchmark analyses

International Nuclear Information System (INIS)

McAfee, W.J.

1982-01-01

This paper presents an overview of an effort initiated in 1976 by the International Working Group on Fast Reactors (IWGFR) of the International Atomic Energy Agency (IAEA) to evaluate detailed and simplified inelastic analysis methods for piping systems with particular emphasis on piping bends. The procedure was to collect from participating member IAEA countries descriptions of tests and test results for piping systems or bends (with emphasis on high temperature inelastic tests), to compile, evaluate, and issue a selected number of these problems for analysis, and to compile and make a preliminary evaluation of the analyses results. Of the problem descriptions submitted three were selected to be used: a 90 0 -elbow at 600 0 C with an in-plane transverse force; a 90 0 -elbow with an in-plane moment; and a 180 0 -elbow at room temperature with a reversed, cyclic, in-plane transverse force. A variety of both detailed and simplified analysis solutions were obtained. A brief comparative assessment of the analyses is contained in this paper. 15 figures

Validations of BWR nuclear design code using ABWR MOX numerical benchmark problems

International Nuclear Information System (INIS)

Takano, Shou; Sasagawa, Masaru; Yamana, Teppei; Ikehara, Tadashi; Yanagisawa, Naoki

2017-01-01

BWR core design code package (the HINES assembly code and the PANACH core simulator), being used for full MOX-ABWR core design, has been benchmarked against the high-fidelity numerical solutions as references, for the purpose of validating its capability of predicting the BWR core design parameters systematically from UO 2 to 100% MOX cores. The reference solutions were created by whole core critical calculations using MCNPs with the precisely modeled ABWR cores both in hot and cold conditions at BOC and EOC of the equilibrium cycle. A Doppler-Broadening Rejection Correction (DCRB) implemented MCNP5-1.4 with ENDF/B-VII.0 was mainly used to evaluate the core design parameters, except for effective delayed neutron fraction (β eff ) and prompt neutron lifetime (l) with MCNP6.1. The discrepancies in the results between the design codes HINES-PANACH and MCNPs for the core design parameters such as the bundle powers, hot pin powers, control rod worth, boron worth, void reactivity, Doppler reactivity, β eff and l, are almost within target accuracy, leading to the conclusion that HINES-PANACH has sufficient fidelity for application to full MOX-ABWR core design. (author)

Benchmarking of the FENDL-3 Neutron Cross-section Data Starter Library for Fusion Applications

Energy Technology Data Exchange (ETDEWEB)

Fischer, U., E-mail: ulrich.fischer@kit.edu [Association KIT-Euratom, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Angelone, M. [Associazione ENEA-Euratom, ENEA Fusion Division, Via E. Fermi 27, I-00044 Frascati (Italy); Bohm, T. [University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706 (United States); Kondo, K. [Association KIT-Euratom, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Konno, C. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Sawan, M. [University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706 (United States); Villari, R. [Associazione ENEA-Euratom, ENEA Fusion Division, Via E. Fermi 27, I-00044 Frascati (Italy); Walker, B. [University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706 (United States)

2014-06-15

This paper summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) on a computational ITER benchmark and a series of 14 MeV neutron benchmark experiments. The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses. In general, FENDL-3 shows an improved performance for fusion neutronics applications.

Piping benchmark problems for the ABB/CE System 80+ Standardized Plant

International Nuclear Information System (INIS)

Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

1994-07-01

To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the ABB/Combustion Engineering System 80+ Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the System 80+ standard design. It will be required that the combined license licensees demonstrate that their solution to these problems are in agreement with the benchmark problem set. The first System 80+ piping benchmark is a uniform support motion response spectrum solution for one section of the feedwater piping subjected to safe shutdown seismic loads. The second System 80+ piping benchmark is a time history solution for the feedwater piping subjected to the transient loading induced by a water hammer. The third System 80+ piping benchmark is a time history solution of the pressurizer surge line subjected to the accelerations induced by a main steam line pipe break. The System 80+ reactor is an advanced PWR type

Benchmarking and performance enhancement framework for multi-staging object-oriented languages

Directory of Open Access Journals (Sweden)

Ahmed H. Yousef

2013-06-01

Full Text Available This paper focuses on verifying the readiness, feasibility, generality and usefulness of multi-staging programming in software applications. We present a benchmark designed to evaluate the performance gain of different multi-staging programming (MSP languages implementations of object oriented languages. The benchmarks in this suite cover different tests that range from classic simple examples (like matrix algebra to advanced examples (like encryption and image processing. The benchmark is applied to compare the performance gain of two different MSP implementations (Mint and Metaphor that are built on object oriented languages (Java and C# respectively. The results concerning the application of this benchmark on these languages are presented and analysed. The measurement technique used in benchmarking leads to the development of a language independent performance enhancement framework that allows the programmer to select which code segments need staging. The framework also enables the programmer to verify the effectiveness of staging on the application performance. The framework is applied to a real case study. The case study results showed the effectiveness of the framework to achieve significant performance enhancement.

Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

International Nuclear Information System (INIS)

Heard, F.J.

1997-01-01

The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

The database 'EDUD Base' for validation of neutron-physics codes used to analyze the WWER-440 cores

International Nuclear Information System (INIS)

Rocek, J.; Belac, J.; Miasnikov, A.

2003-01-01

The program and data system EDUDBase for validation of reactor computing codes was developed at NRI. It is designed for validation and evaluation of the precision of different computer codes used for WWER core analyses. The main goal of this database is to provide data for comparison with calculation results of tested codes and tools for statistical analysis or differences between the calculation results and the test data. The benchmark data sets are based on in-core measurements performed on WWER-440 reactors of Dukovany NPP. The initial data from NPP are verified, errors and inaccuracies are eliminated and data are transferred to a form, which is suitable for comparison with results of calculations. A special reduced operating history data set is created for each operating cycle ('Benchmark Operation History') to be used as an input data for calculation. It contains values of some integral quantities for each time point: effective time, integral thermal power, boron concentration, position of working group control assemblies (group 6) and inlet coolant temperature. At present, sets are available for all completed cycles up to: (unit/cycle) 1/17, 2/16, 3/15, 4/15. Power distribution is described for approx. 40 time steps during each operating cycle. 2D-power distributions are transferred into 60-degree core symmetry sector of reactor core. At present, such data sets are available only for later cycles starting with: (unit/cycle) 1/7, 2/6, 3/5, 4/5 (in other words last II cycles for each unit) (Authors)

Development of a set of benchmark problems to verify numerical methods for solving burnup equations

International Nuclear Information System (INIS)

Lago, Daniel; Rahnema, Farzad

2017-01-01

Highlights: • Description transmutation chain benchmark problems. • Problems for validating numerical methods for solving burnup equations. • Analytical solutions for the burnup equations. • Numerical solutions for the burnup equations. - Abstract: A comprehensive set of transmutation chain benchmark problems for numerically validating methods for solving burnup equations was created. These benchmark problems were designed to challenge both traditional and modern numerical methods used to solve the complex set of ordinary differential equations used for tracking the change in nuclide concentrations over time due to nuclear phenomena. Given the development of most burnup solvers is done for the purpose of coupling with an established transport solution method, these problems provide a useful resource in testing and validating the burnup equation solver before coupling for use in a lattice or core depletion code. All the relevant parameters for each benchmark problem are described. Results are also provided in the form of reference solutions generated by the Mathematica tool, as well as additional numerical results from MATLAB.

Two-dimensional full-core transport theory Benchmarks for the WWER reactors

International Nuclear Information System (INIS)

Petkov, P.T.

2002-01-01

Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

Interim results of the sixth three-dimensional AER dynamic benchmark problem calculation. Solution of problem with DYN3D and RELAP5-3D codes

International Nuclear Information System (INIS)

Hadek, J.; Kral, P.; Macek, J.

2001-01-01

The paper gives a brief survey of the 6 th three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAPS-3D at NRI Rez. This benchmark was defined at the 10 th AER Symposium. Its initiating event is a double ended break in the steam line of steam generator No. I in a WWER-440/213 plant at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations as well as tuning of initial state before the transient were performed with the code DYN3D. Transient calculations were made with the system code RELAPS-3D.The KASSETA library was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP Dukovany was adopted for the 6 th AER dynamic benchmark purposes. The RELAPS-3D full core neutronic model was connected with seven coolant channels thermal-hydraulic model of the core (Authors)

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...

Benchmark test of evaluated nuclear data files for fast reactor neutronics application

International Nuclear Information System (INIS)

Chiba, Go; Hazama, Taira; Iwai, Takehiko; Numata, Kazuyuki

2007-07-01

A benchmark test of the latest evaluated nuclear data files, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0, has been carried out for fast reactor neutronics application. For this benchmark test, experimental data obtained at fast critical assemblies and fast power reactors are utilized. In addition to comparing of numerical solutions with the experimental data, we have extracted several cross sections, in which differences between three nuclear data files affect significantly numerical solutions, by virtue of sensitivity analyses. This benchmark test concludes that ENDF/B-VII.0 predicts well the neutronics characteristics of fast neutron systems rather than the other nuclear data files. (author)

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...

Jendl-3.1 iron validation on the PCA-REPLICA (H2O/Fe) shielding benchmark experiment

International Nuclear Information System (INIS)

Pescarini, M.; Borgia, M. G.

1997-03-01

The PCA-REPLICA (H 2 O/Fe) neutron shielding benchmarks experiment is analysed using the SN 2-D DOT 3.5-E code and the 3-D-equivalent flux synthesis method. This engineering benchmark reproduces the ex-core radial geometry of a PWR, including a mild steel reactor pressure vessel (RPV) simulator, and is designed to test the accuracy of the calculation of the in-vessel neutron exposure parameters. This accuracy is strongly dependent on the quality of the iron neutron cross sections used to describe the nuclear reactions within the RPV simulator. In particular, in this report, the cross sections based on the JENDL-3.1 iron data files are tested, through a comparison of the calculated integral and spectral results with the corresponding experimental data. In addition, the present results are compared, on the same benchmark experiment, with those of a preceding ENEA-Bologna validation of the ENDF/B VI iron cross sections. The integral result comparison indicates that, for all the threshold detectors considered (Rh-103 (n, n') Rh-103m, In-115 (n, n') In-115m and S-32 (n, p) P-32), the JENDL-3.1 natural iron data produce satisfactory results similar to those obtained with the ENDF/B VI iron data. On the contrary, when the JENDL/3.1 Fe-56 data file is used, strongly underestimated results are obtained for the lower energy threshold detectors, Rh-103 and In-115. This fact, in particular, becomes more evident with increasing the neutron penetration depth in the RPV simulator

Three-dimensional coupled kinetics/thermal- hydraulic benchmark TRIGA experiments

International Nuclear Information System (INIS)

Feltus, Madeline Anne; Miller, William Scott

2000-01-01

This research project provides separate effects tests in order to benchmark neutron kinetics models coupled with thermal-hydraulic (T/H) models used in best-estimate codes such as the Nuclear Regulatory Commission's (NRC) RELAP and TRAC code series and industrial codes such as RETRAN. Before this research project was initiated, no adequate experimental data existed for reactivity initiated transients that could be used to assess coupled three-dimensional (3D) kinetics and 3D T/H codes which have been, or are being developed around the world. Using various Test Reactor Isotope General Atomic (TRIGA) reactor core configurations at the Penn State Breazeale Reactor (PSBR), it is possible to determine the level of neutronics modeling required to describe kinetics and T/H feedback interactions. This research demonstrates that the small compact PSBR TRIGA core does not necessarily behave as a point kinetics reactor, but that this TRIGA can provide actual test results for 3D kinetics code benchmark efforts. This research focused on developing in-reactor tests that exhibited 3D neutronics effects coupled with 3D T/H feedback. A variety of pulses were used to evaluate the level of kinetics modeling needed for prompt temperature feedback in the fuel. Ramps and square waves were used to evaluate the detail of modeling needed for the delayed T/H feedback of the coolant. A stepped ramp was performed to evaluate and verify the derived thermal constants for the specific PSBR TRIGA core loading pattern. As part of the analytical benchmark research, the STAR 3D kinetics code (, STAR: Space and time analysis of reactors, Version 5, Level 3, Users Guide, Yankee Atomic Electric Company, YEAC 1758, Bolton, MA) was used to model the transient experiments. The STAR models were coupled with the one-dimensional (1D) WIGL and LRA and 3D COBRA (, COBRA IIIC: A digital computer program for steady-state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements, Battelle

Modal analysis and acoustic transmission through offset-core honeycomb sandwich panels

Science.gov (United States)

Mathias, Adam Dustin

The work presented in this thesis is motivated by an earlier research that showed that double, offset-core honeycomb sandwich panels increased thermal resistance and, hence, decreased heat transfer through the panels. This result lead to the hypothesis that these panels could be used for acoustic insulation. Using commercial finite element modeling software, COMSOL Multiphysics, the acoustical properties, specifically the transmission loss across a variety of offset-core honeycomb sandwich panels, is studied for the case of a plane acoustic wave impacting the panel at normal incidence. The transmission loss results are compared with those of single-core honeycomb panels with the same cell sizes. The fundamental frequencies of the panels are also computed in an attempt to better understand the vibrational modes of these particular sandwich-structured panels. To ensure that the finite element analysis software is adequate for the task at hand, two relevant benchmark problems are solved and compared with theory. Results from these benchmark results compared well to those obtained from theory. Transmission loss results from the offset-core honeycomb sandwich panels show increased transmission loss, especially for large cell honeycombs when compared to single-core honeycomb panels.

Benchmark physics tests in the metallic-fuelled assembly ZPPR-15

International Nuclear Information System (INIS)

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

1987-01-01

In the last two years a shift in emphasis to inherent safety and economic competitiveness has led to a resurgence in US interest in metallic-alloy fuels for LMRs. Argonne National Laboratory initiated an extensive testing program for metallic-fuelled LMR technology that has included benchmark physics as one component. The tests done in the ZPPR-15 Program produced the first physics results in over 20 years for a metal-composition LMR core

Validation of the BUGJEFF311.BOLIB, BUGENDF70.BOLIB and BUGLE-B7 broad-group libraries on the PCA-Replica (H2O/Fe) neutron shielding benchmark experiment

OpenAIRE

Pescarini Massimo; Orsi Roberto; Frisoni Manuela

2016-01-01

The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the TORT-3.2 3D SN code. PCA-Replica reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and U...

Development of three dimensional transient analysis code STTA for SCWR core

International Nuclear Information System (INIS)

Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping

2015-01-01

Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation

Finite element program ARKAS: verification for IAEA benchmark problem analysis on core-wide mechanical analysis of LMFBR cores

International Nuclear Information System (INIS)

Nakagawa, M.; Tsuboi, Y.

1990-01-01

''ARKAS'' code verification, with the problems set in the International Working Group on Fast Reactors (IWGFR) Coordinated Research Programme (CRP) on the inter-comparison between liquid metal cooled fast breeder reactor (LMFBR) Core Mechanics Codes, is discussed. The CRP was co-ordinated by the IWGFR around problems set by Dr. R.G. Anderson (UKAEA) and arose from the IWGFR specialists' meeting on The Predictions and Experience of Core Distortion Behaviour (ref. 2). The problems for the verification (''code against code'') and validation (''code against experiment'') were set and calculated by eleven core mechanics codes from nine countries. All the problems have been completed and were solved with the core structural mechanics code ARKAS. Predictions by ARKAS agreed very well with other solutions for the well-defined verification problems. For the validation problems based on Japanese ex-reactor 2-D thermo-elastic experiments, the agreements between measured and calculated values were fairly good. This paper briefly describes the numerical model of the ARKAS code, and discusses some typical results. (author)

Effects of neutron data libraries and criticality codes on IAEA criticality benchmark problems

International Nuclear Information System (INIS)

Sarker, Md.M.; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka

1993-10-01

In order to compare the effects of neutron data libraries and criticality codes to thermal reactors (LWR), the IAEA criticality benchmark calculations have been performed. The experiments selected in this study include TRX-1 and TRX-2 with a simple geometric configuration. Reactor lattice calculation codes WIMS-D/4, MCNP-4, JACS (MGCL, KENO), and SRAC were used in the present calculations. The TRX cores were analyzed by WIMS-D/4 using WIMS original library and also by MCNP-4, JACS (MGCL, KENO), and SRAC using the libraries generated from JENDL-3 and ENDF/B-IV nuclear data files. An intercomparison work for the above mentioned code systems and cross section libraries was performed by analyzing the LWR benchmark experiments TRX-1 and TRX-2. The TRX cores were also analyzed for supercritical and subcritical conditions and these results were compared. In the case of critical condition, the results were in good agreement. But for the supercritical and subcritical conditions, the difference of the results obtained by using the different cross section libraries become larger than for the critical condition. (author)

International benchmarking of electricity transmission by regulators: A contrast between theory and practice?

International Nuclear Information System (INIS)

Haney, Aoife Brophy; Pollitt, Michael G.

2013-01-01

Benchmarking of electricity networks has a key role in sharing the benefits of efficiency improvements with consumers and ensuring regulated companies earn a fair return on their investments. This paper analyses and contrasts the theory and practice of international benchmarking of electricity transmission by regulators. We examine the literature relevant to electricity transmission benchmarking and discuss the results of a survey of 25 national electricity regulators. While new panel data techniques aimed at dealing with unobserved heterogeneity and the validity of the comparator group look intellectually promising, our survey suggests that they are in their infancy for regulatory purposes. In electricity transmission, relative to electricity distribution, choosing variables is particularly difficult, because of the large number of potential variables to choose from. Failure to apply benchmarking appropriately may negatively affect investors’ willingness to invest in the future. While few of our surveyed regulators acknowledge that regulatory risk is currently an issue in transmission benchmarking, many more concede it might be. In the meantime new regulatory approaches – such as those based on tendering, negotiated settlements, a wider range of outputs or longer term grid planning – are emerging and will necessarily involve a reduced role for benchmarking. -- Highlights: •We discuss how to benchmark electricity transmission. •We report survey results from 25 national energy regulators. •Electricity transmission benchmarking is more challenging than benchmarking distribution. •Many regulators concede benchmarking may raise capital costs. •Many regulators are considering new regulatory approaches

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....

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

Udsættelser af lejere – Udvikling og benchmarking

DEFF Research Database (Denmark)

Christensen, Gunvor; Jeppesen, Anders Gade; Kjær, Agnete Aslaug

I denne rapport undersøges udviklingen i både fogedsager og effektive udsættelser fra 2007-2013. Rapporten indeholder desuden en benchmarking-analyse, der estimerer, om hver enkelt kommune har flere eller færre effektive udsættelser, end hvad man skulle forvente, når der tages højde for bl...... kommuner, der indgår i benchmarking-analysen har desuden flere effektive udsættelser i de almene boliger, end man kunne forvente, når der tages hensyn til kommunernes befolkningsgrundlag, det lokale boligmarked og kommunale forhold som fx størrelsen af kommunen. Rapporten er finansieret af Ministeriet...

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…

Core design and fuel rod analyses of a super fast reactor with high power density

International Nuclear Information System (INIS)

Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

2009-01-01

A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

1993-08-01

To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set

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...

The Monte Carlo performance benchmark test - AIMS, specifications and first results

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J. Eduard, E-mail: j.e.hoogenboom@tudelft.nl [Faculty of Applied Sciences, Delft University of Technology (Netherlands); Martin, William R., E-mail: wrm@umich.edu [Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI (United States); Petrovic, Bojan, E-mail: Bojan.Petrovic@gatech.edu [Nuclear and Radiological Engineering, Georgia Institute of Technology, Atlanta, GA (United States)

2011-07-01

The Monte Carlo performance benchmark for detailed power density calculation in a full-size reactor core is organized under the auspices of the OECD NEA Data Bank. It aims at monitoring over a range of years the increase in performance, measured in terms of standard deviation and computer time, of Monte Carlo calculation of the power density in small volumes. A short description of the reactor geometry and composition is discussed. One of the unique features of the benchmark exercise is the possibility to upload results from participants at a web site of the NEA Data Bank which enables online analysis of results and to graphically display how near we are at the goal of doing a detailed power distribution calculation with acceptable statistical uncertainty in an acceptable computing time. First results are discussed which show that 10 to 100 billion histories must be simulated to reach a standard deviation of a few percent in the estimated power of most of the requested the fuel zones. Even when using a large supercomputer, a considerable speedup is still needed to reach the target of 1 hour computer time. An outlook is given of what to expect from this benchmark exercise over the years. Possible extensions of the benchmark for specific issues relevant in current Monte Carlo calculation for nuclear reactors are also discussed. (author)

The Monte Carlo performance benchmark test - AIMS, specifications and first results

International Nuclear Information System (INIS)

Hoogenboom, J. Eduard; Martin, William R.; Petrovic, Bojan

2011-01-01

The Monte Carlo performance benchmark for detailed power density calculation in a full-size reactor core is organized under the auspices of the OECD NEA Data Bank. It aims at monitoring over a range of years the increase in performance, measured in terms of standard deviation and computer time, of Monte Carlo calculation of the power density in small volumes. A short description of the reactor geometry and composition is discussed. One of the unique features of the benchmark exercise is the possibility to upload results from participants at a web site of the NEA Data Bank which enables online analysis of results and to graphically display how near we are at the goal of doing a detailed power distribution calculation with acceptable statistical uncertainty in an acceptable computing time. First results are discussed which show that 10 to 100 billion histories must be simulated to reach a standard deviation of a few percent in the estimated power of most of the requested the fuel zones. Even when using a large supercomputer, a considerable speedup is still needed to reach the target of 1 hour computer time. An outlook is given of what to expect from this benchmark exercise over the years. Possible extensions of the benchmark for specific issues relevant in current Monte Carlo calculation for nuclear reactors are also discussed. (author)

Three-Dimensional (X,Y,Z) Deterministic Analysis of the PCA-Replica Neutron Shielding Benchmark Experiment using the TORT-3.2 Code and Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

OpenAIRE

Pescarini Massimo; Orsi Roberto; Frisoni Manuela

2016-01-01

The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with ...

Benchmarking road safety performance: Identifying a meaningful reference (best-in-class).

Science.gov (United States)

Chen, Faan; Wu, Jiaorong; Chen, Xiaohong; Wang, Jianjun; Wang, Di

2016-01-01

For road safety improvement, comparing and benchmarking performance are widely advocated as the emerging and preferred approaches. However, there is currently no universally agreed upon approach for the process of road safety benchmarking, and performing the practice successfully is by no means easy. This is especially true for the two core activities of which: (1) developing a set of road safety performance indicators (SPIs) and combining them into a composite index; and (2) identifying a meaningful reference (best-in-class), one which has already obtained outstanding road safety practices. To this end, a scientific technique that can combine the multi-dimensional safety performance indicators (SPIs) into an overall index, and subsequently can identify the 'best-in-class' is urgently required. In this paper, the Entropy-embedded RSR (Rank-sum ratio), an innovative, scientific and systematic methodology is investigated with the aim of conducting the above two core tasks in an integrative and concise procedure, more specifically in a 'one-stop' way. Using a combination of results from other methods (e.g. the SUNflower approach) and other measures (e.g. Human Development Index) as a relevant reference, a given set of European countries are robustly ranked and grouped into several classes based on the composite Road Safety Index. Within each class the 'best-in-class' is then identified. By benchmarking road safety performance, the results serve to promote best practice, encourage the adoption of successful road safety strategies and measures and, more importantly, inspire the kind of political leadership needed to create a road transport system that maximizes safety. Copyright © 2015 Elsevier Ltd. All rights reserved.

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....

Posture Control—Human-Inspired Approaches for Humanoid Robot Benchmarking: Conceptualizing Tests, Protocols and Analyses

Directory of Open Access Journals (Sweden)

Thomas Mergner

2018-05-01

Full Text Available Posture control is indispensable for both humans and humanoid robots, which becomes especially evident when performing sensorimotor tasks such as moving on compliant terrain or interacting with the environment. Posture control is therefore targeted in recent proposals of robot benchmarking in order to advance their development. This Methods article suggests corresponding robot tests of standing balance, drawing inspirations from the human sensorimotor system and presenting examples from robot experiments. To account for a considerable technical and algorithmic diversity among robots, we focus in our tests on basic posture control mechanisms, which provide humans with an impressive postural versatility and robustness. Specifically, we focus on the mechanically challenging balancing of the whole body above the feet in the sagittal plane around the ankle joints in concert with the upper body balancing around the hip joints. The suggested tests target three key issues of human balancing, which appear equally relevant for humanoid bipeds: (1 four basic physical disturbances (support surface (SS tilt and translation, field and contact forces may affect the balancing in any given degree of freedom (DoF. Targeting these disturbances allows us to abstract from the manifold of possible behavioral tasks. (2 Posture control interacts in a conflict-free way with the control of voluntary movements for undisturbed movement execution, both with “reactive” balancing of external disturbances and “proactive” balancing of self-produced disturbances from the voluntary movements. Our proposals therefore target both types of disturbances and their superposition. (3 Relevant for both versatility and robustness of the control, linkages between the posture control mechanisms across DoFs provide their functional cooperation and coordination at will and on functional demands. The suggested tests therefore include ankle-hip coordination. Suggested benchmarking

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.

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...

Jendl-3.1 iron validation on the PCA-REPLICA (H{sub 2}O/Fe) shielding benchmark experiment

Energy Technology Data Exchange (ETDEWEB)

Pescarini, M.; Borgia, M. G. [ENEA, Centro Ricerche ``Ezio Clementel``, Bologna (Italy). Dipt. Energia

1997-03-01

The PCA-REPLICA (H{sub 2}O/Fe) neutron shielding benchmarks experiment is analysed using the SN 2-D DOT 3.5-E code and the 3-D-equivalent flux synthesis method. This engineering benchmark reproduces the ex-core radial geometry of a PWR, including a mild steel reactor pressure vessel (RPV) simulator, and is designed to test the accuracy of the calculation of the in-vessel neutron exposure parameters. This accuracy is strongly dependent on the quality of the iron neutron cross sections used to describe the nuclear reactions within the RPV simulator. In particular, in this report, the cross sections based on the JENDL-3.1 iron data files are tested, through a comparison of the calculated integral and spectral results with the corresponding experimental data. In addition, the present results are compared, on the same benchmark experiment, with those of a preceding ENEA-Bologna validation of the ENDF/B VI iron cross sections. The integral result comparison indicates that, for all the threshold detectors considered (Rh-103 (n, n`) Rh-103m, In-115 (n, n`) In-115m and S-32 (n, p) P-32), the JENDL-3.1 natural iron data produce satisfactory results similar to those obtained with the ENDF/B VI iron data. On the contrary, when the JENDL/3.1 Fe-56 data file is used, strongly underestimated results are obtained for the lower energy threshold detectors, Rh-103 and In-115. This fact, in particular, becomes more evident with increasing the neutron penetration depth in the RPV simulator.

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...

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...

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

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…

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

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

RETRAN-3D Analysis Of The OECD/NRC Peach Bottom 2 Turbine Trip Benchmark

International Nuclear Information System (INIS)

Barten, W.; Coddington, P.

2003-01-01

This paper presents the PSI results on the different Phases of the Peach Bottom BWR Turbine Trip Benchmark using the RETRAN-3D code. In the first part of the paper, the analysis of Phase 1 is presented, in which the system pressure is predicted based on a pre-defined core power distribution. These calculations demonstrate the importance of accurate modelling of the non-equilibrium effects within the steam separator region. In the second part, a selection of the RETRAN-3D results for Phase 2 are given, where the power is predicted using a 3-D core with pre-defined core flow and pressure boundary conditions. A comparison of calculations using the different (Benchmark-specified) boundary conditions illustrates the sensitivity of the power maximum on the various resultant system parameters. In the third part of the paper, the results of the Phase 3 calculation are presented. This phase, which is a combination of the analytical work of Phases 1 and 2, gives good agreement with the measured data. The coupling of the pressure and flow oscillations in the steam line, the mass balance in the core, the (void) reactivity and the core power are all discussed. It is shown that the reactivity effects resulting from the change in the core void can explain the overall behaviour of the transient prior to the reactor scram. The time-dependent, normalized power for different thermal-hydraulic channels in the core is discussed in some detail. Up to the time of reactor scram, the power change was similar in all channels, with differences of the order of only a few percent. The axial shape of the channel powers at the time of maximum (overall) power increased in the core centre (compared with the shape at time zero). These changes occur as a consequence of the relative change in the channel void, which is largest in the region of the onset of boiling, and the influence on the different fuel assemblies of the complex ring pattern of the control rods. (author)

RETRAN-3D Analysis Of The OECD/NRC Peach Bottom 2 Turbine Trip Benchmark

Energy Technology Data Exchange (ETDEWEB)

Barten, W.; Coddington, P

2003-03-01

This paper presents the PSI results on the different Phases of the Peach Bottom BWR Turbine Trip Benchmark using the RETRAN-3D code. In the first part of the paper, the analysis of Phase 1 is presented, in which the system pressure is predicted based on a pre-defined core power distribution. These calculations demonstrate the importance of accurate modelling of the non-equilibrium effects within the steam separator region. In the second part, a selection of the RETRAN-3D results for Phase 2 are given, where the power is predicted using a 3-D core with pre-defined core flow and pressure boundary conditions. A comparison of calculations using the different (Benchmark-specified) boundary conditions illustrates the sensitivity of the power maximum on the various resultant system parameters. In the third part of the paper, the results of the Phase 3 calculation are presented. This phase, which is a combination of the analytical work of Phases 1 and 2, gives good agreement with the measured data. The coupling of the pressure and flow oscillations in the steam line, the mass balance in the core, the (void) reactivity and the core power are all discussed. It is shown that the reactivity effects resulting from the change in the core void can explain the overall behaviour of the transient prior to the reactor scram. The time-dependent, normalized power for different thermal-hydraulic channels in the core is discussed in some detail. Up to the time of reactor scram, the power change was similar in all channels, with differences of the order of only a few percent. The axial shape of the channel powers at the time of maximum (overall) power increased in the core centre (compared with the shape at time zero). These changes occur as a consequence of the relative change in the channel void, which is largest in the region of the onset of boiling, and the influence on the different fuel assemblies of the complex ring pattern of the control rods. (author)

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...

OECD/NRC BWR Turbine Trip Transient Benchmark as a Basis for Comprehensive Qualification and Studying Best-Estimate Coupled Codes

International Nuclear Information System (INIS)

Ivanov, Kostadin; Olson, Andy; Sartori, Enrico

2004-01-01

An Organisation for Economic Co-operation and Development (OECD)/U.S. Nuclear Regulatory Commission (NRC)-sponsored coupled-code benchmark has been initiated for a boiling water reactor (BWR) turbine trip (TT) transient. Turbine trip transients in a BWR are pressurization events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. In addition, the available real plant experimental data make this benchmark problem very valuable. Over the course of defining and coordinating the BWR TT benchmark, a systematic approach has been established to validate best-estimate coupled codes. This approach employs a multilevel methodology that not only allows for a consistent and comprehensive validation process but also contributes to the study of different numerical and computational aspects of coupled best-estimate simulations. This paper provides an overview of the OECD/NRC BWR TT benchmark activities with emphasis on the discussion of the numerical and computational aspects of the benchmark

Development of a 3-dimensional calculation model of the Danish research reactor DR3 to analyse a proposal to a new core design called ring-core

Energy Technology Data Exchange (ETDEWEB)

Nonboel, E

1985-07-01

A 3-dimensional calculation model of the Danish research reactor DR3 has been developed. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. The model has been tested against measurements with satisfying results. Furthermore the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dummy elements to increase the thermal flux in the center of the reactor. (author)

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.

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)

RB reactor as the U-D2O benchmark criticality system

International Nuclear Information System (INIS)

Pesic, M.

1998-01-01

From a rich and valuable database fro 580 different reactor cores formed up to now in the RB nuclear reactor, a selected and well recorded set is carefully chosen and preliminarily proposed as a new uranium-heavy water benchmark criticality system for validation od reactor design computer codes and data libraries. The first results of validation of the MCNP code and adjoining neutron cross section libraries are resented in this paper. (author)

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...

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

Appraisement and benchmarking of third-party logistic service provider by exploration of risk-based approach

Directory of Open Access Journals (Sweden)

Nitin Kumar Sahu

2015-12-01

Full Text Available In the present era, Reverse Logistics Support has monitored as a momentous realm, where stuffs are transferred from point of consumption to origin. The companies who provide the logistic equipments, i.e. Truck, Joseph Cyril Bomford, and Shipment, etc. to its partner’s firms called Third-Party Logistics (3PL service provider. Today, the feasible 3PL service provider evaluation-opt problem is yet an amorous dilemma. The appraisement and benchmarking of logistics service providers in extent of index; allied risk-based indices and their interrelated metrics; outlooked as a great tool for each international firm, in order that firm could obtain their core goals. The novelty of manuscript is that here, a hairy-based approach has been integrated and then implemented upon a novel developed multi hierarchical third-party logistics (3PL service providers appraisement index in purpose to umpire the 3PL provider for their strong and ill’s core indices. Moreover, the overall score (Si system has also been carried out for benchmarking the 3PL provider companies, where s1 has been found as the best 3PL service provider. The developed approach enabled the manager of firms to make the verdict towards the best inclusive evaluation process of 3PL performance appraisement and benchmarking. A numerical illustration has also been provided to validate the verdict support system.

JNC results of BN-600 benchmark calculation (phase 3)

International Nuclear Information System (INIS)

Ishikawa, M.

2002-01-01

The present work is the result of phase 3 BN-600 core benchmark problem, meaning burnup and heterogeneity. Analytical method applied consisted of: JENDL-3.2 nuclear data library, group constants (70 group, ABBN type self shielding transport factors), heterogeneous cell model for fuel and control rod, basic diffusion calculation (CITATION code), transport theory and mesh size correction (NSHEX code based on SN transport nodal method developed by JNC). Burnup and heterogeneity calculation results are presented obtained by applying both diffusion and transport approach for beginning and end of cycle

Real-time advanced nuclear reactor core model

International Nuclear Information System (INIS)

Koclas, J.; Friedman, F.; Paquette, C.; Vivier, P.

1990-01-01

The paper describes a multi-nodal advanced nuclear reactor core model. The model is based on application of modern equivalence theory to the solution of neutron diffusion equation in real time employing the finite differences method. The use of equivalence theory allows the application of the finite differences method to cores divided into hundreds of nodes, as opposed to the much finer divisions (in the order of ten thousands of nodes) where the unmodified method is currently applied. As a result the model can be used for modelling of the core kinetics for real time full scope training simulators. Results of benchmarks, validate the basic assumptions of the model and its applicability to real-time simulation. (orig./HP)

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.

Proposed biopsy performance benchmarks for MRI based on an audit of a large academic center.

Science.gov (United States)

Sedora Román, Neda I; Mehta, Tejas S; Sharpe, Richard E; Slanetz, Priscilla J; Venkataraman, Shambhavi; Fein-Zachary, Valerie; Dialani, Vandana

2018-05-01

Performance benchmarks exist for mammography (MG); however, performance benchmarks for magnetic resonance imaging (MRI) are not yet fully developed. The purpose of our study was to perform an MRI audit based on established MG and screening MRI benchmarks and to review whether these benchmarks can be applied to an MRI practice. An IRB approved retrospective review of breast MRIs was performed at our center from 1/1/2011 through 12/31/13. For patients with biopsy recommendation, core biopsy and surgical pathology results were reviewed. The data were used to derive mean performance parameter values, including abnormal interpretation rate (AIR), positive predictive value (PPV), cancer detection rate (CDR), percentage of minimal cancers and axillary node negative cancers and compared with MG and screening MRI benchmarks. MRIs were also divided by screening and diagnostic indications to assess for differences in performance benchmarks amongst these two groups. Of the 2455 MRIs performed over 3-years, 1563 were performed for screening indications and 892 for diagnostic indications. With the exception of PPV2 for screening breast MRIs from 2011 to 2013, PPVs were met for our screening and diagnostic populations when compared to the MRI screening benchmarks established by the Breast Imaging Reporting and Data System (BI-RADS) 5 Atlas ® . AIR and CDR were lower for screening indications as compared to diagnostic indications. New MRI screening benchmarks can be used for screening MRI audits while the American College of Radiology (ACR) desirable goals for diagnostic MG can be used for diagnostic MRI audits. Our study corroborates established findings regarding differences in AIR and CDR amongst screening versus diagnostic indications. © 2017 Wiley Periodicals, Inc.

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.

Inelastic finite element analysis of a pipe-elbow assembly (benchmark problem 2)

Energy Technology Data Exchange (ETDEWEB)

Knapp, H P [Internationale Atomreaktorbau GmbH (INTERATOM) Bergisch Gladbach (Germany); Prij, J [Netherlands Energy Research Foundation (ECN) Petten (Netherlands)

1979-06-01

In the scope of the international benchmark problem effort on piping systems, benchmark problem 2 consisting of a pipe elbow assembly, subjected to a time dependent in-plane bending moment, was analysed using the finite element program MARC. Numerical results are presented and a comparison with experimental results is made. It is concluded that the main reason for the deviation between the calculated and measured values is due to the fact that creep-plasticity interaction is not taken into account in the analysis. (author)

Comparisons with measured data of the simulated local core parameters by the coupled code ATHLET-BIPR-VVER applying a new enhanced model of the reactor pressure vessel

International Nuclear Information System (INIS)

Nikonov, S.; Pasichnyk, I.; Velkov, K.; Pautz, A.

2011-01-01

The paper describes the performed comparisons of measured and simulated local core data based on the OECD/NEA Benchmark on Kalinin-3 NPP: 'Switching off of one of the four operating main circulation pumps at nominal reactor power'. The local measurements of in core self-powered neutron detectors (SPND) in 64 fuel assemblies on 7 axial levels are used for the comparisons of the assemblies axial power distributions and the thermocouples readings at 93 fuel assembly heads are applied for the fuel assembly coolant temperature comparisons. The analyses are done on the base of benchmark transient calculations performed with the coupled system code ATHLET/BIPR-VVER. In order to describe more realistically the fluid mixing phenomena in a reactor pressure vessel a new enhanced nodalization scheme is being developed. It could take into account asymmetric flow behaviour in the reactor pressure vessel structures like downcomer, reactor core inlet and outlet, control rods' guided tubes, support grids etc. For this purpose details of the core geometry are modelled. About 58000 control volumes and junctions are applied. Cross connection are used to describe the interaction between the fluid objects. The performed comparisons are of great interest because they show some advantages by performing coupled code production pseudo-3D analysis of NPPs applying the parallel thermo-hydraulic channel methodology (or 1D thermo-hydraulic system code modeling). (Authors)

Development and verification of an efficient spatial neutron kinetics method for reactivity-initiated event analyses

International Nuclear Information System (INIS)

Ikeda, Hideaki; Takeda, Toshikazu

2001-01-01

A space/time nodal diffusion code based on the nodal expansion method (NEM), EPISODE, was developed in order to evaluate transient neutron behavior in light water reactor cores. The present code employs the improved quasistatic (IQS) method for spatial neutron kinetics, and neutron flux distribution is numerically obtained by solving the neutron diffusion equation with the nonlinear iteration scheme to achieve fast computation. A predictor-corrector (PC) method developed in the present study enabled to apply a coarse time mesh to the transient spatial neutron calculation than that applicable in the conventional IQS model, which improved computational efficiency further. Its computational advantage was demonstrated by applying to the numerical benchmark problems that simulate reactivity-initiated events, showing reduction of computational times up to a factor of three than the conventional IQS. The thermohydraulics model was also incorporated in EPISODE, and the capability of realistic reactivity event analyses was verified using the SPERT-III/E-Core experimental data. (author)

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.

OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Third Workshop (V1000-CT3)

International Nuclear Information System (INIS)

2005-01-01

The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. The technical topics presented at this workshop were: Review of the benchmark activities after the 2. Workshop; - Discussion of participant's feedback and introduced modifications

Scientific Drilling of Impact Craters - Well Logging and Core Analyses Using Magnetic Methods (Invited)

Science.gov (United States)

Fucugauchi, J. U.; Perez-Cruz, L. L.; Velasco-Villarreal, M.

2013-12-01

Drilling projects of impact structures provide data on the structure and stratigraphy of target, impact and post-impact lithologies, providing insight on the impact dynamics and cratering. Studies have successfully included magnetic well logging and analyses in core and cuttings, directed to characterize the subsurface stratigraphy and structure at depth. There are 170-180 impact craters documented in the terrestrial record, which is a small proportion compared to expectations derived from what is observed on the Moon, Mars and other bodies of the solar system. Knowledge of the internal 3-D deep structure of craters, critical for understanding impacts and crater formation, can best be studied by geophysics and drilling. On Earth, few craters have yet been investigated by drilling. Craters have been drilled as part of industry surveys and/or academic projects, including notably Chicxulub, Sudbury, Ries, Vredefort, Manson and many other craters. As part of the Continental ICDP program, drilling projects have been conducted on the Chicxulub, Bosumtwi, Chesapeake, Ries and El gygytgyn craters. Inclusion of continuous core recovery expanded the range of paleomagnetic and rock magnetic applications, with direct core laboratory measurements, which are part of the tools available in the ocean and continental drilling programs. Drilling studies are here briefly reviewed, with emphasis on the Chicxulub crater formed by an asteroid impact 66 Ma ago at the Cretaceous/Paleogene boundary. Chicxulub crater has no surface expression, covered by a kilometer of Cenozoic sediments, thus making drilling an essential tool. As part of our studies we have drilled eleven wells with continuous core recovery. Magnetic susceptibility logging, magnetostratigraphic, rock magnetic and fabric studies have been carried out and results used for lateral correlation, dating, formation evaluation, azimuthal core orientation and physical property contrasts. Contributions of magnetic studies on impact

Performance of Artificial Intelligence Workloads on the Intel Core 2 Duo Series Desktop Processors

OpenAIRE

Abdul Kareem PARCHUR; Kuppangari Krishna RAO; Fazal NOORBASHA; Ram Asaray SINGH

2010-01-01

As the processor architecture becomes more advanced, Intel introduced its Intel Core 2 Duo series processors. Performance impact on Intel Core 2 Duo processors are analyzed using SPEC CPU INT 2006 performance numbers. This paper studied the behavior of Artificial Intelligence (AI) benchmarks on Intel Core 2 Duo series processors. Moreover, we estimated the task completion time (TCT) @1 GHz, @2 GHz and @3 GHz Intel Core 2 Duo series processors frequency. Our results show the performance scalab...

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)

Performance Against WELCOA's Worksite Health Promotion Benchmarks Across Years Among Selected US Organizations.

Science.gov (United States)

Weaver, GracieLee M; Mendenhall, Brandon N; Hunnicutt, David; Picarella, Ryan; Leffelman, Brittanie; Perko, Michael; Bibeau, Daniel L

2018-05-01

The purpose of this study was to quantify the performance of organizations' worksite health promotion (WHP) activities against the benchmarking criteria included in the Well Workplace Checklist (WWC). The Wellness Council of America (WELCOA) developed a tool to assess WHP with its 100-item WWC, which represents WELCOA's 7 performance benchmarks. Workplaces. This study includes a convenience sample of organizations who completed the checklist from 2008 to 2015. The sample size was 4643 entries from US organizations. The WWC includes demographic questions, general questions about WHP programs, and scales to measure the performance against the WELCOA 7 benchmarks. Descriptive analyses of WWC items were completed separately for each year of the study period. The majority of the organizations represented each year were multisite, multishift, medium- to large-sized companies mostly in the services industry. Despite yearly changes in participating organizations, results across the WELCOA 7 benchmark scores were consistent year to year. Across all years, benchmarks that organizations performed the lowest were senior-level support, data collection, and programming; wellness teams and supportive environments were the highest scoring benchmarks. In an era marked with economic swings and health-care reform, it appears that organizations are staying consistent in their performance across these benchmarks. The WWC could be useful for organizations, practitioners, and researchers in assessing the quality of WHP programs.

Accelerating the SCE-UA Global Optimization Method Based on Multi-Core CPU and Many-Core GPU

Directory of Open Access Journals (Sweden)

Guangyuan Kan

2016-01-01

Full Text Available The famous global optimization SCE-UA method, which has been widely used in the field of environmental model parameter calibration, is an effective and robust method. However, the SCE-UA method has a high computational load which prohibits the application of SCE-UA to high dimensional and complex problems. In recent years, the hardware of computer, such as multi-core CPUs and many-core GPUs, improves significantly. These much more powerful new hardware and their software ecosystems provide an opportunity to accelerate the SCE-UA method. In this paper, we proposed two parallel SCE-UA methods and implemented them on Intel multi-core CPU and NVIDIA many-core GPU by OpenMP and CUDA Fortran, respectively. The Griewank benchmark function was adopted in this paper to test and compare the performances of the serial and parallel SCE-UA methods. According to the results of the comparison, some useful advises were given to direct how to properly use the parallel SCE-UA methods.

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

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...

Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 5. TMI-1 Benchmark Performed by Different Coupled Three-Dimensional Neutronics Thermal- Hydraulic Codes

International Nuclear Information System (INIS)

D'Auria, F.; Galassi, G.M.; Spadoni, A.; Gago, J.L.; Grgic, D.

2001-01-01

A comprehensive analysis of a double-ended main-steam-line-break (MSLB) accident assumed to have occurred in the Babcock and Wilcox Three Mile Island (TMI) Unit 1 nuclear power plant (NPP) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy. The research has been carried out in cooperation with the University of Zagreb, Croatia, and with partial financial support from the European Union through a grant to one of the authors. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development Committee on the Safety of Nuclear Installations-Nuclear Science Committee PWR MSLB Benchmark. Different code versions have been adopted in the analysis. Results from the following codes (or code versions) are described in this paper: 1. RELAP5/mod 3.2.2, gamma version, coupled with the three-dimensional (3-D) neutron kinetics PARCS code; 2. RELAP5/mod 3.2.2, gamma version, coupled with the 3-D neutron kinetics QUABBOX code; 3. RELAP5/3D code coupled with the 3-D neutron kinetics NESTLE code. Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by The Pennsylvania State University in cooperation with GPU Nuclear (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission (NRC). The main challenge for the calculation was the prediction of the return to power (RTP) following the inlet of cold water into the core and one 'stuck-withdrawn' control rod. Non-realistic assumptions were proposed to augment the core power peak following scram. Zero-dimensional neutronics codes were capable of detecting the RTP after scram. However, the application of 3-D neutronics codes to the same scenario allowed the calculation of a similar value for overall core power peak but showed power increase occurrence in about one-tenth of the core volume. The results achieved in phase 1 of

Benchmarking, benchmarks, or best practices? Applying quality improvement principles to decrease surgical turnaround time.

Science.gov (United States)

Mitchell, L

1996-01-01

The processes of benchmarking, benchmark data comparative analysis, and study of best practices are distinctly different. The study of best practices is explained with an example based on the Arthur Andersen & Co. 1992 "Study of Best Practices in Ambulatory Surgery". The results of a national best practices study in ambulatory surgery were used to provide our quality improvement team with the goal of improving the turnaround time between surgical cases. The team used a seven-step quality improvement problem-solving process to improve the surgical turnaround time. The national benchmark for turnaround times between surgical cases in 1992 was 13.5 minutes. The initial turnaround time at St. Joseph's Medical Center was 19.9 minutes. After the team implemented solutions, the time was reduced to an average of 16.3 minutes, an 18% improvement. Cost-benefit analysis showed a potential enhanced revenue of approximately $300,000, or a potential savings of $10,119. Applying quality improvement principles to benchmarking, benchmarks, or best practices can improve process performance. Understanding which form of benchmarking the institution wishes to embark on will help focus a team and use appropriate resources. Communicating with professional organizations that have experience in benchmarking will save time and money and help achieve the desired results.

US/JAERI calculational benchmarks for nuclear data and codes intercomparison. Article 8

International Nuclear Information System (INIS)

Youssef, M.Z.; Jung, J.; Sawan, M.E.; Nakagawa, M.; Mori, T.; Kosako, K.

1986-01-01

Prior to analyzing the integral experiments performed at the FNS facility at JAERI, both US and JAERI's analysts have agreed upon four calculational benchmark problems proposed by JAERI to intercompare results based on various codes and data base used independently by both countries. To compare codes the same data base is used (ENDF/B-IV). To compare nuclear data libraries, common codes were applied. Some of the benchmarks chosen were geometrically simple and consisted of a single material to clearly identify sources of discrepancies and thus help in analysing the integral experiments

Coupled Tort-TD/CTF Capability for high-fidelity LWR core calculations - 321

International Nuclear Information System (INIS)

Christienne, M.; Avramova, M.; Perin, Y.; Seubert, A.

2010-01-01

This paper describes the developed coupling scheme between TORT-TD and CTF. TORT-TD is a time-dependent 3D discrete ordinates neutron transport code. TORT-TD is utilized for high-fidelity reactor core neutronics calculations while CTF is providing the thermal-hydraulics feedback information. CTF is an improved version of the advanced thermal-hydraulic sub-channel code COBRA-TF, which is widely used for best-estimate evaluations of LWR safety margins. CTF is a transient code based on a separated flow representation of the two-phase flow. The coupled code TORT-TD/CTF allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. Steady-state and transient test cases, based on the OECD/NRC PWR MOX/UO 2 Core Transient Benchmark, have been calculated. The steady state cases are based on a quarter core model while the transient test case models a control rod ejection transient in a small PWR mini-core fuel assembly arrangement. The obtained results with TORT-TD/CTF are verified by a code-to-code comparison with the previously developed NEM/CTF and TORT-TD/ATHLET coupled code systems. The performed comparative analysis indicates the applicability and high-fidelity potential of the TORT-TD/CTF coupling. (authors)

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...

Structural code benchmarking for the analysis of impact response of nuclear material shipping casks

International Nuclear Information System (INIS)

Glass, R.E.

1984-01-01

The Transportation Technology Center at Sandia National Laboratories has initiated a program to benchmark thermal and structural codes that are available to the nuclear material transportation community. The program consists of the following five phrases: (1) code inventory and review, (2) development of a cask-like set of problems, (3) multiple independent numerical analyses of the problems, (4) transfer of information, and (5) performance of experiments to obtain data for comparison with the numerical analyses. This paper will summarize the results obtained by the independent numerical analyses. The analyses indicate the variability that can be expected both due to differences in user-controlled parameters and from code-to-code differences. The results show that in purely elastic analyses, differences can be attributed to user controlled parameters. Model problems involving elastic/plastic material behavior and large deformations, however, have greater variability with significant differences reported for implicit and explicit integration schemes in finite element programs. This variability demonstrates the need to obtain experimental data to properly benchmark codes utilizing elastic/plastic material models and large deformation capability

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.

High Energy Physics (HEP) benchmark program

International Nuclear Information System (INIS)

Yasu, Yoshiji; Ichii, Shingo; Yashiro, Shigeo; Hirayama, Hideo; Kokufuda, Akihiro; Suzuki, Eishin.

1993-01-01

High Energy Physics (HEP) benchmark programs are indispensable tools to select suitable computer for HEP application system. Industry standard benchmark programs can not be used for this kind of particular selection. The CERN and the SSC benchmark suite are famous HEP benchmark programs for this purpose. The CERN suite includes event reconstruction and event generator programs, while the SSC one includes event generators. In this paper, we found that the results from these two suites are not consistent. And, the result from the industry benchmark does not agree with either of these two. Besides, we describe comparison of benchmark results using EGS4 Monte Carlo simulation program with ones from two HEP benchmark suites. Then, we found that the result from EGS4 in not consistent with the two ones. The industry standard of SPECmark values on various computer systems are not consistent with the EGS4 results either. Because of these inconsistencies, we point out the necessity of a standardization of HEP benchmark suites. Also, EGS4 benchmark suite should be developed for users of applications such as medical science, nuclear power plant, nuclear physics and high energy physics. (author)

HEXTRAN-SMABRE calculation of the 6th AER Benchmark, main steam line break in a WWER-440 NPP

International Nuclear Information System (INIS)

Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

2003-01-01

The sixth AER benchmark is the second AER benchmark for couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a double end break of one main steam line in a WWER-440 plant. The core is at the end of its first cycle in full power conditions. In VTT HEXTRAN2.9 is used for the core kinetics and dynamics and SMABRE4.8 as a thermal hydraulic model for the primary and secondary loop. The plant model for SMABRE consists mainly of two input models, Loviisa model and a standard WWER-440/213 plant model. The primary loop includes six separate loops, the pressure vessel is divided into six parallel channels in SMABRE and the whole core calculation is performed in the core with HEXTRAN. The horizontal steam generators are modelled with heat transfer tubes in five levels and vertically with two parts, riser and downcomer. With this kind of detailed modelling of steam generators there occurs strong flashing after break opening. As a sequence of the main steam line break at nominal power level, the reactor trip is followed quite soon. The liquid temperature continues to decrease in one core inlet sector which may lead to recriticality and neuron power increase. The situation is very sensitive to small changes in the steam generator and break flow modelling and therefore several sensitivity calculations have been done. Also two stucked control rods have been assumed. Due to boric acid concentration in the high pressure safety injection subcriticality is finally guaranteed in the transient (Authors)

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…

OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Fourth Workshop (V100-CT4)

International Nuclear Information System (INIS)

2006-01-01

The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. Since the previous coupled code benchmarks indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and

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...

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

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)

A study on Monte Carlo analysis of Pebble-type VHTR core for hydrogen production

International Nuclear Information System (INIS)

Kim, Hong Chul

2005-02-01

In order to pursue exact the core analysis for VHTR core which will be developed in future, a study on Monte Carol method was carried out. In Korea, pebble and prism type core are under investigation for VHTR core analysis. In this study, pebble-type core was investigated because it was known that it should not only maintain the nuclear fuel integrity but also have the advantage in economical efficiency and safety. The pebble-bed cores of HTR-PROTEUS critical facility in Swiss were selected for the benchmark model. After the detailed MCNP modeling of the whole facility, calculations of nuclear characteristics were performed. The two core configurations, Core 4.3 and Core 5 (reference state no. 3), among the 10 configurations of the HTR-PROTEUS cores were chosen to be analyzed in order to treat different fuel loading pattern and modeled. The former is a random packing core and the latter deterministic packing core. Based on the experimental data and the benchmark result of other research groups for the two different cores, some nuclear characteristics were calculated. Firstly, keff was calculated for these cores. The effect for TRIO homogeneity model was investigated. Control rod and shutdown rod worths also were calculated and the sensitivity analysis on cross-section library and reflector thickness was pursued. Lastly, neutron flux profiles were investigated in reflector regions. It is noted that Monte Carlo analysis of pebble-type VHTR core was firstly carried out in Korea. Also, this study should not only provide the basic data for pebble-type VHTR core analysis for hydrogen production but also be utilized as the verified data to validate a computer code for VHTR core analysis which will be developed in future

Argonne Code Center: Benchmark problem book.

Energy Technology Data Exchange (ETDEWEB)

None, None

1977-06-01

This book is an outgrowth of activities of the Computational Benchmark Problems Committee of the Mathematics and Computation Division of the American Nuclear Society. This is the second supplement of the original benchmark book which was first published in February, 1968 and contained computational benchmark problems in four different areas. Supplement No. 1, which was published in December, 1972, contained corrections to the original benchmark book plus additional problems in three new areas. The current supplement. Supplement No. 2, contains problems in eight additional new areas. The objectives of computational benchmark work and the procedures used by the committee in pursuing the objectives are outlined in the original edition of the benchmark book (ANL-7416, February, 1968). The members of the committee who have made contributions to Supplement No. 2 are listed below followed by the contributors to the earlier editions of the benchmark book.

Random geometry capability in RMC code for explicit analysis of polytype particle/pebble and applications to HTR-10 benchmark

International Nuclear Information System (INIS)

Liu, Shichang; Li, Zeguang; Wang, Kan; Cheng, Quan; She, Ding

2018-01-01

Highlights: •A new random geometry was developed in RMC for mixed and polytype particle/pebble. •This capability was applied to the full core calculations of HTR-10 benchmark. •Reactivity, temperature coefficient and control rod worth of HTR-10 were compared. •This method can explicitly model different packing fraction of different pebbles. •Monte Carlo code with this method can simulate polytype particle/pebble type reactor. -- Abstract: With the increasing demands of high fidelity neutronics analysis and the development of computer technology, Monte Carlo method is becoming more and more attractive in accurate simulation of pebble bed High Temperature gas-cooled Reactor (HTR), owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. For the double-heterogeneous geometry of pebble bed, traditional Monte Carlo codes can treat it by explicit geometry description. However, packing methods such as Random Sequential Addition (RSA) can only produce a sphere packing up to 38% volume packing fraction, while Discrete Element Method (DEM) is troublesome and also time consuming. Moreover, traditional Monte Carlo codes are difficult and inconvenient to simulate the mixed and polytype particles or pebbles. A new random geometry method was developed in Monte Carlo code RMC to simulate the particle transport in polytype particle/pebble in double heterogeneous geometry systems. This method was verified by some test cases, and applied to the full core calculations of HTR-10 benchmark. The reactivity, temperature coefficient and control rod worth of HTR-10 were compared for full core and initial core in helium and air atmosphere respectively, and the results agree well with the benchmark results and experimental results. This work would provide an efficient tool for the innovative design of pebble bed, prism HTRs and molten salt reactors with polytype particles or pebbles using Monte Carlo method.

Validation of Westinghouse integrated code POLCA-T against OECD NEACRP-L-335 rod ejection benchmark

Energy Technology Data Exchange (ETDEWEB)

Panayotov, Dobromir [Westinghouse Electric Sweden AB, Vaesteraas, SE-721 63 (Sweden)

2008-07-01

This paper describes the work performed and results obtained in the validation of the POLCA-T code against NEACRP PWR Rod Ejection Transients Benchmark. Presented work is a part of the POLCA-T licensing Assessment Data Base for BWR Control Rod Drop Accident (CRDA) Application. The validation against a PWR Rod Ejection Accidents (REA) Benchmark is relevant for the validation of the code for BWR CRDA, as the analyses of both transients require identical phenomena to be modelled. All six benchmark cases have been analyzed in the presented work. Initial state steady-state calculations including boron search, control rod worth, and final state power search have been performed by POLCA7 code. Initial state boron adjustment and steady-state CR worth as well as the transient analyses were performed by POLCA-T code. Benchmark results including 3D transient power distributions are compared with reference PANTHER solutions and published results of other codes. Given the similarity of the kinetics modelling for a BWR CRDA and a PWR REA and the fact that POLCA-T accurately predicts the local transient power and thus, the resulting fuel enthalpy, it is concluded that POLCA-T is a state-of-art tool also for BWR CRDA analysis. (author)

Validation of Westinghouse integrated code POLCA-T against OECD NEACRP-L-335 rod ejection benchmark

International Nuclear Information System (INIS)

Panayotov, Dobromir

2008-01-01

This paper describes the work performed and results obtained in the validation of the POLCA-T code against NEACRP PWR Rod Ejection Transients Benchmark. Presented work is a part of the POLCA-T licensing Assessment Data Base for BWR Control Rod Drop Accident (CRDA) Application. The validation against a PWR Rod Ejection Accidents (REA) Benchmark is relevant for the validation of the code for BWR CRDA, as the analyses of both transients require identical phenomena to be modelled. All six benchmark cases have been analyzed in the presented work. Initial state steady-state calculations including boron search, control rod worth, and final state power search have been performed by POLCA7 code. Initial state boron adjustment and steady-state CR worth as well as the transient analyses were performed by POLCA-T code. Benchmark results including 3D transient power distributions are compared with reference PANTHER solutions and published results of other codes. Given the similarity of the kinetics modelling for a BWR CRDA and a PWR REA and the fact that POLCA-T accurately predicts the local transient power and thus, the resulting fuel enthalpy, it is concluded that POLCA-T is a state-of-art tool also for BWR CRDA analysis. (author)

The coupled code system DORT-TD/THERMIX and its application to the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

International Nuclear Information System (INIS)

Pautz, A.; Tyobeka, B.; Ivanov, K.

2009-01-01

In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)

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.

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.

Comparison of different LMFBR primary containment codes applied to a Benchmark problem

International Nuclear Information System (INIS)

Benuzzi, A.

1986-01-01

The Cont Benchmark calculation exercise is a project sponsored by the Containment Loading and Response Group, a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee - CEC. A full-size typical Pool type LMFBR undergoing a postulated Core Disruptive Accident (CDA) has been defined by Belgonucleaire-Brussels under a study contract financed by the CEC and has been submitted to seven containment code calculations. The results of these calculations are presented and discussed in this paper

featsel: A framework for benchmarking of feature selection algorithms and cost functions

OpenAIRE

Marcelo S. Reis; Gustavo Estrela; Carlos Eduardo Ferreira; Junior Barrera

2017-01-01

In this paper, we introduce featsel, a framework for benchmarking of feature selection algorithms and cost functions. This framework allows the user to deal with the search space as a Boolean lattice and has its core coded in C++ for computational efficiency purposes. Moreover, featsel includes Perl scripts to add new algorithms and/or cost functions, generate random instances, plot graphs and organize results into tables. Besides, this framework already comes with dozens of algorithms and co...

ANN-Benchmarks: A Benchmarking Tool for Approximate Nearest Neighbor Algorithms

DEFF Research Database (Denmark)

Aumüller, Martin; Bernhardsson, Erik; Faithfull, Alexander

2017-01-01

This paper describes ANN-Benchmarks, a tool for evaluating the performance of in-memory approximate nearest neighbor algorithms. It provides a standard interface for measuring the performance and quality achieved by nearest neighbor algorithms on different standard data sets. It supports several...... visualise these as images, Open image in new window plots, and websites with interactive plots. ANN-Benchmarks aims to provide a constantly updated overview of the current state of the art of k-NN algorithms. In the short term, this overview allows users to choose the correct k-NN algorithm and parameters...... for their similarity search task; in the longer term, algorithm designers will be able to use this overview to test and refine automatic parameter tuning. The paper gives an overview of the system, evaluates the results of the benchmark, and points out directions for future work. Interestingly, very different...

Depollution benchmarks for capacitors, batteries and printed wiring boards from waste electrical and electronic equipment (WEEE)

International Nuclear Information System (INIS)

Savi, Daniel; Kasser, Ueli; Ott, Thomas

2013-01-01

Highlights: • We’ve analysed data on the dismantling of electronic and electrical appliances. • Ten years of mass balance data of more than recycling companies have been considered. • Percentages of dismantled batteries, capacitors and PWB have been studied. • Threshold values and benchmarks for batteries and capacitors have been identified. • No benchmark for the dismantling of printed wiring boards should be set. - Abstract: The article compiles and analyses sample data for toxic components removed from waste electronic and electrical equipment (WEEE) from more than 30 recycling companies in Switzerland over the past ten years. According to European and Swiss legislation, toxic components like batteries, capacitors and printed wiring boards have to be removed from WEEE. The control bodies of the Swiss take back schemes have been monitoring the activities of WEEE recyclers in Switzerland for about 15 years. All recyclers have to provide annual mass balance data for every year of operation. From this data, percentage shares of removed batteries and capacitors are calculated in relation to the amount of each respective WEEE category treated. A rationale is developed, why such an indicator should not be calculated for printed wiring boards. The distributions of these de-pollution indicators are analysed and their suitability for defining lower threshold values and benchmarks for the depollution of WEEE is discussed. Recommendations for benchmarks and threshold values for the removal of capacitors and batteries are given

Depollution benchmarks for capacitors, batteries and printed wiring boards from waste electrical and electronic equipment (WEEE)

Energy Technology Data Exchange (ETDEWEB)

Savi, Daniel, E-mail: d.savi@umweltchemie.ch [Dipl. Environmental Sci. ETH, büro für umweltchemie, Zurich (Switzerland); Kasser, Ueli [Lic. Phil. Nat. (Chemist), büro für umweltchemie, Zurich (Switzerland); Ott, Thomas [Dipl. Phys. ETH, Institute of Applied Simulation, Zurich University of Applied Sciences, Wädenswil (Switzerland)

2013-12-15

Highlights: • We’ve analysed data on the dismantling of electronic and electrical appliances. • Ten years of mass balance data of more than recycling companies have been considered. • Percentages of dismantled batteries, capacitors and PWB have been studied. • Threshold values and benchmarks for batteries and capacitors have been identified. • No benchmark for the dismantling of printed wiring boards should be set. - Abstract: The article compiles and analyses sample data for toxic components removed from waste electronic and electrical equipment (WEEE) from more than 30 recycling companies in Switzerland over the past ten years. According to European and Swiss legislation, toxic components like batteries, capacitors and printed wiring boards have to be removed from WEEE. The control bodies of the Swiss take back schemes have been monitoring the activities of WEEE recyclers in Switzerland for about 15 years. All recyclers have to provide annual mass balance data for every year of operation. From this data, percentage shares of removed batteries and capacitors are calculated in relation to the amount of each respective WEEE category treated. A rationale is developed, why such an indicator should not be calculated for printed wiring boards. The distributions of these de-pollution indicators are analysed and their suitability for defining lower threshold values and benchmarks for the depollution of WEEE is discussed. Recommendations for benchmarks and threshold values for the removal of capacitors and batteries are given.

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

Designing a Supply Chain Management Academic Curriculum Using QFD and Benchmarking

Science.gov (United States)

Gonzalez, Marvin E.; Quesada, Gioconda; Gourdin, Kent; Hartley, Mark

2008-01-01

Purpose: The purpose of this paper is to utilize quality function deployment (QFD), Benchmarking analyses and other innovative quality tools to develop a new customer-centered undergraduate curriculum in supply chain management (SCM). Design/methodology/approach: The researchers used potential employers as the source for data collection. Then,…

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

2011-03-01

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

Energy Technology Data Exchange (ETDEWEB)

Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Maddock, Thomas L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Ning [Idaho National Lab. (INL), Idaho Falls, ID (United States); Phillips, Ann Marie [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schreck, Kenneth A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bolin, John M. [General Atomics, San Diego, CA (United States); Veca, Anthony [General Atomics, San Diego, CA (United States); McKnight, Richard D. [Argonne National Lab. (ANL), Argonne, IL (United States); Lell, Richard M. [Argonne National Lab. (ANL), Argonne, IL (United States)

2014-03-01

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

Energy Technology Data Exchange (ETDEWEB)

Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

2014-10-01

Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

International Nuclear Information System (INIS)

Lazaro, A.; Schikorr, M.; Mikityuk, K.; Ammirabile, L.; Bandini, G.; Darmet, G.; Schmitt, D.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D.; Stempniewicz, M.

2014-01-01

Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs

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...

ENDF/B VI iron validation onpca-replica (H2O/FE) shielding benchmark experiment

Energy Technology Data Exchange (ETDEWEB)

Pescarini, M. [ENEA, Bologna (Italy). Centro Ricerche Energia `E. Clementel` - Area Energia e Innovazione

1994-05-01

The PCA-REPLICA (H2O/Fe) neutron shielding benchmark experiment is analysed using the SN 2-D DOT 3.5 code and the 3-D-equivalent flux synthesis method. This engineering benchmark reproduces the ex-core radial geometry of a PWR, including a mild steel reactor pressure vessel (RPV) simulator, and is dsigned to test the accuracy of the calculation of the in-vessel neutron exposure parameters (fast fluence and iron displacement rates). This accuracy is strongly dependent on the quality of the iron neutron cross section used to describe the nuclear reactions within the RPV simulator. In particular, in this report, the cross sections based on the ENDF/B VI iron data files are tested, through a comparison of the calculated integral and spectral results with the corresponding experimental data. In addition, the present results are compared, on the same benchmark experiment, with those of a preceding ENEA (Italian Agency for Energy, New Technologies and Environment)-Bologna validation of the JEF-2.1 iron cross sections. The integral result comparison indicates that, for all the thresold detectors considered (Rh-103 (n,n) Rh-103m, In-115 (n,n) In-115 (n,n) In-115m and S-32 (n.p) P-32), the ENDF/B VI iron data produce better results than the JEF-2.1 iron data. In particular, in the ENDF/B VI calcultaions, an improvement of the in-vessel C/E (Calculated/Experimental) activity ratios for the lower energy threshold detectors, Rh-103 and In-115, is observed. This improvement becomes more evident with increasing neutron penetration depth in the vessel. This is probably attributable to the fact that the inelastic scattering cross section values of the ENDF/B VI Fe-56 data file, approximately in the 0.86 - 1.5 MeV energy range, are lower then the corresponding values of the JEF-2.1 data file.

Preliminary results of the seventh three-dimensional AER dynamic benchmark problem calculation. Solution with DYN3D and RELAP5-3D codes

International Nuclear Information System (INIS)

Bencik, M.; Hadek, J.

2011-01-01

The paper gives a brief survey of the seventh three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at Nuclear Research Institute Rez. This benchmark was defined at the twentieth AER Symposium in Hanassari (Finland). It is focused on investigation of transient behaviour in a WWER-440 nuclear power plant. Its initiating event is opening of the main isolation valve and re-connection of the loop with its main circulation pump in operation. The WWER-440 plant is at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations were performed with the code DYN3D. Transient calculation was made with the system code RELAP5-3D. The two-group homogenized cross sections library HELGD05 created by HELIOS code was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP Dukovany was adopted for the seventh AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was coupled with 49 core thermal-hydraulic channels and 8 reflector channels connected with the three-dimensional model of the reactor vessel. The detailed nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5-3D system code and a short description of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters. (Authors)

Qualification of the core model DYN3D coupled with the code ATHLET as an advanced tool for the accident analysis of VVER type reactors. Pt. 2. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Rohde, U.

2002-10-01

Benchmark calculations for the validation of the coupled neutron kinetics/thermohydraulic code complex DYN3D-ATHLET are described. Two benchmark problems concerning hypothetical accident scenarios with leaks in the steam system for a VVER-440 type reactor and the TMI-1 PWR have been solved. The first benchmark task has been defined by FZR in the frame of the international association 'Atomic Energy Research' (AER), the second exercise has been organized under the auspices of the OECD. While in the first benchmark the break of the main steam collector in the sub-critical hot zero power state of the reactor was considered, the break of one of the two main steam lines at full reactor power was assumed in the OECD benchmark. Therefore, in this exercise the mixing of the coolant from the intact and the defect loops had to be considered, while in the AER benchmark the steam collector break causes a homogeneous overcooling of the primary circuit. In the AER benchmark, each participant had to use its own macroscopic cross section libraries. In the OECD benchmark, the cross sections were given in the benchmark definition. The main task of both benchmark problems was to analyse the re-criticality of the scrammed reactor due to the overcooling. For both benchmark problems, a good agreement of the DYN3D-ATHLET solution with the results of other codes was achieved. Differences in the time of re-criticality and the height of the power peak between various solutions of the AER benchmark can be explained by the use of different cross section data. Significant differences in the thermohydraulic parameters (coolant temperature, pressure) occurred only at the late stage of the transient during the emergency injection of highly borated water. In the OECD benchmark, a broader scattering of the thermohydraulic results can be observed, while a good agreement between the various 3D reactor core calculations with given thermohydraulic boundary conditions was achieved. Reasons for the

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.

Preliminary Results for the OECD/NEA Time Dependent Benchmark using Rattlesnake, Rattlesnake-IQS and TDKENO

Energy Technology Data Exchange (ETDEWEB)

DeHart, Mark D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mausolff, Zander [Univ. of Florida, Gainesville, FL (United States); Weems, Zach [Univ. of Florida, Gainesville, FL (United States); Popp, Dustin [Univ. of Florida, Gainesville, FL (United States); Smith, Kristin [Univ. of Florida, Gainesville, FL (United States); Shriver, Forrest [Univ. of Florida, Gainesville, FL (United States); Goluoglu, Sedat [Univ. of Florida, Gainesville, FL (United States); Prince, Zachary [Texas A & M Univ., College Station, TX (United States); Ragusa, Jean [Texas A & M Univ., College Station, TX (United States)

2016-08-01

One goal of the MAMMOTH M&S project is to validate the analysis capabilities within MAMMOTH. Historical data has shown limited value for validation of full three-dimensional (3D) multi-physics methods. Initial analysis considered the TREAT startup minimum critical core and one of the startup transient tests. At present, validation is focusing on measurements taken during the M8CAL test calibration series. These exercises will valuable in preliminary assessment of the ability of MAMMOTH to perform coupled multi-physics calculations; calculations performed to date are being used to validate the neutron transport solver Rattlesnake\\cite{Rattlesnake} and the fuels performance code BISON. Other validation projects outside of TREAT are available for single-physics benchmarking. Because the transient solution capability of Rattlesnake is one of the key attributes that makes it unique for TREAT transient simulations, validation of the transient solution of Rattlesnake using other time dependent kinetics benchmarks has considerable value. The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has recently developed a computational benchmark for transient simulations. This benchmark considered both two-dimensional (2D) and 3D configurations for a total number of 26 different transients. All are negative reactivity insertions, typically returning to the critical state after some time.

Preliminary Results for the OECD/NEA Time Dependent Benchmark using Rattlesnake, Rattlesnake-IQS and TDKENO

International Nuclear Information System (INIS)

DeHart, Mark D.; Mausolff, Zander; Weems, Zach; Popp, Dustin; Smith, Kristin; Shriver, Forrest; Goluoglu, Sedat; Prince, Zachary; Ragusa, Jean

2016-01-01

One goal of the MAMMOTH M&S project is to validate the analysis capabilities within MAMMOTH. Historical data has shown limited value for validation of full three-dimensional (3D) multi-physics methods. Initial analysis considered the TREAT startup minimum critical core and one of the startup transient tests. At present, validation is focusing on measurements taken during the M8CAL test calibration series. These exercises will valuable in preliminary assessment of the ability of MAMMOTH to perform coupled multi-physics calculations; calculations performed to date are being used to validate the neutron transport solver Rattlesnake\\citelesnake) and the fuels performance code BISON. Other validation projects outside of TREAT are available for single-physics benchmarking. Because the transient solution capability of Rattlesnake is one of the key attributes that makes it unique for TREAT transient simulations, validation of the transient solution of Rattlesnake using other time dependent kinetics benchmarks has considerable value. The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has recently developed a computational benchmark for transient simulations. This benchmark considered both two-dimensional (2D) and 3D configurations for a total number of 26 different transients. All are negative reactivity insertions, typically returning to the critical state after some time.

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

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

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...

A Benchmark for Virtual Camera Control

DEFF Research Database (Denmark)

Burelli, Paolo; Yannakakis, Georgios N.

2015-01-01

Automatically animating and placing the virtual camera in a dynamic environment is a challenging task. The camera is expected to maximise and maintain a set of properties — i.e. visual composition — while smoothly moving through the environment and avoiding obstacles. A large number of different....... For this reason, in this paper, we propose a benchmark for the problem of virtual camera control and we analyse a number of different problems in different virtual environments. Each of these scenarios is described through a set of complexity measures and, as a result of this analysis, a subset of scenarios...

LEU fuel cycle analyses for the Belgian BR2 Research Reactor

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.

1988-01-01

Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the 235 U loading 20% and the fuel meat volume 51%. The first LEU design used 10 B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 ± 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs

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....

Benchmarking – A tool for judgment or improvement?

DEFF Research Database (Denmark)

Rasmussen, Grane Mikael Gregaard

2010-01-01

perceptions of benchmarking will be presented; public benchmarking and best practice benchmarking. These two types of benchmarking are used to characterize and discuss the Danish benchmarking system and to enhance which effects, possibilities and challenges that follow in the wake of using this kind......Change in construction is high on the agenda for the Danish government and a comprehensive effort is done in improving quality and efficiency. This has led to an initiated governmental effort in bringing benchmarking into the Danish construction sector. This paper is an appraisal of benchmarking...... as it is presently carried out in the Danish construction sector. Many different perceptions of benchmarking and the nature of the construction sector, lead to an uncertainty in how to perceive and use benchmarking, hence, generating an uncertainty in understanding the effects of benchmarking. This paper addresses...

BONFIRE: benchmarking computers and computer networks

OpenAIRE

Bouckaert, Stefan; Vanhie-Van Gerwen, Jono; Moerman, Ingrid; Phillips, Stephen; Wilander, Jerker

2011-01-01

The benchmarking concept is not new in the field of computing or computer networking. With “benchmarking tools”, one usually refers to a program or set of programs, used to evaluate the performance of a solution under certain reference conditions, relative to the performance of another solution. Since the 1970s, benchmarking techniques have been used to measure the performance of computers and computer networks. Benchmarking of applications and virtual machines in an Infrastructure-as-a-Servi...

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)

The Isprs Benchmark on Indoor Modelling

Science.gov (United States)

Khoshelham, K.; Díaz Vilariño, L.; Peter, M.; Kang, Z.; Acharya, D.

2017-09-01

Automated generation of 3D indoor models from point cloud data has been a topic of intensive research in recent years. While results on various datasets have been reported in literature, a comparison of the performance of different methods has not been possible due to the lack of benchmark datasets and a common evaluation framework. The ISPRS benchmark on indoor modelling aims to address this issue by providing a public benchmark dataset and an evaluation framework for performance comparison of indoor modelling methods. In this paper, we present the benchmark dataset comprising several point clouds of indoor environments captured by different sensors. We also discuss the evaluation and comparison of indoor modelling methods based on manually created reference models and appropriate quality evaluation criteria. The benchmark dataset is available for download at: html"target="_blank">http://www2.isprs.org/commissions/comm4/wg5/benchmark-on-indoor-modelling.html.

Core story creation: analysing narratives to construct stories for learning.

Science.gov (United States)

Petty, Julia; Jarvis, Joy; Thomas, Rebecca

2018-03-16

Educational research uses narrative enquiry to gain and interpret people's experiences. Narrative analysis is used to organise and make sense of acquired narrative. 'Core story creation' is a way of managing raw data obtained from narrative interviews to construct stories for learning. To explain how core story creation can be used to construct stories from raw narratives obtained by interviewing parents about their neonatal experiences and then use these stories to educate learners. Core story creation involves reconfiguration of raw narratives. Reconfiguration includes listening to and rereading transcribed narratives, identifying elements of 'emplotment' and reordering these to form a constructed story. Thematic analysis is then performed on the story to draw out learning themes informed by the participants. Core story creation using emplotment is a strategy of narrative reconfiguration that produces stories which can be used to develop resources relating to person-centred education about the patient experience. Stories constructed from raw narratives in the context of constructivism can provide a medium or an 'end product' for use in learning resource development. This can then contribute to educating students or health professionals about patients' experiences. ©2018 RCN Publishing Company Ltd. All rights reserved. Not to be copied, transmitted or recorded in any way, in whole or part, without prior permission of the publishers.

Benchmarking the energy efficiency of commercial buildings

International Nuclear Information System (INIS)

Chung, William; Hui, Y.V.; Lam, Y. Miu

2006-01-01

Benchmarking energy-efficiency is an important tool to promote the efficient use of energy in commercial buildings. Benchmarking models are mostly constructed in a simple benchmark table (percentile table) of energy use, which is normalized with floor area and temperature. This paper describes a benchmarking process for energy efficiency by means of multiple regression analysis, where the relationship between energy-use intensities (EUIs) and the explanatory factors (e.g., operating hours) is developed. Using the resulting regression model, these EUIs are then normalized by removing the effect of deviance in the significant explanatory factors. The empirical cumulative distribution of the normalized EUI gives a benchmark table (or percentile table of EUI) for benchmarking an observed EUI. The advantage of this approach is that the benchmark table represents a normalized distribution of EUI, taking into account all the significant explanatory factors that affect energy consumption. An application to supermarkets is presented to illustrate the development and the use of the benchmarking method

Numisheet2005 Benchmark Analysis on Forming of an Automotive Underbody Cross Member: Benchmark 2

International Nuclear Information System (INIS)

Buranathiti, Thaweepat; Cao Jian

2005-01-01

This report presents an international cooperation benchmark effort focusing on simulations of a sheet metal stamping process. A forming process of an automotive underbody cross member using steel and aluminum blanks is used as a benchmark. Simulation predictions from each submission are analyzed via comparison with the experimental results. A brief summary of various models submitted for this benchmark study is discussed. Prediction accuracy of each parameter of interest is discussed through the evaluation of cumulative errors from each submission

Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations. Interim report; Weiterentwicklung moderner Verfahren zu Neutronentransport und Unsicherheitsanalysen fuer Kernberechnungen. Zwischenbericht

Energy Technology Data Exchange (ETDEWEB)

Zwermann, Winfried; Aures, Alexander; Bostelmann, Friederike; Pasichnyk, Ihor; Perin, Yann; Velkov, Kiril; Zilly, Matias

2016-12-15

This report documents the status of the research and development goals reached within the reactor safety research project RS1536 ''Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations'' as of the 3{sup rd} quarter of 2016. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts, in particular fast reactors cooled by liquid metal. The contributing individual goals are the further optimization and validation of deterministic calculation methods with high spatial and energy resolution, the development of a coupled calculation system using the Monte Carlo method for the neutron transport to describe time-dependent reactor core states, the processing and validation of nuclear data, particularly with regard to covariance data, the development, validation, and application of sampling-based methods for uncertainty and sensitivity analyses, the creation of a platform for performing systematic uncertainty analyses for fast reactor systems, as well as the description of states of severe core damage with the Monte Carlo method. Moreover, work regarding the European NURESAFE project, started in the preceding project RS1503, are being continued and completed.

SKaMPI: A Comprehensive Benchmark for Public Benchmarking of MPI

Directory of Open Access Journals (Sweden)

Ralf Reussner

2002-01-01

Full Text Available The main objective of the MPI communication library is to enable portable parallel programming with high performance within the message-passing paradigm. Since the MPI standard has no associated performance model, and makes no performance guarantees, comprehensive, detailed and accurate performance figures for different hardware platforms and MPI implementations are important for the application programmer, both for understanding and possibly improving the behavior of a given program on a given platform, as well as for assuring a degree of predictable behavior when switching to another hardware platform and/or MPI implementation. We term this latter goal performance portability, and address the problem of attaining performance portability by benchmarking. We describe the SKaMPI benchmark which covers a large fraction of MPI, and incorporates well-accepted mechanisms for ensuring accuracy and reliability. SKaMPI is distinguished among other MPI benchmarks by an effort to maintain a public performance database with performance data from different hardware platforms and MPI implementations.

Neutronics benchmark of a MOX assembly with near-weapons-grade plutonium

International Nuclear Information System (INIS)

Difilippo, F.C.; Fisher, S.E.

1998-01-01

One of the proposed ways to dispose of surplus weapons-grade plutonium (Pu) is to irradiate the high-fissile material in light-water reactors in order to reduce the Pu enrichment to the level of spent fuels from commercial reactors. Considerable experience has been accumulated about the behavior of mixed-oxide (MOX) uranium and plutonium fuels for plutonium recycling in commercial reactors, but the experience is related to Pu enrichments typical of spent fuels quite below the values of weapons-grade plutonium. Important decisions related to the kind of reactors to be used for the disposition of the plutonium are going to be based on calculations, so the validation of computational algorithms related to all aspects of the fuel cycle (power distributions, isotopics as function of the burnup, etc.), for weapons-grade isotopics is very important. Analysis of public domain data reveals that the cycle-2 irradiation in the Quad cities boiling-water reactor (BWR) is the most recent US destructive examination. This effort involved the irradiation of five MOX assemblies using 80 and 90% fissile plutonium. These benchmark data were gathered by General Electric under the sponsorship of the Electric Power Research Institute. It is emphasized, however, that global parameters are not the focus of this benchmark, since the five bundles containing MOX fuels did not significantly affect the overall core performance. However, since the primary objective of this work is to compare against measured post-irradiation assembly data, the term benchmark is applied here. One important reason for performing the benchmark on Quad Cities irradiation is that the fissile blends (up to 90%) are higher than reactor-grade and, quite close to, weapons-grade isotopics

Experimental Benchmarking of Fire Modeling Simulations. Final Report

International Nuclear Information System (INIS)

Greiner, Miles; Lopez, Carlos

2003-01-01

A series of large-scale fire tests were performed at Sandia National Laboratories to simulate a nuclear waste transport package under severe accident conditions. The test data were used to benchmark and adjust the Container Analysis Fire Environment (CAFE) computer code. CAFE is a computational fluid dynamics fire model that accurately calculates the heat transfer from a large fire to a massive engulfed transport package. CAFE will be used in transport package design studies and risk analyses

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

Out-of-Core Computations of High-Resolution Level Sets by Means of Code Transformation

DEFF Research Database (Denmark)

Christensen, Brian Bunch; Nielsen, Michael Bang; Museth, Ken

2012-01-01

We propose a storage efficient, fast and parallelizable out-of-core framework for streaming computations of high resolution level sets. The fundamental techniques are skewing and tiling transformations of streamed level set computations which allow for the combination of interface propagation, re...... computations are now CPU bound and consequently the overall performance is unaffected by disk latency and bandwidth limitations. We demonstrate this with several benchmark tests that show sustained out-of-core throughputs close to that of in-core level set simulations....

BENCHMARKING ORTEC ISOTOPIC MEASUREMENTS AND CALCULATIONS

Energy Technology Data Exchange (ETDEWEB)

Dewberry, R; Raymond Sigg, R; Vito Casella, V; Nitin Bhatt, N

2008-09-29

This report represents a description of compiled benchmark tests conducted to probe and to demonstrate the extensive utility of the Ortec ISOTOPIC {gamma}-ray analysis computer program. The ISOTOPIC program performs analyses of {gamma}-ray spectra applied to specific acquisition configurations in order to apply finite-geometry correction factors and sample-matrix-container photon absorption correction factors. The analysis program provides an extensive set of preset acquisition configurations to which the user can add relevant parameters in order to build the geometry and absorption correction factors that the program determines from calculus and from nuclear g-ray absorption and scatter data. The Analytical Development Section field nuclear measurement group of the Savannah River National Laboratory uses the Ortec ISOTOPIC analysis program extensively for analyses of solid waste and process holdup applied to passive {gamma}-ray acquisitions. Frequently the results of these {gamma}-ray acquisitions and analyses are to determine compliance with facility criticality safety guidelines. Another use of results is to designate 55-gallon drum solid waste as qualified TRU waste3 or as low-level waste. Other examples of the application of the ISOTOPIC analysis technique to passive {gamma}-ray acquisitions include analyses of standard waste box items and unique solid waste configurations. In many passive {gamma}-ray acquisition circumstances the container and sample have sufficient density that the calculated energy-dependent transmission correction factors have intrinsic uncertainties in the range 15%-100%. This is frequently the case when assaying 55-gallon drums of solid waste with masses of up to 400 kg and when assaying solid waste in extensive unique containers. Often an accurate assay of the transuranic content of these containers is not required, but rather a good defensible designation as >100 nCi/g (TRU waste) or <100 nCi/g (low level solid waste) is required. In

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)

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)

Electricity consumption in school buildings - benchmark and web tools; Elforbrug i skoler - benchmark og webvaerktoej

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-07-01

The aim of this project has been to produce benchmarks for electricity consumption in Danish schools in order to encourage electricity conservation. An internet programme has been developed with the aim of facilitating schools' access to benchmarks and to evaluate energy consumption. The overall purpose is to create increased attention to the electricity consumption of each separate school by publishing benchmarks which take the schools' age and number of pupils as well as after school activities into account. Benchmarks can be used to make green accounts and work as markers in e.g. energy conservation campaigns, energy management and for educational purposes. The internet tool can be found on www.energiguiden.dk. (BA)

Evaluation of tight-pitch PWR cores

International Nuclear Information System (INIS)

Correa, F.; Driscoll, M.J.; Lanning, D.D.

1979-08-01

The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235 U/UO 2 : Pu/ThO 2 : 233 U/ThO 2 - and the conventional recycle-mode uranium system - 235 U/UO 2 : Pu/UO 2 . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

BEAVRS full core burnup calculation in hot full power condition by RMC code

International Nuclear Information System (INIS)

Liu, Shichang; Liang, Jingang; Wu, Qu; Guo, JuanJuan; Huang, Shanfang; Tang, Xiao; Li, Zeguang; Wang, Kan

2017-01-01

Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.

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...

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...

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.

Forward-Weighted CADIS Method for Variance Reduction of Monte Carlo Reactor Analyses

International Nuclear Information System (INIS)

Wagner, John C.; Mosher, Scott W.

2010-01-01

Current state-of-the-art tools and methods used to perform 'real' commercial reactor analyses use high-fidelity transport codes to produce few-group parameters at the assembly level for use in low-order methods applied at the core level. Monte Carlo (MC) methods, which allow detailed and accurate modeling of the full geometry and energy details and are considered the 'gold standard' for radiation transport solutions, are playing an ever-increasing role in correcting and/or verifying the several-decade-old methodology used in current practice. However, the prohibitive computational requirements associated with obtaining fully converged system-wide solutions restrict the role of MC to benchmarking deterministic results at a limited number of state-points for a limited number of relevant quantities. A goal of current research at Oak Ridge National Laboratory (ORNL) is to change this paradigm by enabling the direct use of MC for full-core reactor analyses. The most significant of the many technical challenges that must be overcome is the slow non-uniform convergence of system-wide MC estimates and the memory requirements associated with detailed solutions throughout a reactor (problems involving hundreds of millions of different material and tally regions due to fuel irradiation, temperature distributions, and the needs associated with multi-physics code coupling). To address these challenges, research has focused on development in the following two areas: (1) a hybrid deterministic/MC method for determining high-precision fluxes throughout the problem space in k-eigenvalue problems and (2) an efficient MC domain-decomposition algorithm that partitions the problem phase space onto multiple processors for massively parallel systems, with statistical uncertainty estimation. The focus of this paper is limited to the first area mentioned above. It describes the FW-CADIS method applied to variance reduction of MC reactor analyses and provides initial results for calculating

A Seafloor Benchmark for 3-dimensional Geodesy

Science.gov (United States)

Chadwell, C. D.; Webb, S. C.; Nooner, S. L.

2014-12-01

We have developed an inexpensive, permanent seafloor benchmark to increase the longevity of seafloor geodetic measurements. The benchmark provides a physical tie to the sea floor lasting for decades (perhaps longer) on which geodetic sensors can be repeatedly placed and removed with millimeter resolution. Global coordinates estimated with seafloor geodetic techniques will remain attached to the benchmark allowing for the interchange of sensors as they fail or become obsolete, or for the sensors to be removed and used elsewhere, all the while maintaining a coherent series of positions referenced to the benchmark. The benchmark has been designed to free fall from the sea surface with transponders attached. The transponder can be recalled via an acoustic command sent from the surface to release from the benchmark and freely float to the sea surface for recovery. The duration of the sensor attachment to the benchmark will last from a few days to a few years depending on the specific needs of the experiment. The recovered sensors are then available to be reused at other locations, or again at the same site in the future. Three pins on the sensor frame mate precisely and unambiguously with three grooves on the benchmark. To reoccupy a benchmark a Remotely Operated Vehicle (ROV) uses its manipulator arm to place the sensor pins into the benchmark grooves. In June 2014 we deployed four benchmarks offshore central Oregon. We used the ROV Jason to successfully demonstrate the removal and replacement of packages onto the benchmark. We will show the benchmark design and its operational capabilities. Presently models of megathrust slip within the Cascadia Subduction Zone (CSZ) are mostly constrained by the sub-aerial GPS vectors from the Plate Boundary Observatory, a part of Earthscope. More long-lived seafloor geodetic measures are needed to better understand the earthquake and tsunami risk associated with a large rupture of the thrust fault within the Cascadia subduction zone

The Late Pliocene Eltanin Impact - Documentation From Sediment Core Analyses

Science.gov (United States)

Gersonde, R.; Kuhn, G.; Kyte, F. T.; Flores, J.; Becquey, S.

2002-12-01

The expeditions ANT-XII/4 (1995) and ANT-XVIII/5a (2001) of the RV POLARSTERN collected extensive bathymetric and seismic data sets as well as sediment cores from an area in the Bellingshausen Sea (eastern Pacific Southern Ocean) that allow the first comprehensive geoscientific documentation of an asteroid impact into a deep ocean (~ 5 km) basin, named the Eltanin impact. Impact deposits have now been recovered from a total of more than 20 sediment cores collected in an area covering about 80,000 km2. Combined biomagnetostratigraphic dating places the impact event into the earliest Matuyama Chron, a period of enhanced climate variability. Sediment texture analyses and studies of sediment composition including grain size and microfossil distribution reveal the pattern of impact-related sediment disturbance and the sedimentary processes immediately following the impact event. The pattern is complicated by the San Martin Seamounts (~57.5 S, 91 W), a large topographic elevation that rises up to 3000 m above the surrounding abyssal plain in the area affected by the Eltanin impact. The impact ripped up sediments as old as Eocene and probably Paleocene that have been redeposited in a chaotic assemblage. This is followed by a sequence sedimented from a turbulent flow at the sea floor, overprinted by fall-out of airborne meteoritic ejecta that settled trough the water column. Grain size distribution reveals the timing and interaction of the different sedimentary processes. The gathered estimate of ejecta mass deposited over the studied area, composed of shock-melted asteroidal matrial and unmelted meteorites including fragments up to 2.5 cm in diameter, point to an Eltanin asteroid larger than the 1 km in diameter size originally suggested as a minimum based on the ANT-XII/4 results. This places the energy released by the impact at the threshold of those considered to cause environmental disturbance at a global scale and it makes the impact a likely transport mechanism

The Concepts "Benchmarks and Benchmarking" Used in Education Planning: Teacher Education as Example

Science.gov (United States)

Steyn, H. J.

2015-01-01

Planning in education is a structured activity that includes several phases and steps that take into account several kinds of information (Steyn, Steyn, De Waal & Wolhuter, 2002: 146). One of the sets of information that are usually considered is the (so-called) "benchmarks" and "benchmarking" regarding the focus of a…

Final results of the fifth three-dimensional dynamic Atomic Energy Research benchmark problem calculations

International Nuclear Information System (INIS)

Hadek, J.

1999-01-01

The paper gives a brief survey of the fifth three-dimensional dynamic Atomic Energy Research benchmark calculation results received with the code DYN3D/ATHLET at NRI Rez. This benchmark was defined at the seventh Atomic Energy Research Symposium (Hoernitz near Zittau, 1997). Its initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and hot shutdown conditions with one stuck out control rod group. The calculations were performed with the externally coupled codes ATHLET Mod.1.1 Cycle C and DYN3DH1.1/M3. The standard WWER-440/213 input deck of ATHLET code was adopted for benchmark purposes and for coupling with the code DYN3D. The first part of paper contains a brief characteristics of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters. In comparison with the results published at the eighth Atomic Energy Research Symposium (Bystrice nad Pernstejnem, 1998), the results published in this paper are based on improved ATHLET descriptions of control and safety systems. (Author)

Aerodynamic Benchmarking of the Deepwind Design

DEFF Research Database (Denmark)

Bedona, Gabriele; Schmidt Paulsen, Uwe; Aagaard Madsen, Helge

2015-01-01

The aerodynamic benchmarking for the DeepWind rotor is conducted comparing different rotor geometries and solutions and keeping the comparison as fair as possible. The objective for the benchmarking is to find the most suitable configuration in order to maximize the power production and minimize...... the blade solicitation and the cost of energy. Different parameters are considered for the benchmarking study. The DeepWind blade is characterized by a shape similar to the Troposkien geometry but asymmetric between the top and bottom parts: this shape is considered as a fixed parameter in the benchmarking...

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

Energy Technology Data Exchange (ETDEWEB)

Lell, R. M.; Schaefer, R. W.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Nuclear Engineering Division; Inst. of Physics and Power Engineering

2007-10-01

Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of

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)

Lesson learned from the SARNET wall condensation benchmarks

International Nuclear Information System (INIS)

Ambrosini, W.; Forgione, N.; Merli, F.; Oriolo, F.; Paci, S.; Kljenak, I.; Kostka, P.; Vyskocil, L.; Travis, J.R.; Lehmkuhl, J.; Kelm, S.; Chin, Y.-S.; Bucci, M.

2014-01-01

Highlights: • The results of the benchmarking activity on wall condensation are reported. • The work was performed in the frame of SARNET. • General modelling techniques for condensation are discussed. • Results of University of Pisa and of other benchmark participants are discussed. • The lesson learned is drawn. - Abstract: The prediction of condensation in the presence of noncondensable gases has received continuing attention in the frame of the Severe Accident Research Network of Excellence, both in the first (2004–2008) and in the second (2009–2013) EC integrated projects. Among the different reasons for considering so relevant this basic phenomenon, coped with by classical treatments dated in the first decades of the last century, there is the interest for developing updated CFD models for reactor containment analysis, requiring validating at a different level the available modelling techniques. In the frame of SARNET, benchmarking activities were undertaken taking advantage of the work performed at different institutions in setting up and developing models for steam condensation in conditions of interest for nuclear reactor containment. Four steps were performed in the activity, involving: (1) an idealized problem freely inspired at the actual conditions occurring in an experimental facility, CONAN, installed at the University of Pisa; (2) a first comparison with experimental data purposely collected by the CONAN facility; (3) a second comparison with data available from experimental campaigns performed in the same apparatus before the inclusion of the activities in SARNET; (4) a third exercise involving data obtained at lower mixture velocity than in previous campaigns, aimed at providing conditions closer to those addressed in reactor containment analyses. The last step of the benchmarking activity required to change the configuration of the experimental apparatus to achieve the lower flow rates involved in the new test specifications. The

Experimental investigations and seismic analyses for benchmark study of 1000 MW WWER type (water-cooled and moderated reactor) nuclear power plant Kozloduy. Final report 15 June 1993 - 14 June 1994

Energy Technology Data Exchange (ETDEWEB)

Sachansky, S [Building Research Institute (NISI), Sofia (Bulgaria)

1995-07-01

This report includes preparation and compilation of all existing studies related to seismic safety assessment of Kozloduy WWER-1000, i.e. Units 5 and 6; description of previous full scale testing of Unit 5; and the results obtained from seismic analyses performed under benchmark experimental studies. The results are concerned with analysis of the geological conditions; analysis of the seismic wave velocities in the soil layers; analysis of the predominant natural periods; dynamic characteristics of the Unit 5; soil-structure interaction and laboratory testing and analysis of the reactor containment tenders.

Experimental investigations and seismic analyses for benchmark study of 1000 MW WWER type (water-cooled and moderated reactor) nuclear power plant Kozloduy. Final report 15 June 1993 - 14 June 1994

International Nuclear Information System (INIS)

Sachansky, S.

1995-01-01

This report includes preparation and compilation of all existing studies related to seismic safety assessment of Kozloduy WWER-1000, i.e. Units 5 and 6; description of previous full scale testing of Unit 5; and the results obtained from seismic analyses performed under benchmark experimental studies. The results are concerned with analysis of the geological conditions; analysis of the seismic wave velocities in the soil layers; analysis of the predominant natural periods; dynamic characteristics of the Unit 5; soil-structure interaction and laboratory testing and analysis of the reactor containment tenders

Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X and NJOY2016

Directory of Open Access Journals (Sweden)

Kabach Ouadie

2017-12-01

Full Text Available To validate the new Evaluated Nuclear Data File (ENDF/B-VIII.0β4 library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016. The results obtained with the ENDF/B-VIII.0β4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X code. All the MCNP(X calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

Medical school benchmarking - from tools to programmes.

Science.gov (United States)

Wilkinson, Tim J; Hudson, Judith N; Mccoll, Geoffrey J; Hu, Wendy C Y; Jolly, Brian C; Schuwirth, Lambert W T

2015-02-01

Benchmarking among medical schools is essential, but may result in unwanted effects. To apply a conceptual framework to selected benchmarking activities of medical schools. We present an analogy between the effects of assessment on student learning and the effects of benchmarking on medical school educational activities. A framework by which benchmarking can be evaluated was developed and applied to key current benchmarking activities in Australia and New Zealand. The analogy generated a conceptual framework that tested five questions to be considered in relation to benchmarking: what is the purpose? what are the attributes of value? what are the best tools to assess the attributes of value? what happens to the results? and, what is the likely "institutional impact" of the results? If the activities were compared against a blueprint of desirable medical graduate outcomes, notable omissions would emerge. Medical schools should benchmark their performance on a range of educational activities to ensure quality improvement and to assure stakeholders that standards are being met. Although benchmarking potentially has positive benefits, it could also result in perverse incentives with unforeseen and detrimental effects on learning if it is undertaken using only a few selected assessment tools.

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.

Post test analyses of Revisa benchmark based on a creep test at 1100 Celsius degrees performed on a notched tube

International Nuclear Information System (INIS)

Fischer, M.; Bernard, A.; Bhandari, S.

2001-01-01

In the Euratom 4. Framework Program of the European Commission, REVISA Project deals with the Reactor Vessel Integrity under Severe Accidents. One of the tasks consists in the experimental validation of the models developed in the project. To do this, a benchmark was designed where the participants use their models to test the results against an experiment. The experiment called RUPTHER 15 was conducted by the coordinating organisation, CEA (Commissariat a l'Energie Atomique) in France. It is a 'delayed fracture' test on a notched tube. Thermal loading is an axial gradient with a temperature of about 1130 C in the mid-part. Internal pressure is maintained at 0.8 MPa. This paper presents the results of Finite Element calculations performed by Framatome-ANP using the SYSTUS code. Two types of analyses were made: -) one based on the 'time hardening' Norton-Bailey creep law, -) the other based on the coupled creep/damage Lemaitre-Chaboche model. The purpose of this paper is in particular to show the influence of temperature on the simulation results. At high temperatures of the kind dealt with here, slight errors in the temperature measurements can lead to very large differences in the deformation behaviour. (authors)

Benchmarking clinical photography services in the NHS.