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

Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

International Nuclear Information System (INIS)

Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

2009-01-01

In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

Benchmark Calculations on Halden IFA-650 LOCA Test Results

International Nuclear Information System (INIS)

Ek, Mirkka; Kekkonen, Laura; Kelppe, Seppo; Stengaard, J.O.; Josek, Radomir; Wiesenack, Wolfgang; Aounallah, Yacine; Wallin, Hannu; Grandjean, Claude; Herb, Joachim; Lerchl, Georg; Trambauer, Klaus; Sonnenburg, Heinz-Guenther; Nakajima, Tetsuo; Spykman, Gerold; Struzik, Christine

2010-01-01

The assessment of the consequences of a loss-of-coolant accident (LOCA) is to a large extent based on calculations carried out with codes especially developed for addressing the phenomena occurring during the transient. Since the time of the first LOCA experiments, which were largely conducted with fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular the move to high burnup have not only generated a need to re-examine the LOCA safety criteria and to verify their continued validity, but also to confirm that codes show an appropriate performance especially with respect to high burnup phenomena influencing LOCA fuel behaviour. As part of international efforts, the OECD Halden Reactor Project program implemented a test series to address particular LOCA issues. Based on recommendations of a group of experts from the US NRC, EPRI, EDF, FRAMATOME-ANP and GNF, the primary objective of the experiments were defined as 1. Measure the extent of fuel (fragment) relocation into the ballooned region and evaluate its possible effect on cladding temperature and oxidation. 2. Investigate the extent (if any) of 'secondary transient hydriding' on the inner side of the cladding above and below the burst region. The Halden LOCA series, using high burnup fuel segments, contains test cases well suited for checking the ability of LOCA analysis codes to predict or reproduce the measurements and to provide clues as to where the codes need to be improved. The NEA Working Group on Fuel Safety, WGFS, therefore decided to conduct a code benchmark based on the Halden LOCA test series. Emphasis was on the codes' ability to predict or reproduce the thermal and mechanical response of fuel and cladding. Before starting the benchmark, participants were given the opportunity to tune their codes to the experimental system applied in the Halden LOCA tests. To this end, the data from the two commissioning runs were made available. The first of these runs went

Standard Guide for Benchmark Testing of Light Water Reactor Calculations

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

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)

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

Monte Carlo benchmark calculations for 400MWTH PBMR core

International Nuclear Information System (INIS)

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

2007-01-01

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

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.

A review on the benchmarking concept in Malaysian construction safety performance

Science.gov (United States)

Ishak, Nurfadzillah; Azizan, Muhammad Azizi

2018-02-01

Construction industry is one of the major industries that propels Malaysia's economy in highly contributes to our nation's GDP growth, yet the high fatality rates on construction sites have caused concern among safety practitioners and the stakeholders. Hence, there is a need of benchmarking in performance of Malaysia's construction industry especially in terms of safety. This concept can create a fertile ground for ideas, but only in a receptive environment, organization that share good practices and compare their safety performance against other benefit most to establish improvement in safety culture. This research was conducted to study the awareness important, evaluate current practice and improvement, and also identify the constraint in implement of benchmarking on safety performance in our industry. Additionally, interviews with construction professionals were come out with different views on this concept. Comparison has been done to show the different understanding of benchmarking approach and how safety performance can be benchmarked. But, it's viewed as one mission, which to evaluate objectives identified through benchmarking that will improve the organization's safety performance. Finally, the expected result from this research is to help Malaysia's construction industry implement best practice in safety performance management through the concept of benchmarking.

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.)

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

Benchmark calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.E.; Cheng, E.T.

1985-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li/sub 17/Pb/sub 83/ and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the TBR to group structure and weighting spectrum increases and Li enrichment decrease with up to 20% discrepancies for thin natural Li/sub 17/Pb/sub 83/ blankets

Classification of criticality calculations with correlation coefficient method and its application to OECD/NEA burnup credit benchmarks phase III-A and II-A

International Nuclear Information System (INIS)

Okuno, Hiroshi

2003-01-01

A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the Nuclear Energy Agency of the Organisation for Economic Cooperation and Development (OECD/NEA). Phase III-A benchmark was a series of criticality calculations for irradiated Boiling Water Reactor (BWR) fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated Pressurized Water Reactor (PWR) fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark calculation results were classified according to the criterion that the values of the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of benchmark calculation results belonged to the same group, one calculation result was found predictable from the other. An example was shown for each of the Benchmarks. While the evaluated nuclear data seemed the main factor for the classification, further investigations were required for finding other factors. (author)

Benchmark calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.L.; Cheng, E.T.

1986-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li 17 Pb 83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li 17 Pb 83 blankets. (author)

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)

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

Benchmark calculation for GT-MHR using HELIOS/MASTER code package and MCNP

International Nuclear Information System (INIS)

Lee, Kyung Hoon; Kim, Kang Seog; Noh, Jae Man; Song, Jae Seung; Zee, Sung Quun

2005-01-01

The latest research associated with the very high temperature gas-cooled reactor (VHTR) is focused on the verification of a system performance and safety under operating conditions for the VHTRs. As a part of those, an international gas-cooled reactor program initiated by IAEA is going on. The key objectives of this program are the validation of analytical computer codes and the evaluation of benchmark models for the projected and actual VHTRs. New reactor physics analysis procedure for the prismatic VHTR is under development by adopting the conventional two-step procedure. In this procedure, a few group constants are generated through the transport lattice calculations using the HELIOS code, and the core physics analysis is performed by the 3-dimensional nodal diffusion code MASTER. We evaluated the performance of the HELIOS/MASTER code package through the benchmark calculations related to the GT-MHR (Gas Turbine-Modular Helium Reactor) to dispose weapon plutonium. In parallel, MCNP is employed as a reference code to verify the results of the HELIOS/MASTER procedure

International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

2014-12-15

Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4® neutron gamma coupled calculations

International Nuclear Information System (INIS)

Lee, Yi-Kang

2016-01-01

Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be

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

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)

Benchmark density functional theory calculations for nanoscale conductance

DEFF Research Database (Denmark)

Strange, Mikkel; Bækgaard, Iben Sig Buur; Thygesen, Kristian Sommer

2008-01-01

We present a set of benchmark calculations for the Kohn-Sham elastic transmission function of five representative single-molecule junctions. The transmission functions are calculated using two different density functional theory methods, namely an ultrasoft pseudopotential plane-wave code...

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)

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-PROTEUS benchmark calculations. Pt. 1. Unit cell results LEUPRO-1 and LEUPRO-2

International Nuclear Information System (INIS)

Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Klippel, H.T.; Kuijper, J.C.

1995-09-01

In the framework of the IAEA Co-ordinated Research Programme (CRP) on 'Validation of Safety Related Physics Calculations for Low-Enriched (LEU) HTGRs' calculational benchmarks are performed on the basis of LEU-HTR pebble-bed critical experiments carried out in the PROTEUS facility at PSI, Switzerland. Of special interest is the treatment of the double heterogeneity of the fuel and the spherical fuel elements of these pebble bed core configurations. Also of interest is the proper calculation of the safety related physics parameters like the effect of water ingress and control rod worth. This document describes the ECN results of the LEUPRO-1 and LEUPRO-2 unitcell calculations performed with the codes WIMS-E, SCALE-4 and MCNP4A. Results of the LEUPRO-1 unit cell with 20% water ingress in the void is also reported for both the single and the double heterogeneous case. Emphasis is put on the intercomparison of the results obtained by the deterministic codes WIMS-E and SCALE-4, and the Monte Carlo code MCNP4A. The LEUPRO whole core calculations will be reported later. (orig.)

Benchmark neutron porosity log calculations

International Nuclear Information System (INIS)

Little, R.C.; Michael, M.; Verghese, K.; Gardner, R.P.

1989-01-01

Calculations have been made for a benchmark neutron porosity log problem with the general purpose Monte Carlo code MCNP and the specific purpose Monte Carlo code McDNL. For accuracy and timing comparison purposes the CRAY XMP and MicroVax II computers have been used with these codes. The CRAY has been used for an analog version of the MCNP code while the MicroVax II has been used for the optimized variance reduction versions of both codes. Results indicate that the two codes give the same results within calculated standard deviations. Comparisons are given and discussed for accuracy (precision) and computation times for the two codes

Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4{sup ®} neutron gamma coupled calculations

Energy Technology Data Exchange (ETDEWEB)

Lee, Yi-Kang, E-mail: yi-kang.lee@cea.fr

2016-11-01

Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4{sup ®} Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries

Reactor calculation benchmark PCA blind test results

International Nuclear Information System (INIS)

Kam, F.B.K.; Stallmann, F.W.

1980-01-01

Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables

Reactor calculation benchmark PCA blind test results

Energy Technology Data Exchange (ETDEWEB)

Kam, F.B.K.; Stallmann, F.W.

1980-01-01

Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables.

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

HEU benchmark calculations and LEU preliminary calculations for IRR-1

International Nuclear Information System (INIS)

Caner, M.; Shapira, M.; Bettan, M.; Nagler, A.; Gilat, J.

2004-01-01

We performed neutronics calculations for the Soreq Research Reactor, IRR-1. The calculations were done for the purpose of upgrading and benchmarking our codes and methods. The codes used were mainly WIMS-D/4 for cell calculations and the three dimensional diffusion code CITATION for full core calculations. The experimental flux was obtained by gold wire activation methods and compared with our calculated flux profile. The IRR-1 is loaded with highly enriched uranium fuel assemblies, of the plate type. In the framework of preparation for conversion to low enrichment fuel, additional calculations were done assuming the presence of LEU fresh fuel. In these preliminary calculations we investigated the effect on the criticality and flux distributions of the increase of U-238 loading, and the corresponding uranium density.(author)

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

Benchmark calculation of subchannel analysis codes

International Nuclear Information System (INIS)

1996-02-01

In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)

Benchmarking of LOFT LRTS-COBRA-FRAP safety analysis model

International Nuclear Information System (INIS)

Hanson, G.H.; Atkinson, S.A.; Wadkins, R.P.

1982-05-01

The purpose of this work was to check out the LOFT LRTS/COBRA-IV/FRAP-T5 safety-analysis models against test data obtained during a LOFT operational transient in which there was a power and fuel-temperature rise. LOFT Experiment L6-3 was an excessive-load-increase anticipated transient test in which the main steam-flow-control valve was driven from its operational position to full-open in seven seconds. The resulting cooldown and reactivity-increase transients provide a good benchmark for the reactivity-and-power-prediction capability of the LRTS calculations, and for the fuel-bundle and fuel-rod temperature-response analysis capability of the LOFT COBRA-IV and FRAP-T5 models

Benchmarking road safety performance by grouping local territories : a study in The Netherlands.

NARCIS (Netherlands)

Aarts, L.T. & Houwing, S.

2015-01-01

The method of benchmarking provides an opportunity to learn from better performing territories to improve the effectiveness and efficiency of activities in a particular field of interest. Such a field of interest could be road safety. Road safety benchmarking can include several indicators, ranging

Benchmark calculation of SCALE-PC 4.3 CSAS6 module and burnup credit criticality analysis

Energy Technology Data Exchange (ETDEWEB)

Shin, Hee Sung; Ro, Seong Gy; Shin, Young Joon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-12-01

Calculation biases of SCALE-PC CSAS6 module for PWR spent fuel, metallized spent fuel and solution of nuclear materials have been determined on the basis of the benchmark to be 0.01100, 0.02650 and 0.00997, respectively. With the aid of the code system, nuclear criticality safety analysis for the spent fuel storage pool has been carried out to determine the minimum burnup of spent fuel required for safe storage. The criticality safety analysis is performed using three types of isotopic composition of spent fuel: ORIGEN2-calculated isotopic compositions; the conservative inventory obtained from the multiplication of ORIGEN2-calculated isotopic compositions by isotopic correction factors; the conservative inventory of only U, Pu and {sup 241}Am. The results show that the minimum burnup for three cases are 990,6190 and 7270 MWd/tU, respectively in the case of 5.0 wt% initial enriched spent fuel. (author). 74 refs., 68 figs., 35 tabs.

CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

International Nuclear Information System (INIS)

Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

1982-01-01

Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed

Benchmarking of World Health Organization surgical safety checklist

International Nuclear Information System (INIS)

Messahel, Farouk M.; AlQahtani, Ali S.

2009-01-01

To compare the quality of our services with the World Health Organization (WHO) surgical safety recommendations as a reference, to improve our services if they fall short of that of the WHO, and to publish our additional standards, so that they may be included in future revision of WHO checklist. We conducted this study on 15th July 2008 at the Armed Forces Hospital, Wadi Al-Dawasir, Kingdom of Saudi Arabia. We compared each WHO safety standard item with its corresponding standard in our checklist. There were 4 possibilities for the comparison: that our performance meet, was less than or exceeded the quality-of-care measures in the WHO checklist, or that there are additional safety measures in either checklist that need to be considered by each party. Since its introduction in 1997, our checklist was applied to 11828 patients and resulted in error-free outcomes. Benchmarking proved that our surgical safety performance does not only match the standards of the WHO surgical safety checklist, but also exceeds it in other safety areas (for example measures to prevent perioperative hypothermia and venous thromboembolism). Benchmarking is a continuous quality improvement process aimed at providing the best available at the time in healthcare, and we recommend its adoption by healthcare providers. The WHO surgical safety checklist is a bold step in the right direction towards safer surgical outcomes. Feedback from other medical establishments should be encouraged. (author)

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

Accelerator shielding benchmark problems

International Nuclear Information System (INIS)

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

1993-01-01

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

Benchmark calculation of nuclear design code for HCLWR

International Nuclear Information System (INIS)

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

1986-01-01

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

Benchmark testing calculations for 232Th

International Nuclear Information System (INIS)

Liu Ping

2003-01-01

The cross sections of 232 Th from CNDC and JENDL-3.3 were processed with NJOY97.45 code in the ACE format for the continuous-energy Monte Carlo Code MCNP4C. The K eff values and central reaction rates based on CENDL-3.0, JENDL-3.3 and ENDF/B-6.2 were calculated using MCNP4C code for benchmark assembly, and the comparisons with experimental results are given. (author)

Serious injuries: an additional indicator to fatalities for road safety benchmarking.

Science.gov (United States)

Shen, Yongjun; Hermans, Elke; Bao, Qiong; Brijs, Tom; Wets, Geert

2015-01-01

Almost all of the current road safety benchmarking studies focus entirely on fatalities, which, however, represent only one measure of the magnitude of the road safety problem. The main objective of this article was to investigate the possibility of including the number of serious injuries in addition to the number of fatalities for road safety benchmarking and to further illuminate its impact on the countries' rankings. We introduced the technique of data envelopment analysis (DEA) to the road safety domain and developed a DEA-based road safety model (DEA-RS) in this study. Moreover, we outlined different types of possible weight restrictions and adopted 2 of them to indicate the relationship between road fatalities and serious injuries for the sake of rational benchmarking. One was a relative weight restriction based on the information of their shadow price, and the other was a virtual weight restriction using a priori knowledge about the importance level of these 2 aspects. By computing the most optimal road safety risk scores of 10 European countries based on the different models, we found that United Kingdom was the only best-performing country no matter which model was utilized. However, countries such as The Netherlands, Sweden, and Switzerland were no longer best-performing when the serious injuries were integrated. On the contrary, Spain, which ranked almost at the bottom among all of the countries when only the number of road fatalities was considered, became a relatively well-performing country when integrating its number of serious injuries in the evaluation. In general, no matter whether the country's road safety ranking was improved or deteriorated, most of the countries achieved a higher risk score when the number of serious injuries was included, which implied that compared to the road fatalities, more policy attention has to be paid to improve the situation of serious injuries in most countries. Given the importance of considering the serious

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)

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

EA-MC Neutronic Calculations on IAEA ADS Benchmark 3.2

Energy Technology Data Exchange (ETDEWEB)

Dahlfors, Marcus [Uppsala Univ. (Sweden). Dept. of Radiation Sciences; Kadi, Yacine [CERN, Geneva (Switzerland). Emerging Energy Technologies

2006-01-15

The neutronics and the transmutation properties of the IAEA ADS benchmark 3.2 setup, the 'Yalina' experiment or ISTC project B-70, have been studied through an extensive amount of 3-D Monte Carlo calculations at CERN. The simulations were performed with the state-of-the-art computer code package EA-MC, developed at CERN. The calculational approach is outlined and the results are presented in accordance with the guidelines given in the benchmark description. A variety of experimental conditions and parameters are examined; three different fuel rod configurations and three types of neutron sources are applied to the system. Reactivity change effects introduced by removal of fuel rods in both central and peripheral positions are also computed. Irradiation samples located in a total of 8 geometrical positions are examined. Calculations of capture reaction rates in {sup 129}I, {sup 237}Np and {sup 243}Am samples and of fission reaction rates in {sup 235}U, {sup 237}Np and {sup 243}Am samples are presented. Simulated neutron flux densities and energy spectra as well as spectral indices inside experimental channels are also given according to benchmark specifications. Two different nuclear data libraries, JAR-95 and JENDL-3.2, are applied for the calculations.

The International Criticality Safety Benchmark Evaluation Project on the Internet

International Nuclear Information System (INIS)

Briggs, J.B.; Brennan, S.A.; Scott, L.

2000-01-01

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in October 1992 by the US Department of Energy's (DOE's) defense programs and is documented in the Transactions of numerous American Nuclear Society and International Criticality Safety Conferences. The work of the ICSBEP is documented as an Organization for Economic Cooperation and Development (OECD) handbook, International Handbook of Evaluated Criticality Safety Benchmark Experiments. The ICSBEP Internet site was established in 1996 and its address is http://icsbep.inel.gov/icsbep. A copy of the ICSBEP home page is shown in Fig. 1. The ICSBEP Internet site contains the five primary links. Internal sublinks to other relevant sites are also provided within the ICSBEP Internet site. A brief description of each of the five primary ICSBEP Internet site links is given

Benchmark calculation programme concerning typical LMFBR structures

International Nuclear Information System (INIS)

Donea, J.; Ferrari, G.; Grossetie, J.C.; Terzaghi, A.

1982-01-01

This programme, which is part of a comprehensive activity aimed at resolving difficulties encountered in using design procedures based on ASME Code Case N-47, should allow to get confidence in computer codes which are supposed to provide a realistic prediction of the LMFBR component behaviour. The calculations started on static analysis of typical structures made of non linear materials stressed by cyclic loads. The fluid structure interaction analysis is also being considered. Reasons and details of the different benchmark calculations are described, results obtained are commented and future computational exercise indicated

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1

International Nuclear Information System (INIS)

Van Der Marck, S. C.

2012-01-01

Three nuclear data libraries have been tested extensively using criticality safety benchmark calculations. The three libraries are the new release of the US library ENDF/B-VII.1 (2011), the new release of the Japanese library JENDL-4.0 (2011), and the OECD/NEA library JEFF-3.1 (2006). All calculations were performed with the continuous-energy Monte Carlo code MCNP (version 4C3, as well as version 6-beta1). Around 2000 benchmark cases from the International Handbook of Criticality Safety Benchmark Experiments (ICSBEP) were used. The results were analyzed per ICSBEP category, and per element. Overall, the three libraries show similar performance on most criticality safety benchmarks. The largest differences are probably caused by elements such as Be, C, Fe, Zr, W. (authors)

FENDL-2 and associated benchmark calculations

International Nuclear Information System (INIS)

Pashchenko, A.B.; Muir, D.W.

1992-03-01

The present Report contains the Summary of the IAEA Advisory Group Meeting on ''The FENDL-2 and Associated Benchmark Calculations'' convened on 18-22 November 1991, at the IAEA Headquarters in Vienna, Austria, by the IAEA Nuclear Data Section. The Advisory Group Meeting Conclusions and Recommendations and the Report on the Strategy for the Future Development of the FENDL and on Future Work towards establishing FENDL-2 are also included in this Summary Report. (author). 1 ref., 4 tabs

Method of characteristics - Based sensitivity calculations for international PWR benchmark

International Nuclear Information System (INIS)

Suslov, I. R.; Tormyshev, I. V.; Komlev, O. G.

2013-01-01

Method to calculate sensitivity of fractional-linear neutron flux functionals to transport equation coefficients is proposed. Implementation of the method on the basis of MOC code MCCG3D is developed. Sensitivity calculations for fission intensity for international PWR benchmark are performed. (authors)

Benchmark Calculations of Noncovalent Interactions of Halogenated Molecules

Czech Academy of Sciences Publication Activity Database

Řezáč, Jan; Riley, Kevin Eugene; Hobza, Pavel

2012-01-01

Roč. 8, č. 11 (2012), s. 4285-4292 ISSN 1549-9618 R&D Projects: GA ČR GBP208/12/G016 Institutional support: RVO:61388963 Keywords : halogenated molecules * noncovalent interactions * benchmark calculations Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 5.389, year: 2012

Benchmarking working conditions for health and safety in the frontline healthcare industry: Perspectives from Australia and Malaysia.

Science.gov (United States)

McLinton, Sarven S; Loh, May Young; Dollard, Maureen F; Tuckey, Michelle M R; Idris, Mohd Awang; Morton, Sharon

2018-04-06

To present benchmarks for working conditions in healthcare industries as an initial effort into international surveillance. The healthcare industry is fundamental to sustaining the health of Australians, yet it is under immense pressure. Budgets are limited, demands are increasing as are workplace injuries and all of these factors compromise patient care. Urgent attention is needed to reduce strains on workers and costs in health care, however, little work has been done to benchmark psychosocial factors in healthcare working conditions in the Asia-Pacific. Intercultural comparisons are important to provide an evidence base for public policy. A cross-sectional design was used (like other studies of prevalence), including a mixed-methods approach with qualitative interviews to better contextualize the results. Data on psychosocial factors and other work variables were collected from healthcare workers in three hospitals in Australia (N = 1,258) and Malaysia (N = 1,125). 2015 benchmarks were calculated for each variable and comparison was conducted via independent samples t tests. Healthcare samples were also compared with benchmarks for non-healthcare general working populations from their respective countries: Australia (N = 973) and Malaysia (N = 225). Our study benchmarks healthcare working conditions in Australia and Malaysia against the general working population, identifying trends that indicate the industry is in need of intervention strategies and job redesign initiatives that better support psychological health and safety. We move toward a better understanding of the precursors of psychosocial safety climate in a broader context, including similarities and differences between Australia and Malaysia in national culture, government occupational health and safety policies and top-level management practices. © 2018 John Wiley & Sons Ltd.

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.

Criticality safety benchmark evaluation project: Recovering the past

Energy Technology Data Exchange (ETDEWEB)

Trumble, E.F.

1997-06-01

A very brief summary of the Criticality Safety Benchmark Evaluation Project of the Westinghouse Savannah River Company is provided in this paper. The purpose of the project is to provide a source of evaluated criticality safety experiments in an easily usable format. Another project goal is to search for any experiments that may have been lost or contain discrepancies, and to determine if they can be used. Results of evaluated experiments are being published as US DOE handbooks.

Companies' opinions and acceptance of global food safety initiative benchmarks after implementation.

Science.gov (United States)

Crandall, Phil; Van Loo, Ellen J; O'Bryan, Corliss A; Mauromoustakos, Andy; Yiannas, Frank; Dyenson, Natalie; Berdnik, Irina

2012-09-01

International attention has been focused on minimizing costs that may unnecessarily raise food prices. One important aspect to consider is the redundant and overlapping costs of food safety audits. The Global Food Safety Initiative (GFSI) has devised benchmarked schemes based on existing international food safety standards for use as a unifying standard accepted by many retailers. The present study was conducted to evaluate the impact of the decision made by Walmart Stores (Bentonville, AR) to require their suppliers to become GFSI compliant. An online survey of 174 retail suppliers was conducted to assess food suppliers' opinions of this requirement and the benefits suppliers realized when they transitioned from their previous food safety systems. The most common reason for becoming GFSI compliant was to meet customers' requirements; thus, supplier implementation of the GFSI standards was not entirely voluntary. Other reasons given for compliance were enhancing food safety and remaining competitive. About 54 % of food processing plants using GFSI benchmarked schemes followed the guidelines of Safe Quality Food 2000 and 37 % followed those of the British Retail Consortium. At the supplier level, 58 % followed Safe Quality Food 2000 and 31 % followed the British Retail Consortium. Respondents reported that the certification process took about 10 months. The most common reason for selecting a certain GFSI benchmarked scheme was because it was widely accepted by customers (retailers). Four other common reasons were (i) the standard has a good reputation in the industry, (ii) the standard was recommended by others, (iii) the standard is most often used in the industry, and (iv) the standard was required by one of their customers. Most suppliers agreed that increased safety of their products was required to comply with GFSI benchmarked schemes. They also agreed that the GFSI required a more carefully documented food safety management system, which often required

COVE 2A Benchmarking calculations using NORIA

International Nuclear Information System (INIS)

Carrigan, C.R.; Bixler, N.E.; Hopkins, P.L.; Eaton, R.R.

1991-10-01

Six steady-state and six transient benchmarking calculations have been performed, using the finite element code NORIA, to simulate one-dimensional infiltration into Yucca Mountain. These calculations were made to support the code verification (COVE 2A) activity for the Yucca Mountain Site Characterization Project. COVE 2A evaluates the usefulness of numerical codes for analyzing the hydrology of the potential Yucca Mountain site. Numerical solutions for all cases were found to be stable. As expected, the difficulties and computer-time requirements associated with obtaining solutions increased with infiltration rate. 10 refs., 128 figs., 5 tabs

Analysis of the impact of correlated benchmark experiments on the validation of codes for criticality safety analysis

International Nuclear Information System (INIS)

Bock, M.; Stuke, M.; Behler, M.

2013-01-01

The validation of a code for criticality safety analysis requires the recalculation of benchmark experiments. The selected benchmark experiments are chosen such that they have properties similar to the application case that has to be assessed. A common source of benchmark experiments is the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments' (ICSBEP Handbook) compiled by the 'International Criticality Safety Benchmark Evaluation Project' (ICSBEP). In order to take full advantage of the information provided by the individual benchmark descriptions for the application case, the recommended procedure is to perform an uncertainty analysis. The latter is based on the uncertainties of experimental results included in most of the benchmark descriptions. They can be performed by means of the Monte Carlo sampling technique. The consideration of uncertainties is also being introduced in the supplementary sheet of DIN 25478 'Application of computer codes in the assessment of criticality safety'. However, for a correct treatment of uncertainties taking into account the individual uncertainties of the benchmark experiments is insufficient. In addition, correlations between benchmark experiments have to be handled correctly. For example, these correlations can arise due to different cases of a benchmark experiment sharing the same components like fuel pins or fissile solutions. Thus, manufacturing tolerances of these components (e.g. diameter of the fuel pellets) have to be considered in a consistent manner in all cases of the benchmark experiment. At the 2012 meeting of the Expert Group on 'Uncertainty Analysis for Criticality Safety Assessment' (UACSA) of the OECD/NEA a benchmark proposal was outlined that aimed for the determination of the impact on benchmark correlations on the estimation of the computational bias of the neutron multiplication factor (k eff ). The analysis presented here is based on this proposal. (orig.)

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

Improving patient safety culture in Saudi Arabia (2012-2015): trending, improvement and benchmarking.

Science.gov (United States)

Alswat, Khalid; Abdalla, Rawia Ahmad Mustafa; Titi, Maher Abdelraheim; Bakash, Maram; Mehmood, Faiza; Zubairi, Beena; Jamal, Diana; El-Jardali, Fadi

2017-08-02

Measuring patient safety culture can provide insight into areas for improvement and help monitor changes over time. This study details the findings of a re-assessment of patient safety culture in a multi-site Medical City in Riyadh, Kingdom of Saudi Arabia (KSA). Results were compared to an earlier assessment conducted in 2012 and benchmarked with regional and international studies. Such assessments can provide hospital leadership with insight on how their hospital is performing on patient safety culture composites as a result of quality improvement plans. This paper also explored the association between patient safety culture predictors and patient safety grade, perception of patient safety, frequency of events reported and number of events reported. We utilized a customized version of the patient safety culture survey developed by the Agency for Healthcare Research and Quality. The Medical City is a tertiary care teaching facility composed of two sites (total capacity of 904 beds). Data was analyzed using SPSS 24 at a significance level of 0.05. A t-Test was used to compare results from the 2012 survey to that conducted in 2015. Two adopted Generalized Estimating Equations in addition to two linear models were used to assess the association between composites and patient safety culture outcomes. Results were also benchmarked against similar initiatives in Lebanon, Palestine and USA. Areas of strength in 2015 included Teamwork within units, and Organizational Learning-Continuous Improvement; areas requiring improvement included Non-Punitive Response to Error, and Staffing. Comparing results to the 2012 survey revealed improvement on some areas but non-punitive response to error and Staffing remained the lowest scoring composites in 2015. Regression highlighted significant association between managerial support, organizational learning and feedback and improved survey outcomes. Comparison to international benchmarks revealed that the hospital is performing at or

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

International Nuclear Information System (INIS)

Bess, John D.; Briggs, J. Blair; Nigg, David W.

2009-01-01

One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

Measuring and benchmarking safety culture: application of the safety attitudes questionnaire to an acute medical admissions unit.

Science.gov (United States)

Relihan, E; Glynn, S; Daly, D; Silke, B; Ryder, S

2009-12-01

To assess the safety culture in an acute medical admissions unit (AMAU) of a teaching hospital in order to benchmark results against international data and guide a unit-based, integrated, risk management strategy. The safety attitudes questionnaire (SAQ), a validated instrument for the measurement of safety culture was applied to an AMAU. All AMAU healthcare staff (n = 92) were surveyed: doctors, nurses, healthcare assistants (HCAs) and allied healthcare professionals (AHPs). Safety attitude scores for the overall unit and individual caregiver types were assessed across six domains of safety culture. When compared against an international benchmark, the AMAU scored significantly higher for four of the six safety domains: p < 0.01 for 'teamwork climate', 'safety climate' and 'stress recognition' and p < 0.05 for 'job satisfaction'. The difference between nurse manager scores and the overall mean for the study group was statistically significant for the domains of 'teamwork climate' (p < 0.05) and 'safety climate' (p < 0.01). HCAs scored significantly lower relative to staff overall with regard to 'working conditions' (p < 0.05) and 'perceptions of management' (p < 0.01). The SAQ was successfully applied to an AMAU setting giving a valuable insight into staff issues of concern across the safety spectrum: employee and environmental safety, clinical risk management and medication safety.

Validation of the Continuous-Energy Monte Carlo Criticality-Safety Analysis System MVP and JENDL-3.2 Using the Internationally Evaluated Criticality Benchmarks

International Nuclear Information System (INIS)

Mitake, Susumu

2003-01-01

Validation of the continuous-energy Monte Carlo criticality-safety analysis system, comprising the MVP code and neutron cross sections based on JENDL-3.2, was examined using benchmarks evaluated in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. Eight experiments (116 configurations) for the plutonium solution and plutonium-uranium mixture systems performed at Valduc, Battelle Pacific Northwest Laboratories, and other facilities were selected and used in the studies. The averaged multiplication factors calculated with MVP and MCNP-4B using the same neutron cross-section libraries based on JENDL-3.2 were in good agreement. Based on methods provided in the Japanese nuclear criticality-safety handbook, the estimated criticality lower-limit multiplication factors to be used as a subcriticality criterion for the criticality-safety evaluation of nuclear facilities were obtained. The analysis proved the applicability of the MVP code to the criticality-safety analysis of nuclear fuel facilities, particularly to the analysis of systems fueled with plutonium and in homogeneous and thermal-energy conditions

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Dourougie, C

1998-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Primm III, RT

2002-05-29

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

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

International Nuclear Information System (INIS)

Primm III, RT

2002-01-01

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

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.

Heavy nucleus resonant absorption calculation benchmarks

International Nuclear Information System (INIS)

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

1993-01-01

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

The Safety Attitudes Questionnaire: psychometric properties, benchmarking data, and emerging research

Directory of Open Access Journals (Sweden)

Boyden James

2006-04-01

Full Text Available Abstract Background There is widespread interest in measuring healthcare provider attitudes about issues relevant to patient safety (often called safety climate or safety culture. Here we report the psychometric properties, establish benchmarking data, and discuss emerging areas of research with the University of Texas Safety Attitudes Questionnaire. Methods Six cross-sectional surveys of health care providers (n = 10,843 in 203 clinical areas (including critical care units, operating rooms, inpatient settings, and ambulatory clinics in three countries (USA, UK, New Zealand. Multilevel factor analyses yielded results at the clinical area level and the respondent nested within clinical area level. We report scale reliability, floor/ceiling effects, item factor loadings, inter-factor correlations, and percentage of respondents who agree with each item and scale. Results A six factor model of provider attitudes fit to the data at both the clinical area and respondent nested within clinical area levels. The factors were: Teamwork Climate, Safety Climate, Perceptions of Management, Job Satisfaction, Working Conditions, and Stress Recognition. Scale reliability was 0.9. Provider attitudes varied greatly both within and among organizations. Results are presented to allow benchmarking among organizations and emerging research is discussed. Conclusion The Safety Attitudes Questionnaire demonstrated good psychometric properties. Healthcare organizations can use the survey to measure caregiver attitudes about six patient safety-related domains, to compare themselves with other organizations, to prompt interventions to improve safety attitudes and to measure the effectiveness of these interventions.

Some comments on cold hydrogenous moderators, simple synthetic kernels and benchmark calculations

International Nuclear Information System (INIS)

Dorning, J.

1997-09-01

The author comments on three general subjects which are not directly related, but which in his opinion are very relevant to the objectives of the workshop. The first of these is parahydrogen moderators, about which recurring questions have been raised during the Workshop. The second topic is related to the use of simple synthetic scattering kernels in conjunction with the neutron transport equation to carry out elementary mathematical analyses and simple computational analyses in order to understand the gross physics of time-dependent neutron transport initiated by pulsed sources in cold moderators. The third subject is that of 'simple' benchmark calculations by which is meant calculations that are simple compared to the very large scale combined spallation, slowing-down, thermalization calculations using MCNP and other large Monte Carlo codes. Such benchmark problems can be created so that they are closely related to both the geometric configuration and material composition of cold moderators of interest and still can be solved using steady-state deterministic transport codes to calculate the asymptotic time-decay constant, and the time-asymptotic energy spectrum of neutrons in the cold moderator and the spectrum of the cold neutrons leaking from it (neither of which should be expected to be Maxwellian in these small leakage-dominated systems). These would provide rather precise benchmark solutions against which the results of the large scale calculations carried out for the whole spallation, slowing-down, thermalization system -- for the same decoupled cold moderator -- could be compared.

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

Benchmark calculations of thermal reaction rates. I - Quantal scattering theory

Science.gov (United States)

Chatfield, David C.; Truhlar, Donald G.; Schwenke, David W.

1991-01-01

The thermal rate coefficient for the prototype reaction H + H2 yields H2 + H with zero total angular momentum is calculated by summing, averaging, and numerically integrating state-to-state reaction probabilities calculated by time-independent quantum-mechanical scattering theory. The results are very carefully converged with respect to all numerical parameters in order to provide high-precision benchmark results for confirming the accuracy of new methods and testing their efficiency.

Benchmarking density-functional-theory calculations of rotational g tensors and magnetizabilities using accurate coupled-cluster calculations.

Science.gov (United States)

Lutnaes, Ola B; Teale, Andrew M; Helgaker, Trygve; Tozer, David J; Ruud, Kenneth; Gauss, Jürgen

2009-10-14

An accurate set of benchmark rotational g tensors and magnetizabilities are calculated using coupled-cluster singles-doubles (CCSD) theory and coupled-cluster single-doubles-perturbative-triples [CCSD(T)] theory, in a variety of basis sets consisting of (rotational) London atomic orbitals. The accuracy of the results obtained is established for the rotational g tensors by careful comparison with experimental data, taking into account zero-point vibrational corrections. After an analysis of the basis sets employed, extrapolation techniques are used to provide estimates of the basis-set-limit quantities, thereby establishing an accurate benchmark data set. The utility of the data set is demonstrated by examining a wide variety of density functionals for the calculation of these properties. None of the density-functional methods are competitive with the CCSD or CCSD(T) methods. The need for a careful consideration of vibrational effects is clearly illustrated. Finally, the pure coupled-cluster results are compared with the results of density-functional calculations constrained to give the same electronic density. The importance of current dependence in exchange-correlation functionals is discussed in light of this comparison.

Statistical Analysis of Reactor Pressure Vessel Fluence Calculation Benchmark Data Using Multiple Regression Techniques

International Nuclear Information System (INIS)

Carew, John F.; Finch, Stephen J.; Lois, Lambros

2003-01-01

The calculated >1-MeV pressure vessel fluence is used to determine the fracture toughness and integrity of the reactor pressure vessel. It is therefore of the utmost importance to ensure that the fluence prediction is accurate and unbiased. In practice, this assurance is provided by comparing the predictions of the calculational methodology with an extensive set of accurate benchmarks. A benchmarking database is used to provide an estimate of the overall average measurement-to-calculation (M/C) bias in the calculations ( ). This average is used as an ad-hoc multiplicative adjustment to the calculations to correct for the observed calculational bias. However, this average only provides a well-defined and valid adjustment of the fluence if the M/C data are homogeneous; i.e., the data are statistically independent and there is no correlation between subsets of M/C data.Typically, the identification of correlations between the errors in the database M/C values is difficult because the correlation is of the same magnitude as the random errors in the M/C data and varies substantially over the database. In this paper, an evaluation of a reactor dosimetry benchmark database is performed to determine the statistical validity of the adjustment to the calculated pressure vessel fluence. Physical mechanisms that could potentially introduce a correlation between the subsets of M/C ratios are identified and included in a multiple regression analysis of the M/C data. Rigorous statistical criteria are used to evaluate the homogeneity of the M/C data and determine the validity of the adjustment.For the database evaluated, the M/C data are found to be strongly correlated with dosimeter response threshold energy and dosimeter location (e.g., cavity versus in-vessel). It is shown that because of the inhomogeneity in the M/C data, for this database, the benchmark data do not provide a valid basis for adjusting the pressure vessel fluence.The statistical criteria and methods employed in

Initialization bias suppression in iterative Monte Carlo calculations: benchmarks on criticality calculation

International Nuclear Information System (INIS)

Richet, Y.; Jacquet, O.; Bay, X.

2005-01-01

The accuracy of an Iterative Monte Carlo calculation requires the convergence of the simulation output process. The present paper deals with a post processing algorithm to suppress the transient due to initialization applied on criticality calculations. It should be noticed that this initial transient suppression aims only at obtaining a stationary output series, then the convergence of the calculation needs to be guaranteed independently. The transient suppression algorithm consists in a repeated truncation of the first observations of the output process. The truncation of the first observations is performed as long as a steadiness test based on Brownian bridge theory is negative. This transient suppression method was previously tuned for a simplified model of criticality calculations, although this paper focuses on the efficiency on real criticality calculations. The efficiency test is based on four benchmarks with strong source convergence problems: 1) a checkerboard storage of fuel assemblies, 2) a pin cell array with irradiated fuel, 3) 3 one-dimensional thick slabs, and 4) an array of interacting fuel spheres. It appears that the transient suppression method needs to be more widely validated on real criticality calculations before any blind using as a post processing in criticality codes

A proposal of a benchmark for calculation of the power distribution next to the absorber

International Nuclear Information System (INIS)

Temesvari, E.; Hordosy, G.; Maraczy, Cs.; Hegyi, Gy.; Kereszturi, A.

1999-01-01

A proposal of a new benchmark problem was formulated to consider the characteristics of the VVER-440 fuel assembly with enrichment zoning, i. e. to study the space dependence of the power distribution near to a control assembly. A quite detailed geometry and the material composition of the fuel and the control assemblies were modeled by the help of MCNP calculations in AEKI. The results of the MCNP calculations were built in the KARATE code system as the new albedo matrices. The comparison of the KARATE calculation results and the MCNP calculations for this benchmark is presented. (Authors)

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

One dimensional benchmark calculations using diffusion theory

International Nuclear Information System (INIS)

Ustun, G.; Turgut, M.H.

1986-01-01

This is a comparative study by using different one dimensional diffusion codes which are available at our Nuclear Engineering Department. Some modifications have been made in the used codes to fit the problems. One of the codes, DIFFUSE, solves the neutron diffusion equation in slab, cylindrical and spherical geometries by using 'Forward elimination- Backward substitution' technique. DIFFUSE code calculates criticality, critical dimensions and critical material concentrations and adjoint fluxes as well. It is used for the space and energy dependent neutron flux distribution. The whole scattering matrix can be used if desired. Normalisation of the relative flux distributions to the reactor power, plotting of the flux distributions and leakage terms for the other two dimensions have been added. Some modifications also have been made for the code output. Two Benchmark problems have been calculated with the modified version and the results are compared with BBD code which is available at our department and uses same techniques of calculation. Agreements are quite good in results such as k-eff and the flux distributions for the two cases studies. (author)

Update of KASHIL-E6 library for shielding analysis and benchmark calculations

International Nuclear Information System (INIS)

Kim, D. H.; Kil, C. S.; Jang, J. H.

2004-01-01

For various shielding and reactor pressure vessel dosimetry applications, a pseudo-problem-independent neutron-photon coupled MATXS-format library based on the last release of ENDF/B-VI has been generated as a part of the update program for KASHIL-E6, which was based on ENDF/B-VI.5. It has VITAMIN-B6 neutron and photon energy group structures, i.e., 199 groups for neutron and 42 groups for photon. The neutron and photon weighting functions and the Legendre order of scattering are same as KASHIL-E6. The library has been validated through some benchmarks: the PCA-REPLICA and NESDIP-2 experiments for LWR pressure vessel facility benchmark, the Winfrith Iron88 experiment for validation of iron data, and the Winfrith Graphite experiment for validation of graphite data. These calculations were performed by the TRANSXlDANTSYS code system. In addition, the substitutions of the JENDL-3.3 and JEFF-3.0 data for Fe, Cr, Cu and Ni, which are very important nuclides for shielding analyses, were investigated to estimate the effects on the benchmark calculation results

Application of a general purpose user's version of the EGS4 code system to a photon skyshine benchmarking calculation

International Nuclear Information System (INIS)

Nojiri, I.; Fukasaku, Y.; Narita, O.

1994-01-01

A general purpose user's version of the EGS4 code system has been developed to make EGS4 easily applicable to the safety analysis of nuclear fuel cycle facilities. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with Kansas State University (KSU) photon skyshine experiment of 1977. The results of the simulation showed that this version of EGS4 would be appicable to the skyshine calculation. (author)

Validation of neutron-transport calculations in benchmark facilities for improved damage-fluence predictions

International Nuclear Information System (INIS)

Williams, M.L.; Stallmann, F.W.; Maerker, R.E.; Kam, F.B.K.

1983-01-01

An accurate determination of damage fluence accumulated by reactor pressure vessels (RPV) as a function of time is essential in order to evaluate the vessel integrity for both pressurized thermal shock (PTS) transients and end-of-life considerations. The desired accuracy for neutron exposure parameters such as displacements per atom or fluence (E > 1 MeV) is of the order of 20 to 30%. However, these types of accuracies can only be obtained realistically by validation of nuclear data and calculational methods in benchmark facilities. The purposes of this paper are to review the needs and requirements for benchmark experiments, to discuss the status of current benchmark experiments, to summarize results and conclusions obtained so far, and to suggest areas where further benchmarking is needed

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

International Nuclear Information System (INIS)

DeHart, M.D.

1993-01-01

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

Safety, codes and standards for hydrogen installations. Metrics development and benchmarking

Energy Technology Data Exchange (ETDEWEB)

Harris, Aaron P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Dedrick, Daniel E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaFleur, Angela Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); San Marchi, Christopher W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-04-01

Automakers and fuel providers have made public commitments to commercialize light duty fuel cell electric vehicles and fueling infrastructure in select US regions beginning in 2014. The development, implementation, and advancement of meaningful codes and standards is critical to enable the effective deployment of clean and efficient fuel cell and hydrogen solutions in the energy technology marketplace. Metrics pertaining to the development and implementation of safety knowledge, codes, and standards are important to communicate progress and inform future R&D investments. This document describes the development and benchmarking of metrics specific to the development of hydrogen specific codes relevant for hydrogen refueling stations. These metrics will be most useful as the hydrogen fuel market transitions from pre-commercial to early-commercial phases. The target regions in California will serve as benchmarking case studies to quantify the success of past investments in research and development supporting safety codes and standards R&D.

Analysis and evaluation of critical experiments for validation of neutron transport calculations

International Nuclear Information System (INIS)

Bazzana, S.; Blaumann, H; Marquez Damian, J.I

2009-01-01

The calculation schemes, computational codes and nuclear data used in neutronic design require validation to obtain reliable results. In the nuclear criticality safety field this reliability also translates into a higher level of safety in procedures involving fissile material. The International Criticality Safety Benchmark Evaluation Project is an OECD/NEA activity led by the United States, in which participants from over 20 countries evaluate and publish criticality safety benchmarks. The product of this project is a set of benchmark experiment evaluations that are published annually in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. With the recent participation of Argentina, this information is now available for use by the neutron calculation and criticality safety groups in Argentina. This work presents the methodology used for the evaluation of experimental data, some results obtained by the application of these methods, and some examples of the data available in the Handbook. [es

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)

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)

MCNP calculations for criticality-safety benchmarks with ENDF/B-V and ENDF/B-VI libraries

International Nuclear Information System (INIS)

Iverson, J.L.; Mosteller, R.D.

1995-01-01

The MCNP Monte Carlo code, in conjunction with its continuous-energy ENDF/B-V and ENDF/B-VI cross-section libraries, has been benchmarked against results from 27 different critical experiments. The predicted values of k eff are in excellent agreement with the benchmarks, except for the ENDF/B-V results for solutions of plutonium nitrate and, to a lesser degree, for the ENDF/B-V and ENDF/B-VI results for a bare sphere of 233 U

Benchmark experiment to verify radiation transport calculations for dosimetry in radiation therapy; Benchmark-Experiment zur Verifikation von Strahlungstransportrechnungen fuer die Dosimetrie in der Strahlentherapie

Energy Technology Data Exchange (ETDEWEB)

Renner, Franziska [Physikalisch-Technische Bundesanstalt (PTB), Braunschweig (Germany)

2016-11-01

Monte Carlo simulations are regarded as the most accurate method of solving complex problems in the field of dosimetry and radiation transport. In (external) radiation therapy they are increasingly used for the calculation of dose distributions during treatment planning. In comparison to other algorithms for the calculation of dose distributions, Monte Carlo methods have the capability of improving the accuracy of dose calculations - especially under complex circumstances (e.g. consideration of inhomogeneities). However, there is a lack of knowledge of how accurate the results of Monte Carlo calculations are on an absolute basis. A practical verification of the calculations can be performed by direct comparison with the results of a benchmark experiment. This work presents such a benchmark experiment and compares its results (with detailed consideration of measurement uncertainty) with the results of Monte Carlo calculations using the well-established Monte Carlo code EGSnrc. The experiment was designed to have parallels to external beam radiation therapy with respect to the type and energy of the radiation, the materials used and the kind of dose measurement. Because the properties of the beam have to be well known in order to compare the results of the experiment and the simulation on an absolute basis, the benchmark experiment was performed using the research electron accelerator of the Physikalisch-Technische Bundesanstalt (PTB), whose beam was accurately characterized in advance. The benchmark experiment and the corresponding Monte Carlo simulations were carried out for two different types of ionization chambers and the results were compared. Considering the uncertainty, which is about 0.7 % for the experimental values and about 1.0 % for the Monte Carlo simulation, the results of the simulation and the experiment coincide.

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)

Elasto-plastic benchmark calculations. Step 1: verification of the numerical accuracy of the computer programs

International Nuclear Information System (INIS)

Corsi, F.

1985-01-01

In connection with the design of nuclear reactors components operating at elevated temperature, design criteria need a level of realism in the prediction of inelastic structural behaviour. This concept leads to the necessity of developing non linear computer programmes, and, as a consequence, to the problems of verification and qualification of these tools. Benchmark calculations allow to carry out these two actions, involving at the same time an increased level of confidence in complex phenomena analysis and in inelastic design calculations. With the financial and programmatic support of the Commission of the European Communities (CEE) a programme of elasto-plastic benchmark calculations relevant to the design of structural components for LMFBR has been undertaken by those Member States which are developing a fast reactor project. Four principal progressive aims were initially pointed out that brought to the decision to subdivide the Benchmark effort in a calculations series of four sequential steps: step 1 to 4. The present document tries to summarize Step 1 of the Benchmark exercise, to derive some conclusions on Step 1 by comparison of the results obtained with the various codes and to point out some concluding comments on the first action. It is to point out that even if the work was designed to test the capabilities of the computer codes, another aim was to increase the skill of the users concerned

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

Energy Technology Data Exchange (ETDEWEB)

DeHart, M.D.

1993-01-01

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

Benchmark calculations with simple phantom for neutron dosimetry (2)

International Nuclear Information System (INIS)

Yukio, Sakamoto; Shuichi, Tsuda; Tatsuhiko, Sato; Nobuaki, Yoshizawa; Hideo, Hirayama

2004-01-01

Benchmark calculations for high-energy neutron dosimetry were undertaken after SATIF-5. Energy deposition in a cylindrical phantom with 100 cm radius and 30 cm depth was calculated for the irradiation of neutrons from 100 MeV to 10 GeV. Using the ICRU four-element loft tissue phantom and four single-element (hydrogen, carbon, nitrogen and oxygen) phantoms, the depth distributions of deposition energy and those total at the central region of phantoms within l cm radius and at the whole region of phantoms within 100 cm radius were calculated. The calculated results of FLUKA, MCNPX, MARS, HETC-3STEP and NMTC/JAM codes were compared. It was found that FLUKA, MARS and NMTC/JAM showed almost the same results. For the high-energy neutron incident, the MCNP-X results showed the largest ones in the total deposition energy and the HETC-3STEP results show'ed smallest ones. (author)

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

International Nuclear Information System (INIS)

Akie, Hiroshi; Ishiguro, Yukio; Takano, Hideki

1988-10-01

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

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

Benchmark calculation for the steady-state temperature distribution of the HTR-10 under full-power operation

International Nuclear Information System (INIS)

Chen Fubing; Dong Yujie; Zheng Yanhua; Shi Lei; Zhang Zuoyi

2009-01-01

Within the framework of a Coordinated Research Project on Evaluation of High Temperature Gas-Cooled Reactor Performance (CRP-5) initiated by the International Atomic Energy Agency (IAEA), the calculation of steady-state temperature distribution of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) under its initial full power experimental operation has been defined as one of the benchmark problems. This paper gives the investigation results obtained by different countries who participate in solving this benchmark problem. The validation works of the THERMIX code used by the Institute of Nuclear and New Energy Technology (INET) are also presented. For the benchmark items defined in this CRP, various calculation results correspond well with each other and basically agree the experimental results. Discrepancies existing among various code results are preliminarily attributed to different methods, models, material properties, and so on used in the computations. Temperatures calculated by THERMIX for the measuring points in the reactor internals agree well with the experimental values. The maximum fuel center temperatures calculated by the participants are much lower than the limited value of 1,230degC. According to the comparison results of code-to-code as well as code-to-experiment, THERMIX is considered to reproduce relatively satisfactory results for the CRP-5 benchmark problem. (author)

Ad hoc committee on reactor physics benchmarks

International Nuclear Information System (INIS)

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

1996-01-01

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

Benchmark calculations on residue production within the EURISOL DS project; Part I: thin targets

CERN Document Server

David, J.C; Boudard, A; Doré, D; Leray, S; Rapp, B; Ridikas, D; Thiollière, N

Report on benchmark calculations on residue production in thin targets. Calculations were performed using MCNPX 2.5.0 coupled to a selection of reaction models. The results were compared to nuclide production cross-sections measured in GSI in inverse kinematics

Benchmark calculations on residue production within the EURISOL DS project; Part II: thick targets

CERN Document Server

David, J.-C; Boudard, A; Doré, D; Leray, S; Rapp, B; Ridikas, D; Thiollière, N

Benchmark calculations on residue production using MCNPX 2.5.0. Calculations were compared to mass-distribution data for 5 different elements measured at ISOLDE, and to specific activities of 28 radionuclides in different places along the thick target measured in Dubna.

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

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.

Benchmarking Global Food Safety Performances: The Era of Risk Intelligence.

Science.gov (United States)

Valleé, Jean-Charles Le; Charlebois, Sylvain

2015-10-01

Food safety data segmentation and limitations hamper the world's ability to select, build up, monitor, and evaluate food safety performance. Currently, there is no metric that captures the entire food safety system, and performance data are not collected strategically on a global scale. Therefore, food safety benchmarking is essential not only to help monitor ongoing performance but also to inform continued food safety system design, adoption, and implementation toward more efficient and effective food safety preparedness, responsiveness, and accountability. This comparative study identifies and evaluates common elements among global food safety systems. It provides an overall world ranking of food safety performance for 17 Organisation for Economic Co-Operation and Development (OECD) countries, illustrated by 10 indicators organized across three food safety risk governance domains: risk assessment (chemical risks, microbial risks, and national reporting on food consumption), risk management (national food safety capacities, food recalls, food traceability, and radionuclides standards), and risk communication (allergenic risks, labeling, and public trust). Results show all countries have very high food safety standards, but Canada and Ireland, followed by France, earned excellent grades relative to their peers. However, any subsequent global ranking study should consider the development of survey instruments to gather adequate and comparable national evidence on food safety.

Benchmark calculations by KENO-Va using the JEF 2.2 library

Energy Technology Data Exchange (ETDEWEB)

Markova, L.

1994-12-01

This work has to be a contribution to the validation of the JEF2.2 neutron cross-section libarary, following the earlier published benchmark calculations having been performed to validate the previous version JEF1.1 of the libarary. Several simple calculational problems and one experimental problem were chosen for a criticality calculations. In addition also a realistic hexagonal arrangement of the VVER-440 fuel assemblies in a spent fuel cask were analyzed in a partly cylindrized model. All criticality calculations, carried out by the KENO-Va code using the JEF2.2 neutron cross-section library in 172 energy groups, resulted in multiplication factors (k{sub eff}) which were tabulated and compared with the results of other available calculations of the same problems. (orig.).

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

International Nuclear Information System (INIS)

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

1985-01-01

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

HELIOS calculations for UO2 lattice benchmarks

International Nuclear Information System (INIS)

Mosteller, R.D.

1998-01-01

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

Calculation and definition of safety indicators

International Nuclear Information System (INIS)

Cristian, I.; Branzeu, N.; Vidican, D.; Vladescu, G.

1997-01-01

This paper presents, based on Cernavoda safety indicators proposal, the purpose definition and calculation formulas for each of the selected safety indicators. Five categories of safety indicators for Cernavoda Unit 1 were identified, namely: overall plant safety performance; initiating events; safety system availability, physical barrier integrity; indirect indicators. Definition, calculation and use of some safety indicators are shown in a tabular form. (authors)

Final Report of the NASA Office of Safety and Mission Assurance Agile Benchmarking Team

Science.gov (United States)

Wetherholt, Martha

2016-01-01

To ensure that the NASA Safety and Mission Assurance (SMA) community remains in a position to perform reliable Software Assurance (SA) on NASAs critical software (SW) systems with the software industry rapidly transitioning from waterfall to Agile processes, Terry Wilcutt, Chief, Safety and Mission Assurance, Office of Safety and Mission Assurance (OSMA) established the Agile Benchmarking Team (ABT). The Team's tasks were: 1. Research background literature on current Agile processes, 2. Perform benchmark activities with other organizations that are involved in software Agile processes to determine best practices, 3. Collect information on Agile-developed systems to enable improvements to the current NASA standards and processes to enhance their ability to perform reliable software assurance on NASA Agile-developed systems, 4. Suggest additional guidance and recommendations for updates to those standards and processes, as needed. The ABT's findings and recommendations for software management, engineering and software assurance are addressed herein.

Benchmark calculations on resonance absorption by 238U in a PWR pin-cell geometry

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Janssen, A.J.

1993-12-01

Very accurate Monte Carlo calculations with MCNP have been performed to serve as a reference for benchmark calculations on resonance absorption by 238 U in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code ROLAIDS-CPM show that this code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (orig.)

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

Benchmark referencing of neutron dosimetry measurements

International Nuclear Information System (INIS)

Eisenhauer, C.M.; Grundl, J.A.; Gilliam, D.M.; McGarry, E.D.; Spiegel, V.

1980-01-01

The concept of benchmark referencing involves interpretation of dosimetry measurements in applied neutron fields in terms of similar measurements in benchmark fields whose neutron spectra and intensity are well known. The main advantage of benchmark referencing is that it minimizes or eliminates many types of experimental uncertainties such as those associated with absolute detection efficiencies and cross sections. In this paper we consider the cavity external to the pressure vessel of a power reactor as an example of an applied field. The pressure vessel cavity is an accessible location for exploratory dosimetry measurements aimed at understanding embrittlement of pressure vessel steel. Comparisons with calculated predictions of neutron fluence and spectra in the cavity provide a valuable check of the computational methods used to estimate pressure vessel safety margins for pressure vessel lifetimes

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

Assessment of the available {sup 233}U cross-section evaluations in the calculation of critical benchmark experiments

Energy Technology Data Exchange (ETDEWEB)

Leal, L.C.; Wright, R.Q.

1996-10-01

In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U.S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the S{sub n} transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

Assessment of the Available (Sup 233)U Cross Sections Evaluations in the Calculation of Critical Benchmark Experiments

Energy Technology Data Exchange (ETDEWEB)

Leal, L.C.

1993-01-01

In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U. S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the Sn transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

Importance of LWR best-estimate safety calculations for analysis of Fukushima-like accidents

International Nuclear Information System (INIS)

Sanchez Espinoza, V.; Ivanov, K.

2011-01-01

The safety assessment of nuclear power plants relies heavily on numerical simulations, which must include the most important physical models that are representative for the reactor type of interest. The current trends in nuclear power generation and regulation are to perform safety studies by 'best-estimate' codes that allow a realistic modeling of nuclear and thermal-hydraulic processes of the reactor core and the entire plant behavior including control and protection functions. Realistic methods are referred to as 'best-estimate' calculations, implying that they use a set of data, correlations, and methods designed to represent the phenomena, using the best available techniques. The application of best-estimate methodologies in the licensing process requires the quantification of the embedded uncertainties of the used codes. In this field many international initiatives are underway under the umbrella of the OECD such as the Light Water Reactor Uncertainty Analysis in Modeling benchmark, Oskarshamn 2 Boiling Water Reactor (BWR) Stability benchmark, Kalinin-3 VVER-1000 benchmark, etc. that underlies the importance of these issues. The Fukushima accident has shown the importance of the knowledge of the initial phase of the accident regarding the state of the core, in-vessel structures, and containment as well as the amount of fissile material inventories that potentially can be released if the safety barriers fail. For the development of mitigation and prevention measures modeling of the sequence of the events along with understanding of the key physical phenomena driving the accident progression is important. The paper presents the best-estimate coupled methodologies implemented, validated and applied at the Karlsruhe Institute Technology (KIT) for both types of LWRs - Pressurized Water Reactors (PWRs) and BWRs. Example are given with a BWR steady state and transient simulations along with corresponding uncertainty quantification. The on-going development of high

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

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

International Nuclear Information System (INIS)

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

1996-06-01

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

Benchmarking promotion and deployment activities regarding intelligent vehicle safety systems in the EU

NARCIS (Netherlands)

Kievit, M. de; Malone, K.M.; Zwijnenberg, H.; Arem, B. van

2008-01-01

This paper presents the results of a Benchmarking study performed in the European Union on Awareness and Promotion & Deployment activities related to Intelligent Vehicle Safety (IVS) systems (1). The study, commissioned by the European Commission under the Intelligent Car Initiative (a i2010

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.

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-15

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

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

International Nuclear Information System (INIS)

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

2006-07-01

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

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.

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.

Benchmarking Promotion and Deployment Activities Regarding Intelligent Vehicle Safety Systems in the European Union

NARCIS (Netherlands)

de Kievit, M.; Malone, K.M.; Zwijnenberg, H.; van Arem, B.

2008-01-01

This paper presents the results of a Benchmarking study performed in the European Union on Awareness and Promotion & Deployment activities related to Intelligent Vehicle Safety (IVS) systems (1). The study, commissioned by the European Commission under the Intelligent Car Initiative (a i2010

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-07-01

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

Development and benchmark verification of a parallelized Monte Carlo burnup calculation program MCBMPI

International Nuclear Information System (INIS)

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

2014-01-01

MCBMPI, a parallelized burnup calculation program, was developed. The program is modularized. Neutron transport calculation module employs the parallelized MCNP5 program MCNP5MPI, and burnup calculation module employs ORIGEN2, with the MPI parallel zone decomposition strategy. The program system only consists of MCNP5MPI and an interface subroutine. The interface subroutine achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, data exchanging with MCNP5MPI. Also, the program was verified with the Pressurized Water Reactor (PWR) cell burnup benchmark, the results showed that it's capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

a Proposed Benchmark Problem for Scatter Calculations in Radiographic Modelling

Science.gov (United States)

Jaenisch, G.-R.; Bellon, C.; Schumm, A.; Tabary, J.; Duvauchelle, Ph.

2009-03-01

Code Validation is a permanent concern in computer modelling, and has been addressed repeatedly in eddy current and ultrasonic modeling. A good benchmark problem is sufficiently simple to be taken into account by various codes without strong requirements on geometry representation capabilities, focuses on few or even a single aspect of the problem at hand to facilitate interpretation and to avoid that compound errors compensate themselves, yields a quantitative result and is experimentally accessible. In this paper we attempt to address code validation for one aspect of radiographic modeling, the scattered radiation prediction. Many NDT applications can not neglect scattered radiation, and the scatter calculation thus is important to faithfully simulate the inspection situation. Our benchmark problem covers the wall thickness range of 10 to 50 mm for single wall inspections, with energies ranging from 100 to 500 keV in the first stage, and up to 1 MeV with wall thicknesses up to 70 mm in the extended stage. A simple plate geometry is sufficient for this purpose, and the scatter data is compared on a photon level, without a film model, which allows for comparisons with reference codes like MCNP. We compare results of three Monte Carlo codes (McRay, Sindbad and Moderato) as well as an analytical first order scattering code (VXI), and confront them to results obtained with MCNP. The comparison with an analytical scatter model provides insights into the application domain where this kind of approach can successfully replace Monte-Carlo calculations.

Benchmark calculations on residue production within the EURISOL DS project. Part 1: thin targets

Energy Technology Data Exchange (ETDEWEB)

David, J.C.; Blideanu, V.; Boudard, A.; Dore, D.; Leray, S.; Rapp, B.; Ridikas, D.; Thiolliere, N

2006-12-15

We have begun this benchmark study using mass distribution data of reaction products obtained at GSI in inverse kinematics. This step has allowed us to make a first selection among 10 spallation models; in this way the first assessment of the quality of the models was obtained. Then, in a second part, experimental mass distributions for some elements, which either are interesting as radioactive ion beams or important due to the safety and radioprotection issues (alpha or gamma emitters), will be also compared to model calculations. These data have been obtained for an equivalent 0.8 or 1.0 GeV proton beam, which is approximately the proposed projectile energy. We note that in realistic thick targets the proton beam will be slowed down and some secondary particles will be produced. Therefore, the residual nuclei production at lower energies is also important. For this reason, we also performed in the third part of this work some excitation function calculations and the associated data obtained with gamma-spectroscopy to test the models in a wide projectile energy range. We conclude that INCL4/Abla and Isabel/Abla are the best model combinations which we recommend. We also note that the agreement between model and data are better with 1 GeV protons than with 100-200 MeV protons.

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)

Calculations to an IAHR-benchmark test using the CFD-code CFX-4

Energy Technology Data Exchange (ETDEWEB)

Krepper, E

1998-10-01

The calculation concerns a test, which was defined as a benchmark for 3-D codes by the working group of advanced nuclear reactor types of IAHR (International Association of Hydraulic Research). The test is well documented and detailed measuring results are available. The test aims at the investigation of phenomena, which are important for heat removal at natural circulation conditions in a nuclear reactor. The task for the calculation was the modelling of the forced flow field of a single phase incompressible fluid with consideration of heat transfer and influence of gravity. These phenomena are typical also for other industrial processes. The importance of correct modelling of these phenomena also for other applications is a motivation for performing these calculations. (orig.)

DRY TRANSFER FACILITY CRITICALITY SAFETY CALCULATIONS

International Nuclear Information System (INIS)

C.E. Sanders

2005-01-01

This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Dry Transfer Facility Description Document'' (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: (1) Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2005 [DIRS 171429], Section 7). (2) Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the ''Canister Handling Facility Criticality Safety Calculations'' (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the

International report to validate criticality safety calculations for fissile material transport

International Nuclear Information System (INIS)

Whitesides, G.E.

1984-01-01

During the past three years a Working Group established by the Organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA) in Paris, France, has been studying the validity and applicability of a variety of criticality safety computer programs and their associated nuclear data for the computation of the neutron multiplication factor, k/sub eff/, for various transport packages used in the fuel cycle. The principal objective of this work has been to provide an internationally acceptable basis for the licensing authorities in a country to honor licensing approvals granted by other participating countries. Eleven countries participated in the initial study which consisted of examining criticality safety calculations for packages designed for spent light water reactor fuel transport. This paper presents a summary of this study which has been completed and reported in an OECD-NEA Report No. CSNI-71. The basic goal of this study was to outline a satisfactory validation procedure for this particular application. First, a set of actual critical experiments were chosen which contained the various material and geometric properties present in typical LWR transport containers. Secondly, calculations were made by each of the methods in order to determine how accurately each method reproduced the experimental values. This successful effort in developing a benchmark procedure for validating criticality calculations for spent LWR transport packages along with the successful intercomparison of a number of methods should provide increased confidence by licensing authorities in the use of these methods for this area of application. 4 references, 2 figures

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)

A benchmark test of computer codes for calculating average resonance parameters

International Nuclear Information System (INIS)

Ribon, P.; Thompson, A.

1983-01-01

A set of resonance parameters has been generated from known, but secret, average values; the parameters have then been adjusted to mimic experimental data by including the effects of Doppler broadening, resolution broadening and statistical fluctuations. Average parameters calculated from the dataset by various computer codes are compared with each other, and also with the true values. The benchmark test is fully described in the report NEANDC160-U (NEA Data Bank Newsletter No. 27 July 1982); the present paper is a summary of this document. (Auth.)

Availability of Neutronics Benchmarks in the ICSBEP and IRPhEP Handbooks for Computational Tools Testing

Energy Technology Data Exchange (ETDEWEB)

Bess, John D.; Briggs, J. Blair; Ivanova, Tatiana; Hill, Ian; Gulliford, Jim

2017-02-01

In the past several decades, numerous experiments have been performed worldwide to support reactor operations, measurements, design, and nuclear safety. Those experiments represent an extensive international investment in infrastructure, expertise, and cost, representing significantly valuable resources of data supporting past, current, and future research activities. Those valuable assets represent the basis for recording, development, and validation of our nuclear methods and integral nuclear data [1]. The loss of these experimental data, which has occurred all too much in the recent years, is tragic. The high cost to repeat many of these measurements can be prohibitive, if not impossible, to surmount. Two international projects were developed, and are under the direction of the Organisation for Co-operation and Development Nuclear Energy Agency (OECD NEA) to address the challenges of not just data preservation, but evaluation of the data to determine its merit for modern and future use. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was established to identify and verify comprehensive critical benchmark data sets; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data [2]. Similarly, the International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications [3]. Annually, contributors from around the world continue to collaborate in the evaluation and review of select benchmark experiments for preservation and dissemination. The extensively peer-reviewed integral benchmark data can then be utilized to support nuclear design and safety analysts to validate the analytical tools, methods, and data needed for next

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-09-01

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

Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark

International Nuclear Information System (INIS)

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

2008-01-01

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

Two-group k-eigenvalue benchmark calculations for planar geometry transport in a binary stochastic medium

International Nuclear Information System (INIS)

Davis, I.M.; Palmer, T.S.

2005-01-01

Benchmark calculations are performed for neutron transport in a two material (binary) stochastic multiplying medium. Spatial, angular, and energy dependence are included. The problem considered is based on a fuel assembly of a common pressurized water reactor. The mean chord length through the assembly is determined and used as the planar geometry system length. According to assumed or calculated material distributions, this system length is populated with alternating fuel and moderator segments of random size. Neutron flux distributions are numerically computed using a discretized form of the Boltzmann transport equation employing diffusion synthetic acceleration. Average quantities (group fluxes and k-eigenvalue) and variances are calculated from an ensemble of realizations of the mixing statistics. The effects of varying two parameters in the fuel, two different boundary conditions, and three different sets of mixing statistics are assessed. A probability distribution function (PDF) of the k-eigenvalue is generated and compared with previous research. Atomic mix solutions are compared with these benchmark ensemble average flux and k-eigenvalue solutions. Mixing statistics with large standard deviations give the most widely varying ensemble solutions of the flux and k-eigenvalue. The shape of the k-eigenvalue PDF qualitatively agrees with previous work. Its overall shape is independent of variations in fuel cross-sections for the problems considered, but its width is impacted by these variations. Statistical distributions with smaller standard deviations alter the shape of this PDF toward a normal distribution. The atomic mix approximation yields large over-predictions of the ensemble average k-eigenvalue and under-predictions of the flux. Qualitatively correct flux shapes are obtained in some cases. These benchmark calculations indicate that a model which includes higher statistical moments of the mixing statistics is needed for accurate predictions of binary

Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

International Nuclear Information System (INIS)

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

2007-01-01

Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm/shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm/shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the ''International Handbook of Evaluated Criticality Safety Benchmark Experiments'' have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement/shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency

Ensuring the validity of calculated subcritical limits

International Nuclear Information System (INIS)

Clark, H.K.

1977-01-01

The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionally subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin

Ward based community road safety performance benchmarking, monitoring and intervention programmes in the City of Johannesburg

CSIR Research Space (South Africa)

Ribbens, H

2008-07-01

Full Text Available benchmarking, monitoring and intervention programme. Community road safety needs in the respective wards are articulated through the ward councillor. The rationale is that the community exactly knows where these problem areas are, because they suffer as a...

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

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)

Calculations of different transmutation concepts. An international benchmark exercise

International Nuclear Information System (INIS)

2000-01-01

In April 1996, the NEA Nuclear Science Committee (NSC) Expert Group on Physics Aspects of Different Transmutation Concepts launched a benchmark exercise to compare different transmutation concepts based on pressurised water reactors (PWRs), fast reactors, and an accelerator-driven system. The aim was to investigate the physics of complex fuel cycles involving reprocessing of spent PWR reactor fuel and its subsequent reuse in different reactor types. The objective was also to compare the calculated activities for individual isotopes as a function of time for different plutonium and minor actinide transmutation scenarios in different reactor systems. This report gives the analysis of results of the 15 solutions provided by the participants: six for the PWRs, six for the fast reactor and three for the accelerator case. Various computer codes and nuclear data libraries were applied. (author)

Benchmark evaluation of the RELAP code to calculate boiling in narrow channels

International Nuclear Information System (INIS)

Kunze, J.F.; Loyalka, S.K.; McKibben, J.C.; Hultsch, R.; Oladiran, O.

1990-01-01

The RELAP code has been tested with benchmark experiments (such as the loss-of-fluid test experiments at the Idaho National Engineering Laboratory) at high pressures and temperatures characteristic of those encountered in loss-of-coolant accidents (LOCAs) in commercial light water power reactors. Application of RELAP to the LOCA analysis of a low pressure (< 7 atm) and low temperature (< 100 degree C), plate-type research reactor, such as the University of Missouri Research Reactor (MURR), the high-flux breeder reactor, high-flux isotope reactor, and Advanced Test Reactor, requires resolution of questions involving overextrapolation to very low pressures and low temperatures, and calculations of the pulsed boiling/reflood conditions in the narrow rectangular cross-section channels (typically 2 mm thick) of the plate fuel elements. The practical concern of this problem is that plate fuel temperatures predicted by RELAP5 (MOD2, version 3) during the pulsed boiling period can reach high enough temperatures to cause plate (clad) weakening, though not melting. Since an experimental benchmark of RELAP under such LOCA conditions is not available and since such conditions present substantial challenges to the code, it is important to verify the code predictions. The comparison of the pulsed boiling experiments with the RELAP calculations involves both visual observations of void fraction versus time and measurements of temperatures near the fuel plate surface

Calculational study of benchmark critical experiments on high-enriched uranyl nitrate solution systems

International Nuclear Information System (INIS)

Oh, I.; Rothe, R.E.

1978-01-01

Criticality calculations on minimally reflected, concrete-reflected, and plastic-reflected single tanks and on arrays of cylinders reflected by concrete and plastic have been performed using the KENO-IV code with 16-group Hansen-Roach neutron cross sections. The fissile material was high-enriched (93.17% 235 U) uranyl nitrate [UO 2 (NO 3 ) 2 ] solution. Calculated results are compared with those from a benchmark critical experiments program to provide the best possible verification of the calculational technique. The calculated k/sub eff/'s underestimate the critical condition by an average of 1.28% for the minimally reflected single tanks, 1.09% for the concrete-reflected single tanks, 0.60% for the plastic-reflected single tanks, 0.75% for the concrete-reflected arrays of cylinders, and 0.51% for the plastic-reflected arrays of cylinders. More than half of the present comparisons were within 1% of the experimental values, and the worst calculational and experimental discrepancy was 2.3% in k/sub eff/ for the KENO calculations

International benchmark on the natural convection test in Phenix reactor

International Nuclear Information System (INIS)

Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

2013-01-01

Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

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

International Nuclear Information System (INIS)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

2002-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Benchmarking study and its application for shielding analysis of large accelerator facilities

Energy Technology Data Exchange (ETDEWEB)

Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk [POSTECH, Pohang (Korea, Republic of)

2015-10-15

Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes.

Benchmarking study and its application for shielding analysis of large accelerator facilities

International Nuclear Information System (INIS)

Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk

2015-01-01

Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes

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.

Specification of phase 3 benchmark (Hex-Z heterogeneous and burnup calculation)

International Nuclear Information System (INIS)

Kim, Y.I.

2002-01-01

During the second RCM of the IAEA Co-ordinated Research Project Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects the following items were identified as important. Heterogeneity will affect absolute core reactivity. Rod worths could be considerably reduced by heterogeneity effects depending on their detailed design. Heterogeneity effects will affect the resonance self-shielding in the treatment of fuel Doppler, steel Doppler and sodium density effects. However, it was considered more important to concentrate on the sodium density effect in order to reduce the calculational effort required. It was also recognized that burnup effects will have an influence on fuel Doppler and sodium worths. A benchmark for the assessment of heterogeneity effect for Phase 3 was defined. It is to be performed for the Hex-Z model of the reactor only. No calculations will be performed for the R-Z model. For comparison with heterogeneous evaluations, the control rod worth will be calculated at the beginning of the equilibrium cycle, based on the homogeneous model. The definitions of rod raised and rod inserted for SHR are given, using the composition numbers

Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

Directory of Open Access Journals (Sweden)

Tanaka Ken-ichi

2016-01-01

Full Text Available We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV of a Boiling Water Reactor (BWR by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au and Nickel (Ni at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

International Nuclear Information System (INIS)

Bowman, S.M.

1991-04-01

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

Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

Energy Technology Data Exchange (ETDEWEB)

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

2007-05-01

Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm / shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm / shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement / shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy

[An approach to care indicators benchmarking. Learning to improve patient safety].

Science.gov (United States)

de Andrés Gimeno, B; Salazar de la Guerra, R M; Ferrer Arnedo, C; Revuelta Zamorano, M; Ayuso Murillo, D; González Soria, J

2014-01-01

Improvements in clinical safety can be achieved by promoting a safety culture, professional training, and learning through benchmarking. The aim of this study was to identify areas for improvement after analysing the safety indicators in two public Hospitals in North-West Madrid Region. Descriptive study performed during 2011 in Hospital Universitario Puerta de Hierro Majadahonda (HUPHM) and Hospital de Guadarrama (HG). The variables under study were 40 indicators on nursing care related to patient safety. Nineteen of them were defined in the SENECA project as care quality standards in order to improve patient safety in the hospitals. The data collected were clinical history, Madrid Health Service assessment reports, care procedures, and direct observation Within the 40 indicators: 22 of them were structured (procedures), HUPHM had 86%, and HG 95% 14 process indicators (training and protocols compliance) with similar results in both hospitals, apart from the care continuity reports and training in hand hygiene. The 4 results indicators (pressure ulcer, falls and pain) showed different results. The analysis of the indicators allowed the following actions to be taken: to identify improvements to be made in each hospital, to develop joint safety recommendations in nursing care protocols in prevention and treatment of chronic wound, to establish systematic pain assessments, and to prepare continuity care reports on all patients transferred from HUPHM to HG. Copyright © 2013 SECA. Published by Elsevier Espana. All rights reserved.

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

Calculations of the IAEA-CRP-6 Benchmark Cases by Using the ABAQUS FE Model for a Comparison with the COPA Results

International Nuclear Information System (INIS)

Cho, Moon-Sung; Kim, Y. M.; Lee, Y. W.; Jeong, K. C.; Kim, Y. K.; Oh, S. C.

2006-01-01

The fundamental design for a gas-cooled reactor relies on an understanding of the behavior of a coated particle fuel. KAERI, which has been carrying out the Korean VHTR (Very High Temperature modular gas cooled Reactor) Project since 2004, is developing a fuel performance analysis code for a VHTR named COPA (COated Particle fuel Analysis). COPA predicts temperatures, stresses, a fission gas release and failure probabilities of a coated particle fuel in normal operating conditions. Validation of COPA in the process of its development is realized partly by participating in the benchmark section of the international CRP-6 program led by IAEA which provides comprehensive benchmark problems and analysis results obtained from the CRP-6 member countries. Apart from the validation effort through the CRP-6, a validation of COPA was attempted by comparing its benchmark results with the visco-elastic solutions obtained from the ABAQUS code calculations for the same CRP-6 TRISO coated particle benchmark problems involving creep, swelling, and pressure. The study shows the calculation results of the IAEA-CRP-6 benchmark cases 5 through 7 by using the ABAQUS FE model for a comparison with the COPA results

Evaluation of PWR and BWR assembly benchmark calculations. Status report of EPRI computational benchmark results, performed in the framework of the Netherlands` PINK programme (Joint project of ECN, IRI, KEMA and GKN)

Energy Technology Data Exchange (ETDEWEB)

Gruppelaar, H. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Klippel, H.T. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Verhagen, F.C.M. [Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands); Bruggink, J.C. [Gemeenschappelijke Kernenergiecentrale Nederland N.V., Dodewaard (Netherlands)

1993-11-01

Benchmark results of the Dutch PINK working group on calculational benchmarks on single pin cell and multipin assemblies as defined by EPRI are presented and evaluated. First a short update of methods used by the various institutes involved is given as well as an update of the status with respect to previous performed pin-cell calculations. Problems detected in previous pin-cell calculations are inspected more closely. Detailed discussion of results of multipin assembly calculations is given. The assembly consists of 9 pins in a multicell square lattice in which the central pin is filled differently, i.e. a Gd pin for the BWR assembly and a control rod/guide tube for the PWR assembly. The results for pin cells showed a rather good overall agreement between the four participants although BWR pins with high void fraction turned out to be difficult to calculate. With respect to burnup calculations good overall agreement for the reactivity swing was obtained, provided that a fine time grid is used. (orig.)

Evaluation of PWR and BWR assembly benchmark calculations. Status report of EPRI computational benchmark results, performed in the framework of the Netherlands' PINK programme (Joint project of ECN, IRI, KEMA and GKN)

International Nuclear Information System (INIS)

Gruppelaar, H.; Klippel, H.T.; Kloosterman, J.L.; Hoogenboom, J.E.; Bruggink, J.C.

1993-11-01

Benchmark results of the Dutch PINK working group on calculational benchmarks on single pin cell and multipin assemblies as defined by EPRI are presented and evaluated. First a short update of methods used by the various institutes involved is given as well as an update of the status with respect to previous performed pin-cell calculations. Problems detected in previous pin-cell calculations are inspected more closely. Detailed discussion of results of multipin assembly calculations is given. The assembly consists of 9 pins in a multicell square lattice in which the central pin is filled differently, i.e. a Gd pin for the BWR assembly and a control rod/guide tube for the PWR assembly. The results for pin cells showed a rather good overall agreement between the four participants although BWR pins with high void fraction turned out to be difficult to calculate. With respect to burnup calculations good overall agreement for the reactivity swing was obtained, provided that a fine time grid is used. (orig.)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

International Nuclear Information System (INIS)

C.E. Sanders

2005-01-01

This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Radiation shielding and safety design

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Ouk; Gil, C. S.; Cho, Y. S.; Kim, D. H.; Kim, H. I.; Kim, J. W.; Lee, C. W.; Kim, K. Y.; Kim, B. H. [KAERI, Daejeon (Korea, Republic of)

2011-07-15

A benchmarking for the test facility, evaluations of the prompt radiation fields, evaluation of the induced activities in the facility, and estimation of the radiological impact on the environment were performed in this study. and the radiation safety analysis report for nuclear licensing was written based on this study. In the benchmark calculation, the neutron spectra was measured in the 20 Mev test facility and the measurements were compared with the computational results to verify the calculation system. In the evaluation of the prompt radiation fields, the shielding design for 100 MeV target rooms, evaluations of the leakage doses from the accidents and skyshine analysis were performed. The evaluation of the induced activities were performed for the coolant, inside air, structural materials, soil and ground-water. At last, the radiation safety analysis report was written based on results from these studies

Diagnostic Algorithm Benchmarking

Science.gov (United States)

Poll, Scott

2011-01-01

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

WWER-1000 Burnup Credit Benchmark (CB5)

International Nuclear Information System (INIS)

Manolova, M.A.

2002-01-01

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

CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

Energy Technology Data Exchange (ETDEWEB)

C.E. Sanders

2005-04-07

This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

Implications of Monte Carlo Statistical Errors in Criticality Safety Assessments

International Nuclear Information System (INIS)

Pevey, Ronald E.

2005-01-01

Most criticality safety calculations are performed using Monte Carlo techniques because of Monte Carlo's ability to handle complex three-dimensional geometries. For Monte Carlo calculations, the more histories sampled, the lower the standard deviation of the resulting estimates. The common intuition is, therefore, that the more histories, the better; as a result, analysts tend to run Monte Carlo analyses as long as possible (or at least to a minimum acceptable uncertainty). For Monte Carlo criticality safety analyses, however, the optimization situation is complicated by the fact that procedures usually require that an extra margin of safety be added because of the statistical uncertainty of the Monte Carlo calculations. This additional safety margin affects the impact of the choice of the calculational standard deviation, both on production and on safety. This paper shows that, under the assumptions of normally distributed benchmarking calculational errors and exact compliance with the upper subcritical limit (USL), the standard deviation that optimizes production is zero, but there is a non-zero value of the calculational standard deviation that minimizes the risk of inadvertently labeling a supercritical configuration as subcritical. Furthermore, this value is shown to be a simple function of the typical benchmarking step outcomes--the bias, the standard deviation of the bias, the upper subcritical limit, and the number of standard deviations added to calculated k-effectives before comparison to the USL

Proposal on the accelerator driven molten-salt reactor (ATW concept) benchmark calculations. (STAGE 1 - without an external neutron source)

International Nuclear Information System (INIS)

Svarny, J.; Mikolas, P.

1999-01-01

The first stage of ATW neutronic benchmark (without an external source), based on the simple modelling of two component concept is presented. The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark is not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (author)

The solution of the LEU and MOX WWER-1000 calculation benchmark with the CARATE - multicell code

International Nuclear Information System (INIS)

Hordosy, G.; Maraczy, Cs.

2000-01-01

Preparations for disposition of weapons grade plutonium in WWER-1000 reactors are in progress. Benchmark: Defined by the Kurchatov Institute (S. Bychkov, M. Kalugin, A. Lazarenko) to assess the applicability of computer codes for weapons grade MOX assembly calculations. Framework: 'Task force on reactor-based plutonium disposition' of OECD Nuclear Energy Agency. (Authors)

Benchmark calculations in multigroup and multidimensional time-dependent transport

International Nuclear Information System (INIS)

Ganapol, B.D.; Musso, E.; Ravetto, P.; Sumini, M.

1990-01-01

It is widely recognized that reliable benchmarks are essential in many technical fields in order to assess the response of any approximation to the physics of the problem to be treated and to verify the performance of the numerical methods used. The best possible benchmarks are analytical solutions to paradigmatic problems where no approximations are actually introduced and the only error encountered is connected to the limitations of computational algorithms. Another major advantage of analytical solutions is that they allow a deeper understanding of the physical features of the model, which is essential for the intelligent use of complicated codes. In neutron transport theory, the need for benchmarks is particularly great. In this paper, the authors propose to establish accurate numerical solutions to some problems concerning the migration of neutron pulses. Use will be made of the space asymptotic theory, coupled with a Laplace transformation inverted by a numerical technique directly evaluating the inversion integral

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

Establishing benchmarks for the management of elevated liver enzymes and/or dilated biliary trees in an urban safety net hospital: analysis of 915 subjects.

Science.gov (United States)

Liu, Laindy; Cripps, Michael W; Riggle, Andrew J; Wolf, Steven E; Nakonezny, Paul A; Phelan, Herb A

2015-12-01

The push for public reporting of outcomes necessitates relevant benchmarks for disease states across different settings. This study establishes benchmarks for choledocholithiasis management in a safety net hospital setting. We reviewed all patients admitted to our acute care surgery service with biochemical evidence of choledocholithiasis who underwent same-admission cholecystectomy (CCY) between July 2012 and December 2013. During this 18-month period, 915 patients were admitted with biochemical evidence of choledocholithiasis. Descriptive statistics for the cohort are provided, which include a 51% rate of obesity and 95% rate of pathologic cholecystitis. Conversion rates of 4% and complication rates of 6% were found. The majority had a CCY without biliary imaging (n = 630, 68.9%). Relevant benchmarks are characterized, and results of a practice pattern of omitting pre- or intraoperative biliary tree imaging are described. These findings serve as a first benchmark of choledocholithiasis management for urban safety net hospitals. Copyright © 2015 Elsevier Inc. All rights reserved.

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.

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)

Radiation Detection Computational Benchmark Scenarios

Energy Technology Data Exchange (ETDEWEB)

Shaver, Mark W.; Casella, Andrew M.; Wittman, Richard S.; McDonald, Ben S.

2013-09-24

Modeling forms an important component of radiation detection development, allowing for testing of new detector designs, evaluation of existing equipment against a wide variety of potential threat sources, and assessing operation performance of radiation detection systems. This can, however, result in large and complex scenarios which are time consuming to model. A variety of approaches to radiation transport modeling exist with complementary strengths and weaknesses for different problems. This variety of approaches, and the development of promising new tools (such as ORNL’s ADVANTG) which combine benefits of multiple approaches, illustrates the need for a means of evaluating or comparing different techniques for radiation detection problems. This report presents a set of 9 benchmark problems for comparing different types of radiation transport calculations, identifying appropriate tools for classes of problems, and testing and guiding the development of new methods. The benchmarks were drawn primarily from existing or previous calculations with a preference for scenarios which include experimental data, or otherwise have results with a high level of confidence, are non-sensitive, and represent problem sets of interest to NA-22. From a technical perspective, the benchmarks were chosen to span a range of difficulty and to include gamma transport, neutron transport, or both and represent different important physical processes and a range of sensitivity to angular or energy fidelity. Following benchmark identification, existing information about geometry, measurements, and previous calculations were assembled. Monte Carlo results (MCNP decks) were reviewed or created and re-run in order to attain accurate computational times and to verify agreement with experimental data, when present. Benchmark information was then conveyed to ORNL in order to guide testing and development of hybrid calculations. The results of those ADVANTG calculations were then sent to PNNL for

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Impact of the 235U Covariance Data in Benchmark Calculations

International Nuclear Information System (INIS)

Leal, Luiz C.; Mueller, D.; Arbanas, G.; Wiarda, D.; Derrien, H.

2008-01-01

The error estimation for calculated quantities relies on nuclear data uncertainty information available in the basic nuclear data libraries such as the U.S. Evaluated Nuclear Data File (ENDF/B). The uncertainty files (covariance matrices) in the ENDF/B library are generally obtained from analysis of experimental data. In the resonance region, the computer code SAMMY is used for analyses of experimental data and generation of resonance parameters. In addition to resonance parameters evaluation, SAMMY also generates resonance parameter covariance matrices (RPCM). SAMMY uses the generalized least-squares formalism (Bayes method) together with the resonance formalism (R-matrix theory) for analysis of experimental data. Two approaches are available for creation of resonance-parameter covariance data. (1) During the data-evaluation process, SAMMY generates both a set of resonance parameters that fit the experimental data and the associated resonance-parameter covariance matrix. (2) For existing resonance-parameter evaluations for which no resonance-parameter covariance data are available, SAMMY can retroactively create a resonance-parameter covariance matrix. The retroactive method was used to generate covariance data for 235U. The resulting 235U covariance matrix was then used as input to the PUFF-IV code, which processed the covariance data into multigroup form, and to the TSUNAMI code, which calculated the uncertainty in the multiplication factor due to uncertainty in the experimental cross sections. The objective of this work is to demonstrate the use of the 235U covariance data in calculations of critical benchmark systems

Impact of the 235U covariance data in benchmark calculations

International Nuclear Information System (INIS)

Leal, Luiz; Mueller, Don; Arbanas, Goran; Wiarda, Dorothea; Derrien, Herve

2008-01-01

The error estimation for calculated quantities relies on nuclear data uncertainty information available in the basic nuclear data libraries such as the U.S. Evaluated Nuclear Data File (ENDF/B). The uncertainty files (covariance matrices) in the ENDF/B library are generally obtained from analysis of experimental data. In the resonance region, the computer code SAMMY is used for analyses of experimental data and generation of resonance parameters. In addition to resonance parameters evaluation, SAMMY also generates resonance parameter covariance matrices (RPCM). SAMMY uses the generalized least-squares formalism (Bayes' method) together with the resonance formalism (R-matrix theory) for analysis of experimental data. Two approaches are available for creation of resonance-parameter covariance data. (1) During the data-evaluation process, SAMMY generates both a set of resonance parameters that fit the experimental data and the associated resonance-parameter covariance matrix. (2) For existing resonance-parameter evaluations for which no resonance-parameter covariance data are available, SAMMY can retroactively create a resonance-parameter covariance matrix. The retroactive method was used to generate covariance data for 235 U. The resulting 235 U covariance matrix was then used as input to the PUFF-IV code, which processed the covariance data into multigroup form, and to the TSUNAMI code, which calculated the uncertainty in the multiplication factor due to uncertainty in the experimental cross sections. The objective of this work is to demonstrate the use of the 235 U covariance data in calculations of critical benchmark systems. (authors)

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

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

Assessment of uncertainty in full core reactor physics calculations using statistical methods

International Nuclear Information System (INIS)

McEwan, C.

2012-01-01

The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (author)

Assessment of uncertainty in full core reactor physics calculations using statistical methods

Energy Technology Data Exchange (ETDEWEB)

McEwan, C., E-mail: mcewac2@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

2012-07-01

The best estimate method of safety analysis involves choosing a realistic set of input parameters for a proposed safety case and evaluating the uncertainty in the results. Determining the uncertainty in code outputs remains a challenge and is the subject of a benchmarking exercise proposed by the Organization for Economic Cooperation and Development. The work proposed in this paper will contribute to this benchmark by assessing the uncertainty in a depletion calculation of the final nuclide concentrations for an experiment performed in the Fukushima-2 reactor. This will be done using lattice transport code DRAGON and a tool known as DINOSAUR. (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.

Attila calculations for the 3-D C5G7 benchmark extension

International Nuclear Information System (INIS)

Wareing, T.A.; McGhee, J.M.; Barnett, D.A.; Failla, G.A.

2005-01-01

The performance of the Attila radiation transport software was evaluated for the 3-D C5G7 MOX benchmark extension, a follow-on study to the MOX benchmark developed by the 'OECD/NEA Expert Group on 3-D Radiation Transport Benchmarks'. These benchmarks were designed to test the ability of modern deterministic transport methods to model reactor problems without spatial homogenization. Attila is a general purpose radiation transport software package with an integrated graphical user interface (GUI) for analysis, set-up and postprocessing. Attila provides solutions to the discrete-ordinates form of the linear Boltzmann transport equation on a fully unstructured, tetrahedral mesh using linear discontinuous finite-element spatial differencing in conjunction with diffusion synthetic acceleration of inner iterations. The results obtained indicate that Attila can accurately solve the benchmark problem without spatial homogenization. (authors)

An international benchmark on safety review practices at nuclear power plants

International Nuclear Information System (INIS)

Wahlstroem, B.; Kettunen, J.

2000-02-01

A benchmarking exercise on safety review practices at nuclear power plants in Finland, Sweden and the United Kingdom has been carried out. In the exercise a comparison was made between documented practices at the Forsmark, Hinkley Point A and Olkiluoto nuclear power plants. In addition a total of 28 persons at FKA, Magnox and TVO were interviewed on their views on the efficiency of the plant modification processes in the later half of 1997. One specific example of a plant modification was selected from each of the nuclear power plant sites to provide a basis for the comparison. The report gives an account of the methodology used, a description of the plant modification projects, impressions from the interviews, potential problem areas and suggestions for possible improvements. (orig.)

H.B. Robinson-2 pressure vessel benchmark

Energy Technology Data Exchange (ETDEWEB)

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

1998-02-01

The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Section 1 of this report describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity. In section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90 {+-} 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 {+-} 0.10, 0.91 {+-} 0.10, and 0.90 {+-} 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively. It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.

Benchmarks for Uncertainty Analysis in Modelling (UAM) for the Design, Operation and Safety Analysis of LWRs - Volume I: Specification and Support Data for Neutronics Cases (Phase I)

International Nuclear Information System (INIS)

Ivanov, K.; Avramova, M.; Kamerow, S.; Kodeli, I.; Sartori, E.; Ivanov, E.; Cabellos, O.

2013-01-01

The objective of the OECD LWR UAM activity is to establish an internationally accepted benchmark framework to compare, assess and further develop different uncertainty analysis methods associated with the design, operation and safety of LWRs. As a result, the LWR UAM benchmark will help to address current nuclear power generation industry and regulation needs and issues related to practical implementation of risk-informed regulation. The realistic evaluation of consequences must be made with best-estimate coupled codes, but to be meaningful, such results should be supplemented by an uncertainty analysis. The use of coupled codes allows us to avoid unnecessary penalties due to incoherent approximations in the traditional decoupled calculations, and to obtain more accurate evaluation of margins regarding licensing limit. This becomes important for licensing power upgrades, improved fuel assembly and control rod designs, higher burn-up and others issues related to operating LWRs as well as to the new Generation 3+ designs being licensed now (ESBWR, AP-1 000, EPR-1 600, etc.). Establishing an internationally accepted LWR UAM benchmark framework offers the possibility to accelerate the licensing process when using best estimate methods. The proposed technical approach is to establish a benchmark for uncertainty analysis in best-estimate modelling and coupled multi-physics and multi-scale LWR analysis, using as bases a series of well-defined problems with complete sets of input specifications and reference experimental data. The objective is to determine the uncertainty in LWR system calculations at all stages of coupled reactor physics/thermal hydraulics calculations. The full chain of uncertainty propagation from basic data, engineering uncertainties, across different scales (multi-scale), and physics phenomena (multi-physics) will be tested on a number of benchmark exercises for which experimental data are available and for which the power plant details have been

OECD/NEA International Benchmark exercises: Validation of CFD codes applied nuclear industry; OECD/NEA internatiion Benchmark exercices: La validacion de los codigos CFD aplicados a la industria nuclear

Energy Technology Data Exchange (ETDEWEB)

Pena-Monferrer, C.; Miquel veyrat, A.; Munoz-Cobo, J. L.; Chiva Vicent, S.

2016-08-01

In the recent years, due, among others, the slowing down of the nuclear industry, investment in the development and validation of CFD codes, applied specifically to the problems of the nuclear industry has been seriously hampered. Thus the International Benchmark Exercise (IBE) sponsored by the OECD/NEA have been fundamental to analyze the use of CFD codes in the nuclear industry, because although these codes are mature in many fields, still exist doubts about them in critical aspects of thermohydraulic calculations, even in single-phase scenarios. The Polytechnic University of Valencia (UPV) and the Universitat Jaume I (UJI), sponsored by the Nuclear Safety Council (CSN), have actively participated in all benchmark's proposed by NEA, as in the expert meetings,. In this paper, a summary of participation in the various IBE will be held, describing the benchmark itself, the CFD model created for it, and the main conclusions. (Author)

Benchmarking van verkeersveiligheid : een inventarisatie en aanbevelingen voor de opzet van verkeersveiligheidsbenchmarks in Nederland.

NARCIS (Netherlands)

Aarts, L.T. & Bax, C.A.

2014-01-01

Benchmarking road safety; Stocktaking and recommendations for the development of road safety benchmarks in The Netherlands. Road safety policy has been decentralized in the Netherlands, giving regional and local governments a greater responsibility for road safety improvement in their jurisdiction.

Benchmark experiments to test plutonium and stainless steel cross sections. Topical report

International Nuclear Information System (INIS)

Jenquin, U.P.; Bierman, S.R.

1978-06-01

The Nuclear Regulatory Commission (NRC) commissioned Battelle, Pacific Northwest Laboratory (PNL) to ascertain the accuracy of the neutron cross sections for the isotopes of plutonium and the constituents of stainless steel and determine if improvements can be made in application to criticality safety analysis. NRC's particular area of interest is in the transportation of light-water reactor spent fuel assemblies. The project was divided into two tasks. The first task was to define a set of integral experimental measurements (benchmarks). The second task is to use these benchmarks in neutronics calculations such that the accuracy of ENDF/B-IV plutonium and stainless steel cross sections can be assessed. The results of the first task are given in this report. A set of integral experiments most pertinent to testing the cross sections has been identified and the code input data for calculating each experiment has been developed

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.

Developing integrated benchmarks for DOE performance measurement

Energy Technology Data Exchange (ETDEWEB)

Barancik, J.I.; Kramer, C.F.; Thode, Jr. H.C.

1992-09-30

The objectives of this task were to describe and evaluate selected existing sources of information on occupational safety and health with emphasis on hazard and exposure assessment, abatement, training, reporting, and control identifying for exposure and outcome in preparation for developing DOE performance benchmarks. Existing resources and methodologies were assessed for their potential use as practical performance benchmarks. Strengths and limitations of current data resources were identified. Guidelines were outlined for developing new or improved performance factors, which then could become the basis for selecting performance benchmarks. Data bases for non-DOE comparison populations were identified so that DOE performance could be assessed relative to non-DOE occupational and industrial groups. Systems approaches were described which can be used to link hazards and exposure, event occurrence, and adverse outcome factors, as needed to generate valid, reliable, and predictive performance benchmarks. Data bases were identified which contain information relevant to one or more performance assessment categories . A list of 72 potential performance benchmarks was prepared to illustrate the kinds of information that can be produced through a benchmark development program. Current information resources which may be used to develop potential performance benchmarks are limited. There is need to develop an occupational safety and health information and data system in DOE, which is capable of incorporating demonstrated and documented performance benchmarks prior to, or concurrent with the development of hardware and software. A key to the success of this systems approach is rigorous development and demonstration of performance benchmark equivalents to users of such data before system hardware and software commitments are institutionalized.

Immunotoxicity of perfluorinated alkylates: calculation of benchmark doses based on serum concentrations in children

DEFF Research Database (Denmark)

Grandjean, Philippe; Budtz-Joergensen, Esben

2013-01-01

BACKGROUND: Immune suppression may be a critical effect associated with exposure to perfluorinated compounds (PFCs), as indicated by recent data on vaccine antibody responses in children. Therefore, this information may be crucial when deciding on exposure limits. METHODS: Results obtained from...... follow-up of a Faroese birth cohort were used. Serum-PFC concentrations were measured at age 5 years, and serum antibody concentrations against tetanus and diphtheria toxoids were obtained at ages 7 years. Benchmark dose results were calculated in terms of serum concentrations for 431 children...

Researches on nuclear criticality safety evaluation

Energy Technology Data Exchange (ETDEWEB)

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

2003-10-01

For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

Researches on nuclear criticality safety evaluation

International Nuclear Information System (INIS)

Okuno, Hiroshi; Suyama, Kenya; Nomura, Yasushi

2003-01-01

For criticality safety evaluation of burnup fuel, the general-purpose burnup calculation code, SWAT, was revised, and its precision was confirmed through comparison with other results from OECD/NEA's burnup credit benchmarks. Effect by replacing the evaluated nuclear data from JENDL-3.2 to ENDF/B-VI and JEF-2.2 was also studied. Correction factors were derived for conservative evaluation of nuclide concentrations obtained with the simplified burnup code ORIGEN2.1. The critical masses of curium were calculated and evaluated for nuclear criticality safety management of minor actinides. (author)

Russian system of computerized analysis for licensing at atomic industry (SCALA) and its validation on ICSBEP handbook data and on some burnup calculations

International Nuclear Information System (INIS)

Ivanova, T.; Polyakov, A.; Saraeva, T.; Tsiboulia, A.

2001-01-01

Validation of criticality calculations using SCALA was performed using data presented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. This paper contains the results of statistical analysis of discrepancies between calculated and benchmark-model k eff and conclusions about uncertainties of criticality prediction for different types of multiplying systems following from this analysis. (authors)

RB reactor benchmark cores

International Nuclear Information System (INIS)

Pesic, M.

1998-01-01

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

Benchmarking quantum mechanical calculations with experimental NMR chemical shifts of 2-HADNT

Science.gov (United States)

Liu, Yuemin; Junk, Thomas; Liu, Yucheng; Tzeng, Nianfeng; Perkins, Richard

2015-04-01

In this study, both GIAO-DFT and GIAO-MP2 calculations of nuclear magnetic resonance (NMR) spectra were benchmarked with experimental chemical shifts. The experimental chemical shifts were determined experimentally for carbon-13 (C-13) of seven carbon atoms for the TNT degradation product 2-hydroxylamino-4,6-dinitrotoluene (2-HADNT). Quantum mechanics GIAO calculations were implemented using Becke-3-Lee-Yang-Parr (B3LYP) and other six hybrid DFT methods (Becke-1-Lee-Yang-Parr (B1LYP), Becke-half-and-half-Lee-Yang-Parr (BH and HLYP), Cohen-Handy-3-Lee-Yang-Parr (O3LYP), Coulomb-attenuating-B3LYP (CAM-B3LYP), modified-Perdew-Wang-91-Lee-Yang-Parr (mPW1LYP), and Xu-3-Lee-Yang-Parr (X3LYP)) which use the same correlation functional LYP. Calculation results showed that the GIAO-MP2 method gives the most accurate chemical shift values, and O3LYP method provides the best prediction of chemical shifts among the B3LYP and other five DFT methods. Three types of atomic partial charges, Mulliken (MK), electrostatic potential (ESP), and natural bond orbital (NBO), were also calculated using MP2/aug-cc-pVDZ method. A reasonable correlation was discovered between NBO partial charges and experimental chemical shifts of carbon-13 (C-13).

Integral parameters for the Godiva benchmark calculated by using theoretical and adjusted fission spectra of 235U

International Nuclear Information System (INIS)

Caldeira, A.D.

1987-05-01

The theoretical and adjusted Watt spectrum representations for 235 U are used as weighting functions to calculate K eff and θ f 28 /θ f 25 for the benchmark Godiva. The results obtained show that the values of K eff and θ f 28 /θ f 25 are not affected by spectrum form change. (author) [pt

RIA Fuel Codes Benchmark - Volume 1

International Nuclear Information System (INIS)

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

2013-01-01

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

Dynamic benchmarking of simulation codes

International Nuclear Information System (INIS)

Henry, R.E.; Paik, C.Y.; Hauser, G.M.

1996-01-01

output includes a plot of the MAAP calculation and the plant data. For the large integral experiments, a major part, but not all of the MAAP code is needed. These use an experiment specific benchmark routine that includes all of the information and boundary conditions for performing the calculation, as well as the information of which parts of MAAP are unnecessary and can be 'bypassed'. Lastly, the separate effects tests only require a few MAAP routines. These are exercised through their own specific benchmark routine that includes the experiment specific information and boundary conditions. This benchmark routine calls the appropriate MAAP routines from the source code, performs the calculations, including integration where necessary and provide the comparison between the MAAP calculation and the experimental observations. (author)

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

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.

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.

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)

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

Vver-1000 Mox core computational benchmark

International Nuclear Information System (INIS)

2006-01-01

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

Benefits of the delta K of depletion benchmarks for burnup credit validation

International Nuclear Information System (INIS)

Lancaster, D.; Machiels, A.

2012-01-01

Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO 2 critical experiments to determine the criticality safety limits on the neutron multiplication factor, k eff . The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

Parametric Criticality Safety Calculations for Arrays of TRU Waste Containers

Energy Technology Data Exchange (ETDEWEB)

Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-10-26

The Nuclear Criticality Safety Division (NCSD) has performed criticality safety calculations for finite and infinite arrays of transuranic (TRU) waste containers. The results of these analyses may be applied in any technical area onsite (e.g., TA-54, TA-55, etc.), as long as the assumptions herein are met. These calculations are designed to update the existing reference calculations for waste arrays documented in Reference 1, in order to meet current guidance on calculational methodology.

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

The lead cooled fast reactor benchmark Brest-300: analysis with sensitivity method

International Nuclear Information System (INIS)

Smirnov, V.; Orlov, V.; Mourogov, A.; Lecarpentier, D.; Ivanova, T.

2005-01-01

Lead cooled fast neutrons reactor is one of the most interesting candidates for the development of atomic energy. BREST-300 is a 300 MWe lead cooled fast reactor developed by the NIKIET (Russia) with a deterministic safety approach which aims to exclude reactivity margins greater than the delayed neutron fraction. The development of innovative reactors (lead coolant, nitride fuel...) and fuel cycles with new constraints such as cycle closure or actinide burning, requires new technologies and new nuclear data. In this connection, the tool and neutron data used for the calculational analysis of reactor characteristics requires thorough validation. NIKIET developed a reactor benchmark fitting of design type calculational tools (including neutron data). In the frame of technical exchanges between NIKIET and EDF (France), results of this benchmark calculation concerning the principal parameters of fuel evolution and safety parameters has been inter-compared, in order to estimate the uncertainties and validate the codes for calculations of this new kind of reactors. Different codes and cross-sections data have been used, and sensitivity studies have been performed to understand and quantify the uncertainties sources.The comparison of results shows that the difference on k eff value between ERANOS code with ERALIB1 library and the reference is of the same order of magnitude than the delayed neutron fraction. On the other hand, the discrepancy is more than twice bigger if JEF2.2 library is used with ERANOS. Analysis of discrepancies in calculation results reveals that the main effect is provided by the difference of nuclear data, namely U 238 , Pu 239 fission and capture cross sections and lead inelastic cross sections

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)

Calculational study for criticality safety data of fissionable actinides

International Nuclear Information System (INIS)

Nojiri, Ichiro; Fukasaku, Yasuhiro.

1997-01-01

This study has been carried out to obtain basic criticality safety characteristics of minor actinides nuclides. Criticality safety data of minor actinides nuclides have been surveyed through public literatures. Critical mass of seven nuclides, Np-237, Am-241, Am-242m, Am-243, Cm-243, Cm-244 and Cm-245, have been calculated by using two code systems of criticality safety analysis, SCALE-4 and MCNP4A, under some material and reflector conditions. Some applicable cross-section libraries have been used for each code systems. Calculated data have been compared with each other and with published data. The results of this comparison shows that there is no discrepancy within the computational codes and the calculated data is strongly depend on the cross-section library. (author)

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

International Nuclear Information System (INIS)

Hazama, T.

2009-07-01

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

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.)

Measurement, Standards, and Peer Benchmarking: One Hospital's Journey.

Science.gov (United States)

Martin, Brian S; Arbore, Mark

2016-04-01

Peer-to-peer benchmarking is an important component of rapid-cycle performance improvement in patient safety and quality-improvement efforts. Institutions should carefully examine critical success factors before engagement in peer-to-peer benchmarking in order to maximize growth and change opportunities. Solutions for Patient Safety has proven to be a high-yield engagement for Children's Hospital of Pittsburgh of University of Pittsburgh Medical Center, with measureable improvement in both organizational process and culture. Copyright © 2016 Elsevier Inc. All rights reserved.

Dose Rate Experiment at JET for Benchmarking the Calculation Direct One Step Method

International Nuclear Information System (INIS)

Angelone, M.; Petrizzi, L.; Pillon, M.; Villari, R.; Popovichev, S.

2006-01-01

Neutrons produced by D-D and D-T plasmas induce the activation of tokamak materials and of components. The development of reliable methods to assess dose rates is a key issue for maintenance and operating nuclear machines, in normal and off-normal conditions. In the frame of the EFDA Fusion Technology work programme, a computational tool based upon MCNP Monte Carlo code has been developed to predict the dose rate after shutdown: it is called Direct One Step Method (D1S). The D1S is an innovative approach in which the decay gammas are coupled to the neutrons as in the prompt case and they are transported in one single step in the same run. Benchmarking of this new tool with experimental data taken in a complex geometry like that of a tokamak is a fundamental step to test the reliability of the D1S method. A dedicated benchmark experiment was proposed for the 2005-2006 experimental campaign of JET. Two irradiation positions have been selected for the benchmark: one inner position inside the vessel, not far from the plasma, called the 2 upper irradiation end (IE2), where neutron fluence is relatively high. The second position is just outside a vertical port in an external position (EX). Here the neutron flux is lower and the dose rate to be measured is not very far from the residual background. Passive detectors are used for in-vessel measurements: the high sensitivity Thermo Luminescent Dosimeters (TLDs) GR-200A (natural LiF), which ensure measurements down to environmental dose level. An active detector of Geiger-Muller (GM) type is used for out of vessel dose rate measurement. Before their use the detectors were calibrated in a secondary gamma-ray standard (Cs-137 and Co-60) facility in term of air-kerma. The background measurement was carried-out in the period July -September 2005 in the outside position EX using the GM tube and in September 2005 inside the vacuum vessel using TLD detectors located in the 2 Upper irradiation end IE2. In the present work

Accident consequence calculations for project W-058 safety analysis

International Nuclear Information System (INIS)

Van Keuren, J.C.

1997-01-01

This document describes the calculations performed to determine the accident consequences for the W-058 safety analysis. Project W-058 is the replacement cross site transfer system (RCSTS), which is designed to transort liquid waste between the 200 W and 200 E areas. Calculations for RCSTS safety analyses used the same methods as the calculations for the Tank Waste Remediation System (TWRS) Basis for Interim Operation (BIO) and its supporting calculation notes. Revised analyses were performed for the spray and pool leak accidents since the RCSTS flows and pressures differ from those assumed in the TWRS BIO. Revision 1 of the document incorporates review comments

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)

Uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. Final report, February 16, 1990--December 31, 1994

International Nuclear Information System (INIS)

Busch, R.D.

1995-01-01

Dr. Robert Busch of the Department of Chemical and Nuclear Engineering was the principal investigator on this project with technical direction provided by the staff in the Nuclear Criticality Safety Group at Los Alamos. During the period of the contract, he had a number of graduate and undergraduate students working on subtasks. The objective of this work was to develop information on uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. During the first year of this project, most of the work was focused on setting up the SUN SPARC-1 Workstation and acquiring the literature which described the critical experiments. By august 1990, the Workstation was operational with the current version of TWODANT loaded on the system. MCNP, version 4 tape was made available from Los Alamos late in 1990. Various documents were acquired which provide the initial descriptions of the critical experiments under consideration as benchmarks. The next four years were spent working on various benchmark projects. A number of publications and presentations were made on this material. These are briefly discussed in this report

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)

Performance assessment of new neutron cross section libraries using MCNP code and some critical benchmarks

International Nuclear Information System (INIS)

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

2007-01-01

Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm [fr

Validation of calculational methods for nuclear criticality safety - approved 1975

International Nuclear Information System (INIS)

Anon.

1977-01-01

The American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, N16.1-1975, states in 4.2.5: In the absence of directly applicable experimental measurements, the limits may be derived from calculations made by a method shown to be valid by comparison with experimental data, provided sufficient allowances are made for uncertainties in the data and in the calculations. There are many methods of calculation which vary widely in basis and form. Each has its place in the broad spectrum of problems encountered in the nuclear criticality safety field; however, the general procedure to be followed in establishing validity is common to all. The standard states the requirements for establishing the validity and area(s) of applicability of any calculational method used in assessing nuclear criticality safety

Kowledge-based dynamic network safety calculations. Wissensbasierte dynamische Netzsicherheitsberechnungen

Energy Technology Data Exchange (ETDEWEB)

Kulicke, B [Inst. fuer Hochspannungstechnik und Starkstromanlagen, Berlin (Germany); Schlegel, S [Inst. fuer Hochspannungstechnik und Starkstromanlagen, Berlin (Germany)

1993-06-28

An important part of network operation management is the estimation and maintenance of the security of supply. So far the control personnel has only been supported by static network analyses and safety calculations. The authors describe an expert system, which is coupled to a real time simulation program on a transputer basis, for dynamic network safety calculations. They also introduce the system concept and the most important functions of the expert system. (orig.)

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)

RISKIND verification and benchmark comparisons

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

RISKIND verification and benchmark comparisons

International Nuclear Information System (INIS)

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

1997-08-01

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

Evaluation of neutron thermalization parameters and benchmark reactor calculations using a synthetic scattering function for molecular gases

International Nuclear Information System (INIS)

Gillete, V.H.; Patino, N.E.; Granada, J.E.; Mayer, R.E.

1988-01-01

Using a synthetic scattering function which describes the interaction of neutrons with molecular gases we provide analytical expressions for zero-and first-order scattering kernels, σ 0 (E 0 →E), σ 1 (E 0 →E), and total cross section σ 0 (E 0 ). Based on these quantities, we have performed calculations of thermalization parameters and transport coefficients for H 2 O, D 2 O, C 6 H 6 and (CH 2 ) n at room temperature. Comparasion of such values with available experimental data and other calculations is satisfactory. We also generated nuclear data libraries for H 2 O with 47 thermal groups at 300K and performed some benchmark calculations ( 235 U, 239 Pu, PWR cell and typical APWR cell); the resulting reactivities are compared with experimental data and ENDF/B-IV calculations. (author) [pt

Quantum computing applied to calculations of molecular energies: CH2 benchmark.

Science.gov (United States)

Veis, Libor; Pittner, Jiří

2010-11-21

Quantum computers are appealing for their ability to solve some tasks much faster than their classical counterparts. It was shown in [Aspuru-Guzik et al., Science 309, 1704 (2005)] that they, if available, would be able to perform the full configuration interaction (FCI) energy calculations with a polynomial scaling. This is in contrast to conventional computers where FCI scales exponentially. We have developed a code for simulation of quantum computers and implemented our version of the quantum FCI algorithm. We provide a detailed description of this algorithm and the results of the assessment of its performance on the four lowest lying electronic states of CH(2) molecule. This molecule was chosen as a benchmark, since its two lowest lying (1)A(1) states exhibit a multireference character at the equilibrium geometry. It has been shown that with a suitably chosen initial state of the quantum register, one is able to achieve the probability amplification regime of the iterative phase estimation algorithm even in this case.

Monte Carlo benchmarking: Validation and progress

International Nuclear Information System (INIS)

Sala, P.

2010-01-01

Document available in abstract form only. Full text of publication follows: Calculational tools for radiation shielding at accelerators are faced with new challenges from the present and next generations of particle accelerators. All the details of particle production and transport play a role when dealing with huge power facilities, therapeutic ion beams, radioactive beams and so on. Besides the traditional calculations required for shielding, activation predictions have become an increasingly critical component. Comparison and benchmarking with experimental data is obviously mandatory in order to build up confidence in the computing tools, and to assess their reliability and limitations. Thin target particle production data are often the best tools for understanding the predictive power of individual interaction models and improving their performances. Complex benchmarks (e.g. thick target data, deep penetration, etc.) are invaluable in assessing the overall performances of calculational tools when all ingredients are put at work together. A review of the validation procedures of Monte Carlo tools will be presented with practical and real life examples. The interconnections among benchmarks, model development and impact on shielding calculations will be highlighted. (authors)

PHEBUS-FPTO Benchmark calculations

International Nuclear Information System (INIS)

Shepherd, I.; Ball, A.; Trambauer, K.; Barbero, F.; Olivar Dominguez, F.; Herranz, L.; Biasi, L.; Fermandjian, J.; Hocke, K.

1991-01-01

This report summarizes a set of pre-test predictions made for the first Phebus-FP test, FPT-O. There were many different calculations, performed by various organizations and they represent the first attempt to calculate the whole experimental sequence, from bundle to containment. Quantitative agreement between the various calculations was not good but the particular models in the code responsible for disagreements were mostly identified. A consensus view was formed as to how the test would proceed. It was found that a successful execution of the test will require a different operating procedure than had been assumed here. Critical areas which require close attention are the need to devize a strategy for the power and flow in the bundle that takes account of uncertainties in the modelling and the shroud conductivity and the necessity to develop a reliable method to achieve the desired thermalhydraulic conditions in the containment

Preparation of data for criticality safety evaluation of nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Okuno, Hiroshi; Suyama, Kenya; Yoshiyama, Hiroshi; Tonoike, Kotaro; Miyoshi, Yoshinori

2005-01-01

Nuclear Criticality Safety Handbook/Data Collection, Version 2 was submitted to the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan as a contract report. In this presentation paper, its outline and related recent works are presented. After an introduction in Chapter 1, useful information to obtain the atomic number densities was collected in Chapter 2. The nuclear characteristic parameters for 11 nuclear fuels were provided in Chapter 3, and subcriticality judgment graphs were given in Chapter 4. The estimated critical and estimated lower-limit critical values were supplied for the 11 nuclear fuels as results of calculations by using the Japanese Evaluated Nuclear Data Library, JENDL-3.2, and the continuous energy Monte Carlo neutron transport code MVP in Chapter 5. The results of benchmark calculations based on the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook were summarized into six fuel categories in Chapter 6. As for recent works, subcriticality judgment graphs for U-SiO 2 and Pu-SiO 2 were obtained. Benchmark calculations were made with the combination of the latest version of the library JENDL-3.3 and MVP code for a series of STACY experiments and the estimated critical and estimated lower-limit critical values of 10 wt%-enriched uranium nitrate solutions were calculated. (author)

NPP Krsko core calculations to improve operational safety

International Nuclear Information System (INIS)

Ivekovic, I.; Grgic, D.; Nemec, T.

2007-01-01

Calculation tools and methodology used to perform independent calculations of cumulative influence of different changes related to fuel and core operation of NPP Krsko were described. Some examples of steady state and transient results are used to illustrate potential improvements to understanding and reviewing plant safety. (author)

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

IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

International Nuclear Information System (INIS)

Sartori, Enrico; Blair Briggs, J.

2008-01-01

1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

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)

Benchmarking and validation of a Geant4-SHADOW Monte Carlo simulation for dose calculations in microbeam radiation therapy.

Science.gov (United States)

Cornelius, Iwan; Guatelli, Susanna; Fournier, Pauline; Crosbie, Jeffrey C; Sanchez Del Rio, Manuel; Bräuer-Krisch, Elke; Rosenfeld, Anatoly; Lerch, Michael

2014-05-01

Microbeam radiation therapy (MRT) is a synchrotron-based radiotherapy modality that uses high-intensity beams of spatially fractionated radiation to treat tumours. The rapid evolution of MRT towards clinical trials demands accurate treatment planning systems (TPS), as well as independent tools for the verification of TPS calculated dose distributions in order to ensure patient safety and treatment efficacy. Monte Carlo computer simulation represents the most accurate method of dose calculation in patient geometries and is best suited for the purpose of TPS verification. A Monte Carlo model of the ID17 biomedical beamline at the European Synchrotron Radiation Facility has been developed, including recent modifications, using the Geant4 Monte Carlo toolkit interfaced with the SHADOW X-ray optics and ray-tracing libraries. The code was benchmarked by simulating dose profiles in water-equivalent phantoms subject to irradiation by broad-beam (without spatial fractionation) and microbeam (with spatial fractionation) fields, and comparing against those calculated with a previous model of the beamline developed using the PENELOPE code. Validation against additional experimental dose profiles in water-equivalent phantoms subject to broad-beam irradiation was also performed. Good agreement between codes was observed, with the exception of out-of-field doses and toward the field edge for larger field sizes. Microbeam results showed good agreement between both codes and experimental results within uncertainties. Results of the experimental validation showed agreement for different beamline configurations. The asymmetry in the out-of-field dose profiles due to polarization effects was also investigated, yielding important information for the treatment planning process in MRT. This work represents an important step in the development of a Monte Carlo-based independent verification tool for treatment planning in MRT.

Revaluering benchmarking - A topical theme for the construction industry

DEFF Research Database (Denmark)

Rasmussen, Grane Mikael Gregaard

2011-01-01

and questioning the concept objectively. This paper addresses the underlying nature of benchmarking, and accounts for the importance of focusing attention on the sociological impacts benchmarking has in organizations. To understand these sociological impacts, benchmarking research needs to transcend...... the perception of benchmarking systems as secondary and derivative and instead studying benchmarking as constitutive of social relations and as irredeemably social phenomena. I have attempted to do so in this paper by treating benchmarking using a calculative practice perspective, and describing how...

Evaluation of cross sections of 56Fe up to 3 GeV and integral benchmark calculation for thick target yield

International Nuclear Information System (INIS)

Yoshizawa, Nobuaki; Meigo, Shin-ichiro

2001-01-01

The neutron and proton cross sections of 56 Fe were evaluated up to 3 GeV. JENDL High Energy File of 56 Fe were developed for use in transport calculation. For neutrons, the high-energy data are merged with JENDL3.3-file. Integral benchmark calculations for thick target neutron yields (TTY) for 113 MeV and 256 MeV proton bombardment of Fe targets were performed using the evaluated libraries. Calculated TTY neutron spectra were compared with experimental data. For 113 MeV, calculated TTY at 7.5 degree underestimated in the emitted neutron energy range above 10 MeV. For 256 MeV, calculated TTY well agree with experimental data except below 10 MeV. (author)

Proposal on the accelerator driven molten-salt reactor (ATW-concept) benchmark calculation (stage-1 without an external neutron sources)

International Nuclear Information System (INIS)

Svarny, J.; Mikolas, P.

1999-01-01

The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark will be not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (Authors)

Sensitivity and Uncertainty Analysis of IAEA CRP HTGR Benchmark Using McCARD

International Nuclear Information System (INIS)

Jang, Sang Hoon; Shim, Hyung Jin

2016-01-01

The benchmark consists of 4 phases starting from the local standalone modeling (Phase I) to the safety calculation of coupled system with transient situation (Phase IV). As a preliminary study of UAM on HTGR, this paper covers the exercise 1 and 2 of Phase I which defines the unit cell and lattice geometry of MHTGR-350 (General Atomics). The objective of these exercises is to quantify the uncertainty of the multiplication factor induced by perturbing nuclear data as well as to analyze the specific features of HTGR such as double heterogeneity and self-shielding treatment. The uncertainty quantification of IAEA CRP HTGR UAM benchmarks were conducted using first-order AWP method in McCARD. Uncertainty of the multiplication factor was estimated only for the microscopic cross section perturbation. To reduce the computation time and memory shortage, recently implemented uncertainty analysis module in MC wielandt calculation was adjusted. The covariance data of cross section was generated by NJOY/ERRORR module with ENDF/B-VII.1. The numerical result was compared with evaluation result of DeCART/MUSAD code system developed by KAERI. IAEA CRP HTGR UAM benchmark problems were analyzed using McCARD. The numerical results were compared with Serpent for eigenvalue calculation and DeCART/MUSAD for S/U analysis. In eigenvalue calculation, inconsistencies were found in the result with ENDF/B-VII.1 cross section library and it was found to be the effect of thermal scattering data of graphite. As to S/U analysis, McCARD results matched well with DeCART/MUSAD, but showed some discrepancy in 238U capture regarding implicit uncertainty.

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

Benchmark Calculation of Radial Expectation Value for Confined Hydrogen-Like Atoms and Isotropic Harmonic Oscillators

International Nuclear Information System (INIS)

Yu, Rong Mei; Zan, Li Rong; Jiao, Li Guang; Ho, Yew Kam

2017-01-01

Spatially confined atoms have been extensively investigated to model atomic systems in extreme pressures. For the simplest hydrogen-like atoms and isotropic harmonic oscillators, numerous physical quantities have been established with very high accuracy. However, the expectation value of which is of practical importance in many applications has significant discrepancies among calculations by different methods. In this work we employed the basis expansion method with cut-off Slater-type orbitals to investigate these two confined systems. Accurate values for several low-lying bound states were obtained by carefully examining the convergence with respect to the size of basis. A scaling law for was derived and it is used to verify the accuracy of numerical results. Comparison with other calculations show that the present results establish benchmark values for this quantity, which may be useful in future studies. (author)

SU-E-J-30: Benchmark Image-Based TCP Calculation for Evaluation of PTV Margins for Lung SBRT Patients

Energy Technology Data Exchange (ETDEWEB)

Li, M [Wayne State Univeristy, Detroit, MI (United States); Chetty, I [Henry Ford Health System, Detroit, MI (United States); Zhong, H [Henry Ford Hospital System, Detroit, MI (United States)

2014-06-01

Purpose: Tumor control probability (TCP) calculated with accumulated radiation doses may help design appropriate treatment margins. Image registration errors, however, may compromise the calculated TCP. The purpose of this study is to develop benchmark CT images to quantify registration-induced errors in the accumulated doses and their corresponding TCP. Methods: 4DCT images were registered from end-inhale (EI) to end-exhale (EE) using a “demons” algorithm. The demons DVFs were corrected by an FEM model to get realistic deformation fields. The FEM DVFs were used to warp the EI images to create the FEM-simulated images. The two images combined with the FEM DVF formed a benchmark model. Maximum intensity projection (MIP) images, created from the EI and simulated images, were used to develop IMRT plans. Two plans with 3 and 5 mm margins were developed for each patient. With these plans, radiation doses were recalculated on the simulated images and warped back to the EI images using the FEM DVFs to get the accumulated doses. The Elastix software was used to register the FEM-simulated images to the EI images. TCPs calculated with the Elastix-accumulated doses were compared with those generated by the FEM to get the TCP error of the Elastix registrations. Results: For six lung patients, the mean Elastix registration error ranged from 0.93 to 1.98 mm. Their relative dose errors in PTV were between 0.28% and 6.8% for 3mm margin plans, and between 0.29% and 6.3% for 5mm-margin plans. As the PTV margin reduced from 5 to 3 mm, the mean TCP error of the Elastix-reconstructed doses increased from 2.0% to 2.9%, and the mean NTCP errors decreased from 1.2% to 1.1%. Conclusion: Patient-specific benchmark images can be used to evaluate the impact of registration errors on the computed TCPs, and may help select appropriate PTV margins for lung SBRT patients.

SU-E-J-30: Benchmark Image-Based TCP Calculation for Evaluation of PTV Margins for Lung SBRT Patients

International Nuclear Information System (INIS)

Li, M; Chetty, I; Zhong, H

2014-01-01

Purpose: Tumor control probability (TCP) calculated with accumulated radiation doses may help design appropriate treatment margins. Image registration errors, however, may compromise the calculated TCP. The purpose of this study is to develop benchmark CT images to quantify registration-induced errors in the accumulated doses and their corresponding TCP. Methods: 4DCT images were registered from end-inhale (EI) to end-exhale (EE) using a “demons” algorithm. The demons DVFs were corrected by an FEM model to get realistic deformation fields. The FEM DVFs were used to warp the EI images to create the FEM-simulated images. The two images combined with the FEM DVF formed a benchmark model. Maximum intensity projection (MIP) images, created from the EI and simulated images, were used to develop IMRT plans. Two plans with 3 and 5 mm margins were developed for each patient. With these plans, radiation doses were recalculated on the simulated images and warped back to the EI images using the FEM DVFs to get the accumulated doses. The Elastix software was used to register the FEM-simulated images to the EI images. TCPs calculated with the Elastix-accumulated doses were compared with those generated by the FEM to get the TCP error of the Elastix registrations. Results: For six lung patients, the mean Elastix registration error ranged from 0.93 to 1.98 mm. Their relative dose errors in PTV were between 0.28% and 6.8% for 3mm margin plans, and between 0.29% and 6.3% for 5mm-margin plans. As the PTV margin reduced from 5 to 3 mm, the mean TCP error of the Elastix-reconstructed doses increased from 2.0% to 2.9%, and the mean NTCP errors decreased from 1.2% to 1.1%. Conclusion: Patient-specific benchmark images can be used to evaluate the impact of registration errors on the computed TCPs, and may help select appropriate PTV margins for lung SBRT patients

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)

A national standard for psychosocial safety climate (PSC): PSC 41 as the benchmark for low risk of job strain and depressive symptoms.

Science.gov (United States)

Bailey, Tessa S; Dollard, Maureen F; Richards, Penny A M

2015-01-01

Despite decades of research from around the world now permeating occupational health and safety (OHS) legislation and guidelines, there remains a lack of tools to guide practice. Our main goal was to establish benchmark levels of psychosocial safety climate (PSC) that would signify risk of job strain (jobs with high demands and low control) and depression in organizations. First, to justify our focus on PSC, using interview data from Australian employees matched at 2 time points 12 months apart (n = 1081), we verified PSC as a significant leading predictor of job strain and in turn depression. Next, using 2 additional data sets (n = 2097 and n = 1043) we determined benchmarks of organizational PSC (range 12-60) for low-risk (PSC at 41 or above) and high-risk (PSC at 37 or below) of employee job strain and depressive symptoms. Finally, using the newly created benchmarks we estimated the population attributable risk (PAR) and found that improving PSC in organizations to above 37 could reduce 14% of job strain and 16% of depressive symptoms in the working population. The results provide national standards that organizations and regulatory agencies can utilize to promote safer working environments and lower the risk of harm to employee mental health. PsycINFO Database Record (c) 2014 APA, all rights reserved.

Benchmarking statewide trauma mortality using Agency for Healthcare Research and Quality's patient safety indicators.

Science.gov (United States)

Ang, Darwin; McKenney, Mark; Norwood, Scott; Kurek, Stanley; Kimbrell, Brian; Liu, Huazhi; Ziglar, Michele; Hurst, James

2015-09-01

Improving clinical outcomes of trauma patients is a challenging problem at a statewide level, particularly if data from the state's registry are not publicly available. Promotion of optimal care throughout the state is not possible unless clinical benchmarks are available for comparison. Using publicly available administrative data from the State Department of Health and the Agency for Healthcare Research and Quality (AHRQ) patient safety indicators (PSIs), we sought to create a statewide method for benchmarking trauma mortality and at the same time also identifying a pattern of unique complications that have an independent influence on mortality. Data for this study were obtained from State of Florida Agency for Health Care Administration. Adult trauma patients were identified as having International Classification of Disease ninth edition codes defined by the state. Multivariate logistic regression was used to create a predictive inpatient expected mortality model. The expected value of PSIs was created using the multivariate model and their beta coefficients provided by the AHRQ. Case-mix adjusted mortality results were reported as observed to expected (O/E) ratios to examine mortality, PSIs, failure to prevent complications, and failure to rescue from death. There were 50,596 trauma patients evaluated during the study period. The overall fit of the expected mortality model was very strong at a c-statistic of 0.93. Twelve of 25 trauma centers had O/E ratios benchmarking method that screens at risk trauma centers in the state for higher than expected mortality. Stratifying mortality based on failure to prevent PSIs may identify areas of needed improvement at a statewide level. Copyright © 2015 Elsevier Inc. All rights reserved.

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...

The spent fuel safety experiment

International Nuclear Information System (INIS)

Harmms, G.A.; Davis, F.J.; Ford, J.T.

1995-01-01

The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

Energy Technology Data Exchange (ETDEWEB)

Kahler, A. [Los Alamos National Laboratory (LANL); Macfarlane, R E [Los Alamos National Laboratory (LANL); Mosteller, R D [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Frankle, S C [Los Alamos National Laboratory (LANL); Chadwick, M. B. [Los Alamos National Laboratory (LANL); Mcknight, R D [Argonne National Laboratory (ANL); Lell, R M [Argonne National Laboratory (ANL); Palmiotti, G [Idaho National Laboratory (INL); Hiruta, h [Idaho National Laboratory (INL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Arcilla, r [Brookhaven National Laboratory (BNL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Sublet, J C [Culham Science Center, Abington, UK; Trkov, A. [Jozef Stefan Institute, Slovenia; Trumbull, T H [Knolls Atomic Power Laboratory; Dunn, Michael E [ORNL

2011-01-01

The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unrnoderated and uranium reflected (235)U and (239)Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as (236)U; (238,242)Pu and (241,243)Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical

Criticality Benchmark Results Using Various MCNP Data Libraries

International Nuclear Information System (INIS)

Frankle, Stephanie C.

1999-01-01

A suite of 86 criticality benchmarks has been recently implemented in MCNPtrademark as part of the nuclear data validation effort. These benchmarks have been run using two sets of MCNP continuous-energy neutron data: ENDF/B-VI based data through Release 2 (ENDF60) and the ENDF/B-V based data. New evaluations were completed for ENDF/B-VI for a number of the important nuclides such as the isotopes of H, Be, C, N, O, Fe, Ni, 235,238 U, 237 Np, and 239,240 Pu. When examining the results of these calculations for the five manor categories of 233 U, intermediate-enriched 235 U (IEU), highly enriched 235 U (HEU), 239 Pu, and mixed metal assembles, we find the following: (1) The new evaluations for 9 Be, 12 C, and 14 N show no net effect on k eff ; (2) There is a consistent decrease in k eff for all of the solution assemblies for ENDF/B-VI due to 1 H and 16 O, moving k eff further from the benchmark value for uranium solutions and closer to the benchmark value for plutonium solutions; (3) k eff decreased for the ENDF/B-VI Fe isotopic data, moving the calculated k eff further from the benchmark value; (4) k eff decreased for the ENDF/B-VI Ni isotopic data, moving the calculated k eff closer to the benchmark value; (5) The W data remained unchanged and tended to calculate slightly higher than the benchmark values; (6) For metal uranium systems, the ENDF/B-VI data for 235 U tends to decrease k eff while the 238 U data tends to increase k eff . The net result depends on the energy spectrum and material specifications for the particular assembly; (7) For more intermediate-energy systems, the changes in the 235,238 U evaluations tend to increase k eff . For the mixed graphite and normal uranium-reflected assembly, a large increase in k eff due to changes in the 238 U evaluation moved the calculated k eff much closer to the benchmark value. (8) There is little change in k eff for the uranium solutions due to the new 235,238 U evaluations; and (9) There is little change in k eff

Reactor group constants and benchmark test

Energy Technology Data Exchange (ETDEWEB)

Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-08-01

The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

Three-dimensional RAMA fluence methodology benchmarking

International Nuclear Information System (INIS)

Baker, S. P.; Carter, R. G.; Watkins, K. E.; Jones, D. B.

2004-01-01

This paper describes the benchmarking of the RAMA Fluence Methodology software, that has been performed in accordance with U. S. Nuclear Regulatory Commission Regulatory Guide 1.190. The RAMA Fluence Methodology has been developed by TransWare Enterprises Inc. through funding provided by the Electric Power Research Inst., Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP). The purpose of the software is to provide an accurate method for calculating neutron fluence in BWR pressure vessels and internal components. The Methodology incorporates a three-dimensional deterministic transport solution with flexible arbitrary geometry representation of reactor system components, previously available only with Monte Carlo solution techniques. Benchmarking was performed on measurements obtained from three standard benchmark problems which include the Pool Criticality Assembly (PCA), VENUS-3, and H. B. Robinson Unit 2 benchmarks, and on flux wire measurements obtained from two BWR nuclear plants. The calculated to measured (C/M) ratios range from 0.93 to 1.04 demonstrating the accuracy of the RAMA Fluence Methodology in predicting neutron flux, fluence, and dosimetry activation. (authors)

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

Analysis of a molten salt reactor benchmark

International Nuclear Information System (INIS)

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

2013-01-01

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

Spent reactor fuel benchmark composition data for code validation

International Nuclear Information System (INIS)

Bierman, S.R.

1991-09-01

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

Base data for looking-up tables of calculation errors in JACS code system

International Nuclear Information System (INIS)

Murazaki, Minoru; Okuno, Hiroshi

1999-03-01

The report intends to clarify the base data for the looking-up tables of calculation errors cited in 'Nuclear Criticality Safety Handbook'. The tables were obtained by classifying the benchmarks made by JACS code system, and there are two kinds: One kind is for fuel systems in general geometry with a reflected and another kind is for fuel systems specific to simple geometry with a reflector. Benchmark systems were further categorized into eight groups according to the fuel configuration: homogeneous or heterogeneous; and fuel kind: uranium, plutonium and their mixtures, etc. The base data for fuel systems in general geometry with a reflected are summarized in this report for the first time. The base data for fuel systems in simple geometry with a reflector were summarized in a technical report published in 1987. However, the data in a group named homogeneous low-enriched uranium were further selected out later by the working group for making the Nuclear Criticality Safety Handbook. This report includes the selection. As a project has been organized by OECD/NEA for evaluation of criticality safety benchmark experiments, the results are also described. (author)

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; J. Blair Briggs; Jim Gulliford; Ian Hill

2014-10-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

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

Recommended nuclear criticality safety experiments in support of the safe transportation of fissile material

International Nuclear Information System (INIS)

Tollefson, D.A.; Elliott, E.P.; Dyer, H.R.; Thompson, S.A.

1993-01-01

Validation of computer codes and nuclear data (cross-section) libraries using benchmark quality critical (or certain subcritical) experiments is an essential part of a nuclear criticality safety evaluation. The validation results establish the credibility of the calculational tools for use in evaluating a particular application. Validation of the calculational tools is addressed in several American National Standards Institute/American Nuclear Society (ANSI/ANS) standards, with ANSI/ANS-8.1 being the most relevant. Documentation of the validation is a required part of all safety analyses involving significant quantities of fissile materials. In the case of transportation of fissile materials, the safety analysis report for packaging (SARP) must contain a thorough discussion of benchmark experiments, detailing how the experiments relate to the significant packaging and contents materials (fissile, moderating, neutron absorbing) within the package. The experiments recommended in this paper are needed to address certain areas related to transportation of unirradiated fissile materials in drum-type containers (packagings) for which current data are inadequate or are lacking

Spin-orbit couplings within the equation-of-motion coupled-cluster framework: Theory, implementation, and benchmark calculations

Energy Technology Data Exchange (ETDEWEB)

Epifanovsky, Evgeny [Department of Chemistry, University of Southern California, Los Angeles, California 90089-0482 (United States); Department of Chemistry, University of California, Berkeley, California 94720 (United States); Q-Chem Inc., 6601 Owens Drive, Suite 105, Pleasanton, California 94588 (United States); Klein, Kerstin; Gauss, Jürgen [Institut für Physikalische Chemie, Universität Mainz, D-55099 Mainz (Germany); Stopkowicz, Stella [Department of Chemistry, Centre for Theoretical and Computational Chemistry, University of Oslo, N-0315 Oslo (Norway); Krylov, Anna I. [Department of Chemistry, University of Southern California, Los Angeles, California 90089-0482 (United States)

2015-08-14

We present a formalism and an implementation for calculating spin-orbit couplings (SOCs) within the EOM-CCSD (equation-of-motion coupled-cluster with single and double substitutions) approach. The following variants of EOM-CCSD are considered: EOM-CCSD for excitation energies (EOM-EE-CCSD), EOM-CCSD with spin-flip (EOM-SF-CCSD), EOM-CCSD for ionization potentials (EOM-IP-CCSD) and electron attachment (EOM-EA-CCSD). We employ a perturbative approach in which the SOCs are computed as matrix elements of the respective part of the Breit-Pauli Hamiltonian using zeroth-order non-relativistic wave functions. We follow the expectation-value approach rather than the response-theory formulation for property calculations. Both the full two-electron treatment and the mean-field approximation (a partial account of the two-electron contributions) have been implemented and benchmarked using several small molecules containing elements up to the fourth row of the periodic table. The benchmark results show the excellent performance of the perturbative treatment and the mean-field approximation. When used with an appropriate basis set, the errors with respect to experiment are below 5% for the considered examples. The findings regarding basis-set requirements are in agreement with previous studies. The impact of different correlation treatment in zeroth-order wave functions is analyzed. Overall, the EOM-IP-CCSD, EOM-EA-CCSD, EOM-EE-CCSD, and EOM-SF-CCSD wave functions yield SOCs that agree well with each other (and with the experimental values when available). Using an EOM-CCSD approach that provides a more balanced description of the target states yields more accurate results.

Working group on VVER safety analysis - report of the 2010 meeting

International Nuclear Information System (INIS)

Kliem, S.

2010-01-01

The AER Working Group D on WWER reactor safety analysis held its nineteenth meeting in Pisa, Italy, during the period 15-16 April, 2010. The meeting was hosted by the San Piero a Grado Nuclear Research Group of the University of Pisa and was held in conjunction with the second workshop on the OECD/NEA Benchmark for the Kalinin-3 WWER-1000 NPP and the fourth workshop on the OECD Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of LWRs. Altogether 12 participants attended the meeting of the working group D, 8 from AER member organizations and 4 guests from non-member organization. The co-ordinator of the working group, Mr. S. Kliem, served as chairman of the meeting. The meeting started with a general information exchange about the recent activities in the participating organizations. The given presentations and the discussions can be attributed to the following topics:-Code validation and benchmarking including the calculation of the OECD/NEA Benchmark for the Kalinin-3 WWER-1000 NPP;-Safety analyses and code developments;-Future activities A list of the participants and a list of the handouts distributed at the meeting are attached to the report. The corresponding PDF-files can be obtained from the chairman. (Author)

Benchmark calculations on fluid coupled co-axial cylinders typical to LMFBR structures

International Nuclear Information System (INIS)

Dostal, M.; Descleve, P.; Gantenbein, F.; Lazzeri, L.

1983-01-01

This paper describes a joint effort promoted and funded by the Commission of European Community under the umbrella of Fast Reactor Co-ordinating Committee and working group on Codes and Standards No. 2 with the purpose to test several programs currently used for dynamic analysis of fluid-coupled structures. The scope of the benchmark calculations is limited to beam type modes of vibration, small displacement of the structures and small pressure variation such as encountered in seismic or flow induced vibration problems. Five computer codes were used: ANSYS, AQUAMODE, NOVAX, MIAS/SAP4 and ZERO where each program employs a different structural-fluid formulation. The calculations were performed for four different geometrical configurations of concentric cylinders where the effect of gap size, water level, and support conditions were considered. The analytical work was accompanied by experiments carried out on a purpose-built rig. The test rig consisted of two concentric cylinders independently supported on flexible cantilevers. A geometrical simplicity and attention in the rig design to eliminate the structural coupling between the cylinders lead to unambiguous test results. Only the beam natural frequencies, in phase and out of phase were measured. The comparison of different analytical methods and experimental results is presented and discussed. The degree of agreement varied between very good and unacceptable. (orig./GL)

Handbook of critical experiments benchmarks

International Nuclear Information System (INIS)

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

1978-03-01

Data from critical experiments have been collected together for use as benchmarks in evaluating calculational techniques and nuclear data. These benchmarks have been selected from the numerous experiments performed on homogeneous plutonium systems. No attempt has been made to reproduce all of the data that exists. The primary objective in the collection of these data is to present representative experimental data defined in a concise, standardized format that can easily be translated into computer code input

Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark. Re-connection of an isolated loop

Energy Technology Data Exchange (ETDEWEB)

Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

2016-09-15

The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center

International Nuclear Information System (INIS)

Podlazov, L. N.

1998-01-01

Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions

Nuclear critical safety analysis for UX-30 transport of freight package

International Nuclear Information System (INIS)

Quan Yanhui; Zhou Qi; Yin Shenggui

2014-01-01

The nuclear critical safety analysis and evaluation for UX-30 transport freight package in the natural condition and accident condition were carried out with MONK-9A code and MCNP code. Firstly, the critical benchmark experiment data of public in international were selected, and the deflection and subcritical limiting value with MONK-9A code and MCNP code in calculating same material form were validated and confirmed. Secondly, the neutron efficiency multiplication factors in the natural condition and accident condition were calculated and analyzed, and the safety in transport process was evaluated by taking conservative suppose of nuclear critical safety. The calculation results show that the max value of k eff for UX-30 transport freight package is less than the subcritical limiting value, and the UX-30 transport freight package is in the state of subcritical safety. Moreover, the critical safety index (CSI) for UX-30 package can define zero based on the definition of critical safety index. (authors)

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

International Nuclear Information System (INIS)

2000-01-01

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

A BENCHMARKING ANALYSIS FOR FIVE RADIONUCLIDE VADOSE ZONE MODELS (CHAIN, MULTIMED_DP, FECTUZ, HYDRUS, AND CHAIN 2D) IN SOIL SCREENING LEVEL CALCULATIONS

Science.gov (United States)

Five radionuclide vadose zone models with different degrees of complexity (CHAIN, MULTIMED_DP, FECTUZ, HYDRUS, and CHAIN 2D) were selected for use in soil screening level (SSL) calculations. A benchmarking analysis between the models was conducted for a radionuclide (99Tc) rele...

Prismatic Core Coupled Transient Benchmark

International Nuclear Information System (INIS)

Ortensi, J.; Pope, M.A.; Strydom, G.; Sen, R.S.; DeHart, M.D.; Gougar, H.D.; Ellis, C.; Baxter, A.; Seker, V.; Downar, T.J.; Vierow, K.; Ivanov, K.

2011-01-01

The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivated the development of more accurate and efficient tools for the design and safety evaluations of the PMR. In addition to the work invested in new methods, it is essential to develop appropriate benchmarks to verify and validate the new methods in computer codes. The purpose of this benchmark is to establish a well-defined problem, based on a common given set of data, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events. The benchmark-working group is currently seeking OECD/NEA sponsorship. This benchmark is being pursued and is heavily based on the success of the PBMR-400 exercise.

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

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

Energy Technology Data Exchange (ETDEWEB)

Kahler, A.C.; Herman, M.; Kahler,A.C.; MacFarlane,R.E.; Mosteller,R.D.; Kiedrowski,B.C.; Frankle,S.C.; Chadwick,M.B.; McKnight,R.D.; Lell,R.M.; Palmiotti,G.; Hiruta,H.; Herman,M.; Arcilla,R.; Mughabghab,S.F.; Sublet,J.C.; Trkov,A.; Trumbull,T.H.; Dunn,M.

2011-12-01

The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also

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)

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

International Nuclear Information System (INIS)

Liu Guisheng

1999-01-01

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

Finite-element solutions of the AER-2 rod ejection benchmark by CRONOS

International Nuclear Information System (INIS)

Kolev, N.P.; Lenain, R.; Fedon-Magnaud, C.

2001-01-01

The finite-element option in CRONOS was used to analyse the AER-2 rod-ejection benchmark for WWER-440. The objective is to obtain spatially converged solutions by means of node subdivision and approximation refinement. This paper presents the first phase of analysis dealing with the initial and just-ejected states used for calculation of the initial reactivity. Fine-mesh and extrapolated to zero mesh size solutions were obtained and verified by comparison to MAG code solutions. These differences provide potential for large deviations in the transient results and deserve further attention in reactor safety analysis (Authors)

International Benchmark on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume II: Benchmark Results of Phase I: Void Distribution

International Nuclear Information System (INIS)

Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

2016-03-01

This report summarised the first phase of the Nuclear Energy Agency (NEA) and the US Nuclear Regulatory Commission Benchmark based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of void distribution models in participants' codes. This phase was composed of four exercises; Exercise 1: steady-state single sub-channel benchmark, Exercise 2: steady-state rod bundle benchmark, Exercise 3: transient rod bundle benchmark and Exercise 4: a pressure drop benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both Boiling Water Reactors (BWRs) and Pressurised Water Reactors (PWRs). These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Twenty-one institutions from nine countries participated in this benchmark. Seventeen different computer codes were used in Exercises 1, 2, 3 and 4. Among the computer codes were porous media, sub-channel, systems thermal-hydraulic code and Computational Fluid Dynamics (CFD) codes. It was observed that the codes tended to overpredict the thermal equilibrium quality at lower elevations and under predict it at higher elevations. There was also a tendency to overpredict void fraction at lower elevations and underpredict it at high elevations for the bundle test cases. The overprediction of void fraction at low elevations is likely caused by the x-ray densitometer measurement method used. Under sub-cooled boiling conditions, the voids accumulate at heated surfaces (and are therefore not seen in the centre of the sub-channel, where the measurements are being taken), so the experimentally-determined void fractions will be lower than the actual void fraction. Some of the best

Radiation shielding calculation using MCNP

International Nuclear Information System (INIS)

Masukawa, Fumihiro

2001-01-01

To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

Influence of the ab initio n–d cross sections in the critical heavy-water benchmarks

International Nuclear Information System (INIS)

Morillon, B.; Lazauskas, R.; Carbonell, J.

2013-01-01

Highlights: ► We solve the three nucleon problem using different NN potential (MT, AV18 and INOY) to calculate the Neutron–deuteron cross sections. ► These cross sections are compared to the existing experimental data and to international libraries. ► We describe the different sets of heavy water benchmarks for which the Monte Carlo simulations have been performed including our new Neutron–deuteron cross sections. ► The results obtained by the ab initio INOY potential have been compared with the calculations based on the international library cross sections and are found to be of the same quality. - Abstract: The n–d elastic and breakup cross sections are computed by solving the three-body Faddeev equations for realistic and semi-realistic nucleon–nucleon potentials. These cross sections are inserted in the Monte Carlo simulation of the nuclear processes considered in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results obtained using thes ab initio n–d cross sections are compared with those provided by the most renown international libraries

OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the First Workshop (V1000-CT1)

International Nuclear Information System (INIS)

2003-01-01

The first workshop for the VVER-1000 Coolant Transient Benchmark TT Benchmark was hosted by the Commissariat a l'Energie Atomique, Centre d'Etudes de Saclay, France. The V1000CT benchmark defines standard problems for validation of coupled three-dimensional (3-D) neutron-kinetics/system thermal-hydraulics codes for application to Soviet-designed VVER-1000 reactors using actual plant data without any scaling. The overall objective is to access computer codes used in the safety analysis of VVER power plants, specifically for their use in reactivity transient simulations in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 - simulation of the switching on of one main coolant pump (MCP) while the other three MCP are in operation, and V1000CT- 2 - calculation of coolant mixing tests and Main Steam Line Break (MSLB) scenario. Further background information on this benchmark can be found at the OECD/NEA benchmark web site . The purpose of the first workshop was to review the benchmark activities after the Starter Meeting held last year in Dresden, Germany: to discuss the participants' feedback and modifications introduced in the Benchmark Specifications on Phase 1; to present and to discuss modelling issues and preliminary results from the three exercises of Phase 1; to discuss the modelling issues of Exercise 1 of Phase 2; and to define work plan and schedule in order to complete the two phases

Validation of the criticality calculation for fuel elements using the Gamtec 2 - Keno 2 and 4

International Nuclear Information System (INIS)

Teixeira, M.C.C.; Andrade, M.C. de

1990-01-01

For criticality safety in the fabrication, storage and transportation of fuel assemblies, subcriticality analysis must be done. The calculations are performed at CDTN with the GAMTEC computer code, to homogenize the fuel assembly in order to create 16 group cross-section library, and with KENO code, for determining the multiplication factor. To validate the calculational method, suitable Benchmark experiments have been done. The results show that the calculational model overestimates kef when kef+ 2 σ was considered. (author) [pt

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)

JNC results of BFS-62-3A benchmark calculation (CRP: Phase 5)

International Nuclear Information System (INIS)

Ishikawa, M.

2004-01-01

The present work is the results of JNC, Japan, for the Phase 5 of IAEA CRP benchmark problem (BFS-62-3A critical experiment). Analytical Method of JNC is based on Nuclear Data Library JENDL-3.2; Group Constant Set JFS-3-J3.2R: 70-group, ABBN-type self-shielding factor table based on JENDL-3.2; Effective Cross-section - Current-weighted multigroup transport cross-section. Cell model for the BFS as-built tube and pellets was (Case 1) Homogeneous Model based on IPPE definition; (Case 2) Homogeneous atomic density equivalent to JNC's heterogeneous calculation only to cross-check the adjusted correction factors; (Case 3) Heterogeneous model based on JNC's evaluation, One-dimensional plate-stretch model with Tone's background cross-section method (CASUP code). Basic diffusion Calculation was done in 18-groups and three-dimensional Hex-Z model (by the CITATION code), with Isotropic diffusion coefficients (Case 1 and 2), and Benoist's anisotropic diffusion coefficients (Case 3). For sodium void reactivity, the exact perturbation theory was applied both to basic calculation and correction calculations, ultra-fine energy group correction - approx. 100,000 group constants below 50 keV, and ABBN-type 175 group constants with shielding factors above 50 keV. Transport theory and mesh size correction 18-group, was used for three-dimensional Hex-Z model (the MINIHEX code based on the S4-P0 transport method, which was developed by JNC. Effective delayed Neutron fraction in the reactivity scale was fixed at 0.00623 by IPPE evaluation. Analytical Results of criticality values and sodium void reactivity coefficient obtained by JNC are presented. JNC made a cross-check of the homogeneous model and the adjusted correction factors submitted by IPPE, and confirmed they are consistent. JNC standard system showed quite satisfactory analytical results for the criticality and the sodium void reactivity of BFS-62-3A experiment. JNC calculated the cross-section sensitivity coefficients of BFS

Utilization of the MCNP-3A code for criticality safety analysis

International Nuclear Information System (INIS)

Maragni, M.G.; Moreira, J.M.L.

1996-01-01

In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

The University of Pisa calculations for the Phase I of the OECD/NEA UAM Benchmark

International Nuclear Information System (INIS)

Ball, M.; Parisi, C.; D'Auria, F.

2009-01-01

In this paper we present the Univ. of Pisa preliminary results for the first exercise of the Phase I of the OECD/NEA Benchmark on the Uncertainty in Analysis and Modeling. The scope of exercise one is to address the uncertainties due to the basic nuclear data as well as the impact of processing the nuclear and covariance data, selection of multi-group structure and self-shielding treatment. DRAGON code and TSUNAMI code were employed, using the available covariance data matrix. The execution of DRAGON calculations required the use of ANGELO and LAMBDA codes for the extension of the covariance matrix from the original SCALE 44 group structure to DRAGON 69 group structure. The uncertainties for the main cross sections were evaluated and are presented here. (authors)

Atomic Energy Research benchmark activity

International Nuclear Information System (INIS)

Makai, M.

1998-01-01

The test problems utilized in the validation and verification process of computer programs in Atomic Energie Research are collected into one bunch. This is the first step towards issuing a volume in which tests for VVER are collected, along with reference solutions and a number of solutions. The benchmarks do not include the ZR-6 experiments because they have been published along with a number of comparisons in the Final reports of TIC. The present collection focuses on operational and mathematical benchmarks which cover almost the entire range of reaktor calculation. (Author)

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)

Gamma ray benchmark on the spent fuel shipping cask TN 12

International Nuclear Information System (INIS)

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

1983-05-01

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

Benchmark Analysis of Subcritical Noise Measurements on a Nickel-Reflected Plutonium Metal Sphere

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; Jesson Hutchinson

2009-09-01

Subcritical experiments using californium source-driven noise analysis (CSDNA) and Feynman variance-to-mean methods were performed with an alpha-phase plutonium sphere reflected by nickel shells, up to a maximum thickness of 7.62 cm. Both methods provide means of determining the subcritical multiplication of a system containing nuclear material. A benchmark analysis of the experiments was performed for inclusion in the 2010 edition of the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Benchmark models have been developed that represent these subcritical experiments. An analysis of the computed eigenvalues and the uncertainty in the experiment and methods was performed. The eigenvalues computed using the CSDNA method were very close to those calculated using MCNP5; however, computed eigenvalues are used in the analysis of the CSDNA method. Independent calculations using KENO-VI provided similar eigenvalues to those determined using the CSDNA method and MCNP5. A slight trend with increasing nickel-reflector thickness was seen when comparing MCNP5 and KENO-VI results. For the 1.27-cm-thick configuration the MCNP eigenvalue was approximately 300 pcm greater. The calculated KENO eigenvalue was about 300 pcm greater for the 7.62-cm-thick configuration. The calculated results were approximately the same for a 5-cm-thick shell. The eigenvalues determined using the Feynman method are up to approximately 2.5% lower than those determined using either the CSDNA method or the Monte Carlo codes. The uncertainty in the results from either method was not large enough to account for the bias between the two experimental methods. An ongoing investigation is being performed to assess what potential uncertainties and/or biases exist that have yet to be properly accounted for. The dominant uncertainty in the CSDNA analysis was the uncertainty in selecting a neutron cross-section library for performing the analysis of the data. The uncertainty in the

A PWR Thorium Pin Cell Burnup Benchmark

Energy Technology Data Exchange (ETDEWEB)

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

2000-05-01

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

New calculations for critical assemblies using MCNP4B

International Nuclear Information System (INIS)

Adams, A.A.; Frankle, S.C.; Little, R.C.

1997-07-01

A suite of 41 criticality benchmarks has been modeled using MCNP trademark (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes 233 U, 235 U, 238 U, and 239 Pu. The values of k eff for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI.2 evaluations for H, O, N, B, 235 U, 238 U, and 239 Pu can have a significant impact on the values of k eff . In addition to the integral quantity k eff , several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k eff . Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI.2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data

AER working group D on WWER safety analysis - report of the 2007 meeting

International Nuclear Information System (INIS)

Kliem, S.

2007-01-01

The AER working group D on WWER reactor safety analysis held its sixteenth meeting in Paris, France during the period 08-09 May 2007. The meeting was hosted by the CEA France. It followed the final workshop on the OECD/DOE/CEA WWER-1000 Coolant Transient Benchmark held at 07 May. Altogether 11 participants attend the meeting of the working group D, 7 from AER member organizations and 4 guests from non-member organizations. The co-ordinator of the working group, Mr. S. Kliem, served as chairman of the meeting. The meeting started with a general information exchange about the recent activities in the participating organizations. The given presentations and the discussions can be attributed to the following topics: -Code development and benchmarking for reactor dynamics applications; -Safety analysis methodology and results; -Future activities. New solutions for three different benchmarks were presented and discussed. These are the Second AER Dynamic Benchmark on control rod ejection at hot zero power (S. Kliem, FZD), the WWER-1000 Coolant Transient Benchmark (E. Syrjaelahti, VTT) and the stationary AER-FCM101 Benchmark considering a WWER-1000 reactor (C. Parisi, UniPisa). A. Kereszturi (AEKI) presented a statistical evaluation of the possibility to observe a fuel assembly mis loading event. The second presentation of E. Syrjaelahti was dedicated to the description how best-estimate coupled code calculations at VTT are supported by uncertainty and sensitivity analyses. K. Velkov (GRS) presented preliminary results of BIPR8KN/ATHLET calculations with a very detailed resolution of the calculation grid on the assessment of coolant mixing inside WWER-1000 assembly heads. Coolant mixing experiments at three different mixing test facilities, modeling different reactor types, were presented and compared by S. Kliem. A calculation study using the coupled code system KORSAR/GP on the consequences of the injection of a slug of un borated water into the reactor core was

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)

Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry

International Nuclear Information System (INIS)

Ait Abderrahim, H.; D'Hondt, P.; Oeyen, J.; Risch, P.; Bioux, P.

1993-09-01

The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The 'Concrete Benchmark' experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the 'Concrete Benchmark' experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs

BN-600 Phase III benchmark calculations

International Nuclear Information System (INIS)

Hill, R.N.; Grimm, K.N.

2002-01-01

Calculations for a Hexagonal-Z model of the BN-600 reactor with a partial mixed oxide loading, based on a joint IPPE/OBMK loading configuration that contained three uranium enrichment zones and one plutonium enrichment zone in the core, have been performed at ANL. Control-rod worths and reactivity feedback coefficients were calculated using both homogeneous and heterogeneous models. These values were calculated with either first-order perturbation theory methods (Triangle-Z geometry), nodal eigenvalue differences (Hexagonal-Z geometry), or Monte Carlo eigenvalue differences. Both spatially-dependent and region integrated values are shown

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.

Numerical methods: Analytical benchmarking in transport theory

International Nuclear Information System (INIS)

Ganapol, B.D.

1988-01-01

Numerical methods applied to reactor technology have reached a high degree of maturity. Certainly one- and two-dimensional neutron transport calculations have become routine, with several programs available on personal computer and the most widely used programs adapted to workstation and minicomputer computational environments. With the introduction of massive parallelism and as experience with multitasking increases, even more improvement in the development of transport algorithms can be expected. Benchmarking an algorithm is usually not a very pleasant experience for the code developer. Proper algorithmic verification by benchmarking involves the following considerations: (1) conservation of particles, (2) confirmation of intuitive physical behavior, and (3) reproduction of analytical benchmark results. By using today's computational advantages, new basic numerical methods have been developed that allow a wider class of benchmark problems to be considered

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)

Validation of new 240Pu cross section and covariance data via criticality calculation

International Nuclear Information System (INIS)

Kim, Do Heon; Gil, Choong-Sup; Kim, Hyeong Il; Lee, Young-Ouk; Leal, Luiz C.; Dunn, Michael E.

2011-01-01

Recent collaboration between KAERI and ORNL has completed an evaluation for 240 Pu neutron cross section with covariance data. The new 240 Pu cross section data has been validated through 28 criticality safety benchmark problems taken from the ICSBEP and/or CSEWG specifications with MCNP calculations. The calculation results based on the new evaluation have been compared with those based on recent evaluations such as ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. In addition, the new 240 Pu covariance data has been tested for some criticality benchmarks via the DANTSYS/SUSD3D-based nuclear data sensitivity and uncertainty analysis of k eff . The k eff uncertainty estimates by the new covariance data has been compared with those by JENDL-4.0, JENDL-3.3, and Low-Fidelity covariance data. (author)

Numerical benchmarking of SPEEDUP trademark against point kinetics solutions

International Nuclear Information System (INIS)

Gregory, M.V.

1993-02-01

SPEEDUP trademark is a state-of-the-art, dynamic, chemical process modeling package offered by Aspen Technology. In anticipation of new customers' needs for new analytical tools to support the site's waste management activities, SRTC has secured a multiple-user license to SPEEDUP trademark. In order to verify both the installation and mathematical correctness of the algorithms in SPEEDUP trademark, we have performed several numerical benchmarking calculations. These calculations are the first steps in establishing an on-site quality assurance pedigree for SPEEDUP trademark. The benchmark calculations consisted of SPEEDUP trademark Version 5.3L representations of five neutron kinetics benchmarks (each a mathematically stiff system of seven coupled ordinary differential equations), whose exact solutions are documented in the open literature. In all cases, SPEEDUP trademark solutions to be in excellent agreement with the reference solutions. A minor peculiarity in dealing with a non-existent discontinuity in the OPERATION section of the model made itself evident

A comparison of recent results from HONDO III with the JSME nuclear shipping cask benchmark calculations

International Nuclear Information System (INIS)

Key, S.W.

1985-01-01

The results of two calculations related to the impact response of spent nuclear fuel shipping casks are compared to the benchmark results reported in a recent study by the Japan Society of Mechanical Engineers Subcommittee on Structural Analysis of Nuclear Shipping Casks. Two idealized impacts are considered. The first calculation utilizes a right circular cylinder of lead subjected to a 9.0 m free fall onto a rigid target, while the second calculation utilizes a stainless steel clad cylinder of lead subjected to the same impact conditions. For the first problem, four calculations from graphical results presented in the original study have been singled out for comparison with HONDO III. The results from DYNA3D, STEALTH, PISCES, and ABAQUS are reproduced. In the second problem, the results from four separate computer programs in the original study, ABAQUS, ANSYS, MARC, and PISCES, are used and compared with HONDO III. The current version of HONDO III contains a fully automated implementation of the explicit-explicit partitioning procedure for the central difference method time integration which results in a reduction of computational effort by a factor in excess of 5. The results reported here further support the conclusion of the original study that the explicit time integration schemes with automated time incrementation are effective and efficient techniques for computing the transient dynamic response of nuclear fuel shipping casks subject to impact loading. (orig.)

Computational Benchmark Calculations Relevant to the Neutronic Design of the Spallation Neutron Source (SNS)

International Nuclear Information System (INIS)

Gallmeier, F.X.; Glasgow, D.C.; Jerde, E.A.; Johnson, J.O.; Yugo, J.J.

1999-01-01

The Spallation Neutron Source (SNS) will provide an intense source of low-energy neutrons for experimental use. The low-energy neutrons are produced by the interaction of a high-energy (1.0 GeV) proton beam on a mercury (Hg) target and slowed down in liquid hydrogen or light water moderators. Computer codes and computational techniques are being benchmarked against relevant experimental data to validate and verify the tools being used to predict the performance of the SNS. The LAHET Code System (LCS), which includes LAHET, HTAPE ad HMCNP (a modified version of MCNP version 3b), have been applied to the analysis of experiments that were conducted in the Alternating Gradient Synchrotron (AGS) facility at Brookhaven National Laboratory (BNL). In the AGS experiments, foils of various materials were placed around a mercury-filled stainless steel cylinder, which was bombarded with protons at 1.6 GeV. Neutrons created in the mercury target, activated the foils. Activities of the relevant isotopes were accurately measured and compared with calculated predictions. Measurements at BNL were provided in part by collaborating scientists from JAERI as part of the AGS Spallation Target Experiment (ASTE) collaboration. To date, calculations have shown good agreement with measurements

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

International Nuclear Information System (INIS)

Anno, J.; Poullot, G.

1995-01-01

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

Benchmarking road safety performances of countries.

NARCIS (Netherlands)

Wegman, F.C.M. & Oppe, S.

2014-01-01

In order to obtain political interest in road safety problems and to learn from other countries’ ‘good practices’, it is often helpful to compare one’s own safety situation with that of other countries. In a number of projects tools have been developed for such comparisons. These tools range from

Critical power prediction by CATHARE2 of the OECD/NRC BFBT benchmark

Energy Technology Data Exchange (ETDEWEB)

Lutsanych, Sergii, E-mail: s.lutsanych@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122, San Piero a Grado, Pisa (Italy); Sabotinov, Luben, E-mail: luben.sabotinov@irsn.fr [Institut for Radiological Protection and Nuclear Safety (IRSN), 31 avenue de la Division Leclerc, 92262 Fontenay-aux-Roses (France); D’Auria, Francesco, E-mail: francesco.dauria@dimnp.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122, San Piero a Grado, Pisa (Italy)

2015-03-15

Highlights: • We used CATHARE code to calculate the critical power exercises of the OECD/NRC BFBT benchmark. • We considered both steady-state and transient critical power tests of the benchmark. • We used both the 1D and 3D features of the CATHARE code to simulate the experiments. • Acceptable prediction of the critical power and its location in the bundle is obtained using appropriate modelling. - Abstract: This paper presents an application of the French best estimate thermal-hydraulic code CATHARE 2 to calculate the critical power and departure from nucleate boiling (DNB) exercises of the International OECD/NRC BWR Fuel Bundle Test (BFBT) benchmark. The assessment activity is performed comparing the code calculation results with available in the framework of the benchmark experimental data from Japanese Nuclear Power Engineering Corporation (NUPEC). Two-phase flow calculations on prediction of the critical power have been carried out both in steady state and transient cases, using one-dimensional and three-dimensional modelling. Results of the steady-state critical power tests calculation have shown the ability of CATHARE code to predict reasonably the critical power and its location, using appropriate modelling.

Resolution for the Loviisa benchmark problem

International Nuclear Information System (INIS)

Garcia, C.R.; Quintero, R.; Milian, D.

1992-01-01

In the present paper, the Loviisa benchmark problem for cycles 11 and 8, and reactor blocks 1 and 2 from Loviisa NPP, is calculated. This problem user law leakage reload patterns and was posed at the second thematic group of TIC meeting held in Rheinsberg GDR, march 1989. SPPS-1 coarse mesh code has been used for the calculations

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

International Nuclear Information System (INIS)

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

2015-01-01

a discussion of the spread of the eff k results. Following this, the evaluation of the end effect is accomplished starting with a discussion of the spread of the end effect results following from the eff k results. Then the functional dependence of the end effect on the control rod insertion depth is described by introducing and deriving model functions. After that the fission density results are evaluated by introducing and deriving fission density model functions describing the axial fission probability density for the different control rod insertion depths. Using these fission density model functions the fission probability content of the top end region of the active zone of the fuel assemblies is estimated. Predictions of the qualitative behaviour of the neutron multiplication factor and the end effect as a function of the control rod insertion depth were already made in the Phase II-C report and verified in this report which demonstrates the practical relevance of the relations established in the Phase II-C report. In addition, it turns out that parameters describing the average burn-up transformation characteristics of these relations play important roles in comparisons of the end effect model functions derived for the two Phase II-E axial burn-up profiles. Thus, the Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. The applicability of the knowledge gained from the results of all these three exercises to burn-up credit criticality safety design calculations is demonstrated

Application of PHEBUS results to benchmarking of nuclear plant safety codes

International Nuclear Information System (INIS)

Birchley, J.; Cripps, R.; Guentay, S.; Hosemann, J.P.

2001-01-01

The PHEBUS Fission Product project comprises six nuclear reactor severe accident simulations, using prototypic core materials and representative geometry and boundary conditions for the coolant loop and containment. The data thus produced are being used to benchmark the computer tools used for nuclear plant accident analysis to reduce the excessive conservatism typical for estimates of the radiological source term. A set of calculations has been carried out to simulate the results of experiment PHEBUS FPT-1 through each of its main stages, using computer models and methods analogous to those currently employed at PSI for assessments of Swiss nuclear plants. Good agreement for the core degradation and containment behaviour builds confidence in the models, while some open questions remain concerning some aspects of the release of fission products from the fuel, their transport and chemical speciation. Of potentially great importance to the reduction in source term estimates is the formation of the non-volatile species, silver iodide. Current investigations are focused on the uncertainty concerning fission product behaviour and the stability of silver iodide under irradiation. (author)

Application of an integrated PC-based neutronics code system to criticality safety

International Nuclear Information System (INIS)

Briggs, J.B.; Nigg, D.W.

1991-01-01

An integrated system of neutronics and radiation transport software suitable for operation in an IBM PC-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past four years. Four modules within the system are particularly useful for criticality safety applications. Using the neutronics portion of the integrated code system, effective neutron multiplication values (k eff values) have been calculated for a variety of benchmark critical experiments for metal systems (Plutonium and Uranium), Aqueous Systems (Plutonium and Uranium) and LWR fuel rod arrays. A description of the codes and methods used in the analysis and the results of the benchmark critical experiments are presented in this paper. In general, excellent agreement was found between calculated and experimental results. (Author)

Second benchmark problem for WIPP structural computations

International Nuclear Information System (INIS)

Krieg, R.D.; Morgan, H.S.; Hunter, T.O.

1980-12-01

This report describes the second benchmark problem for comparison of the structural codes used in the WIPP project. The first benchmark problem consisted of heated and unheated drifts at a depth of 790 m, whereas this problem considers a shallower level (650 m) more typical of the repository horizon. But more important, the first problem considered a homogeneous salt configuration, whereas this problem considers a configuration with 27 distinct geologic layers, including 10 clay layers - 4 of which are to be modeled as possible slip planes. The inclusion of layering introduces complications in structural and thermal calculations that were not present in the first benchmark problem. These additional complications will be handled differently by the various codes used to compute drift closure rates. This second benchmark problem will assess these codes by evaluating the treatment of these complications

Electron-helium S-wave model benchmark calculations. I. Single ionization and single excitation

Science.gov (United States)

Bartlett, Philip L.; Stelbovics, Andris T.

2010-02-01

A full four-body implementation of the propagating exterior complex scaling (PECS) method [J. Phys. B 37, L69 (2004)] is developed and applied to the electron-impact of helium in an S-wave model. Time-independent solutions to the Schrödinger equation are found numerically in coordinate space over a wide range of energies and used to evaluate total and differential cross sections for a complete set of three- and four-body processes with benchmark precision. With this model we demonstrate the suitability of the PECS method for the complete solution of the full electron-helium system. Here we detail the theoretical and computational development of the four-body PECS method and present results for three-body channels: single excitation and single ionization. Four-body cross sections are presented in the sequel to this article [Phys. Rev. A 81, 022716 (2010)]. The calculations reveal structure in the total and energy-differential single-ionization cross sections for excited-state targets that is due to interference from autoionization channels and is evident over a wide range of incident electron energies.

A Benchmarking Study of High Energy Carbon Ion Induced Neutron Using Several Monte Carlo Codes

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Oh, J. H.; Jung, N. S.; Lee, H. S. [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Shin, Y. S.; Kwon, D. Y.; Kim, Y. M. [Catholic Univ., Gyeongsan (Korea, Republic of); Oranj, L. Mokhtari [POSTECH, Pohang (Korea, Republic of)

2014-10-15

In this study, the benchmarking study was done for the representative particle interaction of the heavy ion accelerator, especially carbon-induced reaction. The secondary neutron is an important particle in the shielding analysis to define the source term and penetration ability of radiation fields. The performance of each Monte Carlo codes were verified for selected codes: MCNPX 2.7, PHITS 2.64 and FLUKA 2011.2b.6. For this benchmarking study, the experimental data of Kurosawa et al. in the SINBAD database of NEA was applied. The calculated results of the differential neutron yield produced from several materials irradiated by high energy carbon beam reproduced the experimental data well in small uncertainty. But the MCNPX results showed large discrepancy with experimental data, especially at the forward angle. The calculated results were lower a little than the experimental and it was clear in the cases of lower incident carbon energy, thinner target and forward angle. As expected, the influence of different model was found clearly at forward direction. In the shielding analysis, these characteristics of each Monte Carlo codes should be considered and utilized to determine the safety margin of a shield thickness.

A Benchmarking Study of High Energy Carbon Ion Induced Neutron Using Several Monte Carlo Codes

International Nuclear Information System (INIS)

Kim, D. H.; Oh, J. H.; Jung, N. S.; Lee, H. S.; Shin, Y. S.; Kwon, D. Y.; Kim, Y. M.; Oranj, L. Mokhtari

2014-01-01

In this study, the benchmarking study was done for the representative particle interaction of the heavy ion accelerator, especially carbon-induced reaction. The secondary neutron is an important particle in the shielding analysis to define the source term and penetration ability of radiation fields. The performance of each Monte Carlo codes were verified for selected codes: MCNPX 2.7, PHITS 2.64 and FLUKA 2011.2b.6. For this benchmarking study, the experimental data of Kurosawa et al. in the SINBAD database of NEA was applied. The calculated results of the differential neutron yield produced from several materials irradiated by high energy carbon beam reproduced the experimental data well in small uncertainty. But the MCNPX results showed large discrepancy with experimental data, especially at the forward angle. The calculated results were lower a little than the experimental and it was clear in the cases of lower incident carbon energy, thinner target and forward angle. As expected, the influence of different model was found clearly at forward direction. In the shielding analysis, these characteristics of each Monte Carlo codes should be considered and utilized to determine the safety margin of a shield thickness

Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry; Experience ``Benchmark beton`` pour la dosimetrie hors cuve dans les reacteurs a eau legere

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, H.; D`Hondt, P.; Oeyen, J.; Risch, P.; Bioux, P.

1993-09-01

The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The `Concrete Benchmark` experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the `Concrete Benchmark` experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs.

The Department of Energy nuclear criticality safety program

International Nuclear Information System (INIS)

Felty, J.R.

2004-01-01

This paper broadly covers key events and activities from which the Department of Energy Nuclear Criticality Safety Program (NCSP) evolved. The NCSP maintains fundamental infrastructure that supports operational criticality safety programs. This infrastructure includes continued development and maintenance of key calculational tools, differential and integral data measurements, benchmark compilation, development of training resources, hands-on training, and web-based systems to enhance information preservation and dissemination. The NCSP was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 97-2, Criticality Safety, and evolved from a predecessor program, the Nuclear Criticality Predictability Program, that was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 93-2, The Need for Critical Experiment Capability. This paper also discusses the role Dr. Sol Pearlstein played in helping the Department of Energy lay the foundation for a robust and enduring criticality safety infrastructure.

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)

Solution of the 'MIDICORE' WWER-1000 core periphery power distribution benchmark by KARATE and MCNP

International Nuclear Information System (INIS)

Temesvari, E.; Hegyi, G.; Hordosy, G.; Maraczy, C.

2011-01-01

The 'MIDICORE' WWER-1000 core periphery power distribution benchmark was proposed by Mr. Mikolas on the twentieth Symposium of AER in Finland in 2010. This MIDICORE benchmark is a two-dimensional calculation benchmark based on the WWER-1000 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of the benchmark is to test the pin by pin power distribution in selected fuel assemblies at the periphery of the WWER-1000 core. In this paper we present our results (k eff , integral fission power) calculated by MCNP and the KARATE code system in KFKI-AEKI and the comparison to the preliminary reference Monte Carlo calculation results made by NRI, Rez. (Authors)

Cluster monte carlo method for nuclear criticality safety calculation

International Nuclear Information System (INIS)

Pei Lucheng

1984-01-01

One of the most important applications of the Monte Carlo method is the calculation of the nuclear criticality safety. The fair source game problem was presented at almost the same time as the Monte Carlo method was applied to calculating the nuclear criticality safety. The source iteration cost may be reduced as much as possible or no need for any source iteration. This kind of problems all belongs to the fair source game prolems, among which, the optimal source game is without any source iteration. Although the single neutron Monte Carlo method solved the problem without the source iteration, there is still quite an apparent shortcoming in it, that is, it solves the problem without the source iteration only in the asymptotic sense. In this work, a new Monte Carlo method called the cluster Monte Carlo method is given to solve the problem further

Criticality safety analysis for mockup facility

International Nuclear Information System (INIS)

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

2000-03-01

Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO 2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum K eff is 0.28356 well below than the critical limit, K eff =0.95 at normal condition. In a hypothetical accidental condition, the maximum K eff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. K eff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the K eff increases as the water volume ratio increases. It is also revealed that the K eff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum K eff value is 0.93960 lower than the subcritical limit

Methodology for calculating guideline concentrations for safety shot sites

International Nuclear Information System (INIS)

1997-06-01

Residual plutonium (Pu), with trace quantities of depleted uranium (DU) or weapons grade uranium (WU), exists in surficial soils at the Nevada Test Site (NTS), Nellis Air Force Range (NAFR), and the Tonopah Test Range (TTR) as the result of the above-ground testing of nuclear weapons and special experiments involving the detonation of plutonium-bearing devices. The special experiments (referred to as safety shots) involving plutonium-bearing devices were conducted to study the behavior of Pu as it was being explosively compressed; ensure that the accidental detonation of the chemical explosive in a production weapon would not result in criticality; evaluate the ability of personnel to manage large-scale Pu dispersal accidents; and develop criteria for transportation and storage of nuclear weapons. These sites do not pose a health threat to either workers or the general public because they are under active institutional control. The DOE is committed to remediating the safety shot sites so that radiation exposure to the public, both now and in the future, will be maintained within the established limits and be as low as reasonably achievable. Remediation requires calculation of a guideline concentration for the Pu, U, and their decay products that are present in the surface soil. This document presents the methodology for calculating guideline concentrations of weapons grade plutonium, weapons grade uranium, and depleted uranium in surface soils at the safety shot sites. Emphasis is placed on obtaining site-specific data for use in calculating dose to potential residents from the residual soil contamination

Methodology for calculating guideline concentrations for safety shot sites

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

Residual plutonium (Pu), with trace quantities of depleted uranium (DU) or weapons grade uranium (WU), exists in surficial soils at the Nevada Test Site (NTS), Nellis Air Force Range (NAFR), and the Tonopah Test Range (TTR) as the result of the above-ground testing of nuclear weapons and special experiments involving the detonation of plutonium-bearing devices. The special experiments (referred to as safety shots) involving plutonium-bearing devices were conducted to study the behavior of Pu as it was being explosively compressed; ensure that the accidental detonation of the chemical explosive in a production weapon would not result in criticality; evaluate the ability of personnel to manage large-scale Pu dispersal accidents; and develop criteria for transportation and storage of nuclear weapons. These sites do not pose a health threat to either workers or the general public because they are under active institutional control. The DOE is committed to remediating the safety shot sites so that radiation exposure to the public, both now and in the future, will be maintained within the established limits and be as low as reasonably achievable. Remediation requires calculation of a guideline concentration for the Pu, U, and their decay products that are present in the surface soil. This document presents the methodology for calculating guideline concentrations of weapons grade plutonium, weapons grade uranium, and depleted uranium in surface soils at the safety shot sites. Emphasis is placed on obtaining site-specific data for use in calculating dose to potential residents from the residual soil contamination.

Validation of the EIR LWR calculation methods for criticality assessment of storage pools

International Nuclear Information System (INIS)

Grimm, P.; Paratte, J.M.

1986-11-01

The EIR code system for the calculation of light water reactors is presented and the methods used are briefly described. The application of the system to various types of critical experiments and benchmark problems proves its good accuracy, even for heterogeneous configurations containing strong neutron absorbers such as Boral. Since the multiplication factor k eff is normally somewhat overpredicted and the spread of the results is small, this code system is validated for the calculation of storage pools, taking into account a safety margins of 1.5% on k eff . (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)

LCEs for Naval Reactor Benchmark Calculations

International Nuclear Information System (INIS)

W.J. Anderson

1999-01-01

The purpose of this engineering calculation is to document the MCNP4B2LV evaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 22 different cases with varied design parameters. Some of these LCEs (10) are documented in existing references (Ref. 7.1 and 7.2), but were re-run for this calculation file using more neutron histories. The objective of this analysis is to quantify the MCNP4B2LV code system's ability to accurately calculate the effective neutron multiplication factor (k eff ) for various critical configurations. These LCE evaluations support the development and validation of the neutronics methodology used for criticality analyses involving Naval reactor spent nuclear fuel in a geologic repository

Criticality benchmark comparisons leading to cross-section upgrades

International Nuclear Information System (INIS)

Alesso, H.P.; Annese, C.E.; Heinrichs, D.P.; Lloyd, W.R.; Lent, E.M.

1993-01-01

For several years criticality benchmark calculations with COG. COG is a point-wise Monte Carlo code developed at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The principle consideration in developing COG was that the resulting calculation would be as accurate as the point-wise cross-sectional data, since no physics computational approximations were used. The objective of this paper is to report on COG results for criticality benchmark experiments in concert with MCNP comparisons which are resulting in corrections an upgrades to the point-wise ENDL cross-section data libraries. Benchmarking discrepancies reported here indicated difficulties in the Evaluated Nuclear Data Livermore (ENDL) cross-sections for U-238 at thermal neutron energy levels. This led to a re-evaluation and selection of the appropriate cross-section values from several cross-section sets available (ENDL, ENDF/B-V). Further cross-section upgrades anticipated

Nonparametric estimation of benchmark doses in environmental risk assessment

Science.gov (United States)

Piegorsch, Walter W.; Xiong, Hui; Bhattacharya, Rabi N.; Lin, Lizhen

2013-01-01

Summary An important statistical objective in environmental risk analysis is estimation of minimum exposure levels, called benchmark doses (BMDs), that induce a pre-specified benchmark response in a dose-response experiment. In such settings, representations of the risk are traditionally based on a parametric dose-response model. It is a well-known concern, however, that if the chosen parametric form is misspecified, inaccurate and possibly unsafe low-dose inferences can result. We apply a nonparametric approach for calculating benchmark doses, based on an isotonic regression method for dose-response estimation with quantal-response data (Bhattacharya and Kong, 2007). We determine the large-sample properties of the estimator, develop bootstrap-based confidence limits on the BMDs, and explore the confidence limits’ small-sample properties via a short simulation study. An example from cancer risk assessment illustrates the calculations. PMID:23914133

Links among available integral benchmarks and differential date evaluations, computational biases and uncertainties, and nuclear criticality safety biases on potential MOX production throughput

International Nuclear Information System (INIS)

Goluoglu, S.; Hopper, C.M.

2004-01-01

Through the use of Oak Ridge National Laboratory's recently developed and applied sensitivity and uncertainty computational analysis techniques, this paper presents the relevance and importance of available and needed integral benchmarks and differential data evaluations impacting potential MOX production throughput determinations relative to low-moderated MOX fuel blending operations. The relevance and importance in the availability of or need for critical experiment benchmarks and data evaluations are presented in terms of computational biases as influenced by computational and experimental sensitivities and uncertainties relative to selected MOX production powder blending processes. Recent developments for estimating the safe margins of subcriticality for assuring nuclear criticality safety for process approval are presented. In addition, the impact of the safe margins (due to computational biases and uncertainties) on potential MOX production throughput will also be presented. (author)

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

Food safety performance indicators to benchmark food safety output of food safety management systems.

Science.gov (United States)

Jacxsens, L; Uyttendaele, M; Devlieghere, F; Rovira, J; Gomez, S Oses; Luning, P A

2010-07-31

There is a need to measure the food safety performance in the agri-food chain without performing actual microbiological analysis. A food safety performance diagnosis, based on seven indicators and corresponding assessment grids have been developed and validated in nine European food businesses. Validation was conducted on the basis of an extensive microbiological assessment scheme (MAS). The assumption behind the food safety performance diagnosis is that food businesses which evaluate the performance of their food safety management system in a more structured way and according to very strict and specific criteria will have a better insight in their actual microbiological food safety performance, because food safety problems will be more systematically detected. The diagnosis can be a useful tool to have a first indication about the microbiological performance of a food safety management system present in a food business. Moreover, the diagnosis can be used in quantitative studies to get insight in the effect of interventions on sector or governmental level. Copyright 2010 Elsevier B.V. All rights reserved.

Reactor physics calculations on HTR type configurations

Energy Technology Data Exchange (ETDEWEB)

Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

1995-04-01

In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.).

Reactor physics calculations on HTR type configurations

International Nuclear Information System (INIS)

Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

1995-04-01

In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.)

Benchmarking in pathology: development of a benchmarking complexity unit and associated key performance indicators.

Science.gov (United States)

Neil, Amanda; Pfeffer, Sally; Burnett, Leslie

2013-01-01

This paper details the development of a new type of pathology laboratory productivity unit, the benchmarking complexity unit (BCU). The BCU provides a comparative index of laboratory efficiency, regardless of test mix. It also enables estimation of a measure of how much complex pathology a laboratory performs, and the identification of peer organisations for the purposes of comparison and benchmarking. The BCU is based on the theory that wage rates reflect productivity at the margin. A weighting factor for the ratio of medical to technical staff time was dynamically calculated based on actual participant site data. Given this weighting, a complexity value for each test, at each site, was calculated. The median complexity value (number of BCUs) for that test across all participating sites was taken as its complexity value for the Benchmarking in Pathology Program. The BCU allowed implementation of an unbiased comparison unit and test listing that was found to be a robust indicator of the relative complexity for each test. Employing the BCU data, a number of Key Performance Indicators (KPIs) were developed, including three that address comparative organisational complexity, analytical depth and performance efficiency, respectively. Peer groups were also established using the BCU combined with simple organisational and environmental metrics. The BCU has enabled productivity statistics to be compared between organisations. The BCU corrects for differences in test mix and workload complexity of different organisations and also allows for objective stratification into peer groups.

The benchmark testing of 9Be of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2002-01-01

CENDL-3, the latest version of China Evaluated Nuclear Data Library was finished. The data of 9 Be were updated, and distributed for benchmark analysis recently. The calculated results were presented, and compared with the experimental data and the results based on other evaluated nuclear data libraries. The results show that CENDL-3 is better than others for most benchmarks

Development and validation of a criticality calculation scheme based on French deterministic transport codes

International Nuclear Information System (INIS)

Santamarina, A.

1991-01-01

A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)

Development of Audit Calculation Methodology for RIA Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Joosuk; Kim, Gwanyoung; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

The interim criteria contain more stringent limits than previous ones. For example, pellet-to-cladding mechanical interaction(PCMI) was introduced as a new failure criteria. And both short-term (e.g. fuel-to coolant interaction, rod burst) and long-term(e.g., fuel rod ballooning, flow blockage) phenomena should be addressed for core coolability assurance. For dose calculations, transient-induced fission gas release has to be accounted additionally. Traditionally, the approved RIA analysis methodologies for licensing application are developed based on conservative approach. But newly introduced safety criteria tend to reduce the margins to the criteria. Thereby, licensees are trying to improve the margins by utilizing a less conservative approach. In this situation, to cope with this trend, a new audit calculation methodology needs to be developed. In this paper, the new methodology, which is currently under developing in KINS, was introduced. For the development of audit calculation methodology of RIA safety analysis based on the realistic evaluation approach, preliminary calculation by utilizing the best estimate code has been done on the initial core of APR1400. Followings are main conclusions. - With the assumption of single full-strength control rod ejection in HZP condition, rod failure due to PCMI is not predicted. - And coolability can be assured in view of entalphy and fuel melting. - But, rod failure due to DNBR is expected, and there is possibility of fuel failure at the rated power conditions also.

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

Comparative sensitivity study of some criticality safety benchmark experiments using JEFF-3.1.2, JEFF-3.2T and ENDF/B-VII.1

International Nuclear Information System (INIS)

Kooyman, Timothee; Messaoudia, Nadia

2014-01-01

A sensitivity study on a set of evaluated criticality benchmarks with two versions of the JEFF nuclear data library, namely JEFF-3.1.2 and JEFF-3.2T, and ENDF/B-VII.1 was performed using MNCP(X) 2.6.0. As these benchmarks serve to estimate the upper safety limit for criticality risk analysis at SCK.CEN the sensitivity of their results to nuclear data is an important parameter to asses. Several nuclides were identified as being responsible for an evident change in the effective multiplication factor k eff : 235 U, 239 Pu, 240 Pu, 54 Fe, 56 Fe, 57 Fe and 208 Pb. A high sensitivity was found to the fission cross-section of all the fissile material in the study. Additionally, a smaller sensitivity to inelastic and capture cross-section of 235 U and 240 Pu was also found. Sensitivity to the scattering law for non-fissile material was postulated. The biggest change in the k eff due to non-fissile material was due to 208 Pb evaluation (±700 pcm), followed by 56 Fe (±360 pcm) for both versions of the JEFF library. Changes due to 235 U (±300 pcm) and Pu isotopes (±120 pcm for 239 Pu and ±80 pcm for 240 Pu) were found only with JEFF-3.1.2. 238 U was found to have no effect on the k eff . Significant improvements were identified between the two versions of the JEFF library. No further differences were found between the JEFF-3.2T and the ENDF/B-VII.1 calculations involving 235 U or Pu. (authors)

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)

The aug-cc-pVnZ-F12 basis set family: Correlation consistent basis sets for explicitly correlated benchmark calculations on anions and noncovalent complexes.

Science.gov (United States)

Sylvetsky, Nitai; Kesharwani, Manoj K; Martin, Jan M L

2017-10-07

We have developed a new basis set family, denoted as aug-cc-pVnZ-F12 (or aVnZ-F12 for short), for explicitly correlated calculations. The sets included in this family were constructed by supplementing the corresponding cc-pVnZ-F12 sets with additional diffuse functions on the higher angular momenta (i.e., additional d-h functions on non-hydrogen atoms and p-g on hydrogen atoms), optimized for the MP2-F12 energy of the relevant atomic anions. The new basis sets have been benchmarked against electron affinities of the first- and second-row atoms, the W4-17 dataset of total atomization energies, the S66 dataset of noncovalent interactions, the Benchmark Energy and Geometry Data Base water cluster subset, and the WATER23 subset of the GMTKN24 and GMTKN30 benchmark suites. The aVnZ-F12 basis sets displayed excellent performance, not just for electron affinities but also for noncovalent interaction energies of neutral and anionic species. Appropriate CABSs (complementary auxiliary basis sets) were explored for the S66 noncovalent interaction benchmark: between similar-sized basis sets, CABSs were found to be more transferable than generally assumed.

Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2014-01-01

The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that - beyond the uncertainties of the geometry and the boundary conditions - it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.

Neutronic safety parameters of the BN-600 type reactor with hybrid core. Diffusion and transport approach. R-Z homogeneous media

International Nuclear Information System (INIS)

Cherny, V.; Danilytchev, A.; Korobeinikov, V.; Korobeinikova, L.; Stogov, V.

2000-01-01

The present paper includes the results of neutronic safety calculations of the BN-600 hybrid core benchmark problem. Results presented include: multiplication factors, Doppler coefficients, fuel and structure density coefficients, expansion coefficients, power distribution, beta-effective values, reaction rate distributions

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)

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)

Evaluation of PWR and BWR pin cell benchmark results

International Nuclear Information System (INIS)

Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J.; Hoogenboom, J.E.; Leege, P.F.A. de; Voet, J. van der; Verhagen, F.C.M.

1991-12-01

Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs

Evaluation of PWR and BWR pin cell benchmark results

Energy Technology Data Exchange (ETDEWEB)

Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

1991-12-01

Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

Evaluation of PWR and BWR pin cell benchmark results

Energy Technology Data Exchange (ETDEWEB)

Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

1991-12-01

Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-12-31

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

Deuteron cross section evaluation for safety and radioprotection calculations of IFMIF/EVEDA accelerator prototype

International Nuclear Information System (INIS)

Blideanu, Valentin; Garcia, Mauricio; Joyer, Philippe; Lopez, Daniel; Mayoral, Alicia; Ogando, Francisco; Ortiz, Felix; Sanz, Javier; Sauvan, Patrick

2011-01-01

In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is under construction in Japan. Interaction of these deuterons with matter will generate high levels of neutrons and induced activation, whose predicted yields depend strongly on the models used to calculate the different cross sections. A benchmark test was performed to validate these data for deuteron energies up to 20 MeV and to define a reasonable methodology for calculating the cross sections needed for EVEDA. Calculations were performed using the nuclear models included in MCNPX and PHITS, and the dedicated nuclear model code TALYS. Although the results obtained using TALYS (global parameters) or Monte Carlo codes disagree with experimental values, a solution is proposed to compute cross sections that are a good fit to experimental data. A consistent computational procedure is also suggested to improve both transport simulations/prompt dose and activation/residual dose calculations required for EVEDA.

Deuteron cross section evaluation for safety and radioprotection calculations of IFMIF/EVEDA accelerator prototype

Energy Technology Data Exchange (ETDEWEB)

Blideanu, Valentin [Commissariat a l' energie atomique CEA/IRFU, Centre de Saclay, 91191 Gif sur Yvette cedex (France); Garcia, Mauricio [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain); Joyer, Philippe, E-mail: philippe.joyer@cea.fr [Commissariat a l' energie atomique CEA/IRFU, Centre de Saclay, 91191 Gif sur Yvette cedex (France); Lopez, Daniel; Mayoral, Alicia; Ogando, Francisco [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain); Ortiz, Felix [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Sanz, Javier; Sauvan, Patrick [Universidad Nacional de Educacion a Distancia, UNED, Madrid (Spain); Instituto de Fusion Nuclear, UPM, Madrid (Spain)

2011-10-01

In the frame of IFMIF/EVEDA activities, a prototype accelerator delivering a high power deuteron beam is under construction in Japan. Interaction of these deuterons with matter will generate high levels of neutrons and induced activation, whose predicted yields depend strongly on the models used to calculate the different cross sections. A benchmark test was performed to validate these data for deuteron energies up to 20 MeV and to define a reasonable methodology for calculating the cross sections needed for EVEDA. Calculations were performed using the nuclear models included in MCNPX and PHITS, and the dedicated nuclear model code TALYS. Although the results obtained using TALYS (global parameters) or Monte Carlo codes disagree with experimental values, a solution is proposed to compute cross sections that are a good fit to experimental data. A consistent computational procedure is also suggested to improve both transport simulations/prompt dose and activation/residual dose calculations required for EVEDA.

Monte Carlo burnup simulation of the TAKAHAMA-3 benchmark experiment

International Nuclear Information System (INIS)

Dalle, Hugo M.

2009-01-01

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

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.

OECD/NEA BENCHMARK FOR UNCERTAINTY ANALYSIS IN MODELING (UAM FOR LWRS – SUMMARY AND DISCUSSION OF NEUTRONICS CASES (PHASE I

Directory of Open Access Journals (Sweden)

RYAN N. BRATTON

2014-06-01

Full Text Available A Nuclear Energy Agency (NEA, Organization for Economic Co-operation and Development (OECD benchmark for Uncertainty Analysis in Modeling (UAM is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the “Neutronics Phase”, which is devoted mostly to the propagation of nuclear data (cross-section uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: 238U radiative capture and inelastic scattering (n, n’ as well as the average number of neutrons released per fission event of 239Pu.

Preparation of a criticality benchmark based on experiments performed at the RA-6 reactor

International Nuclear Information System (INIS)

Bazzana, S.; Blaumann, H; Marquez Damian, J.I

2009-01-01

The operation and fuel management of a reactor uses neutronic modeling to predict its behavior in operational and accidental conditions. This modeling uses computational tools and nuclear data that must be contrasted against benchmark experiments to ensure its accuracy. These benchmarks have to be simple enough to be possible to model with the desired computer code and have quantified and bound uncertainties. The start-up of the RA-6 reactor, final stage of the conversion and renewal project, allowed us to obtain experimental results with fresh fuel. In this condition the material composition of the fuel elements is precisely known, which contributes to a more precise modeling of the critical condition. These experimental results are useful to evaluate the precision of the models used to design the core, based on U 3 Si 2 and cadmium wires as burnable poisons, for which no data was previously available. The analysis of this information can be used to validate models for the analysis of similar configurations, which is necessary to follow the operational history of the reactor and perform fuel management. The analysis of the results and the generation of the model were done following the methodology established by International Criticality Safety Benchmark Evaluation Project, which gathers and analyzes experimental data for critical systems. The results were very satisfactory resulting on a value for the multiplication factor of the model of 1.0000 ± 0.0044, and a calculated value of 0.9980 ± 0.0001 using MCNP 5 and ENDF/B-VI. The utilization of as-built dimensions and compositions, and the sensitivity analysis allowed us to review the design calculations and analyze their precision, accuracy and error compensation. [es

Estimation of ΔR/R values by benchmark study of the Mössbauer Isomer shifts for Ru, Os complexes using relativistic DFT calculations

Energy Technology Data Exchange (ETDEWEB)

Kaneko, Masashi [Japan Atomic Energy Agency, Nuclear Science and Engineering Center (Japan); Yasuhara, Hiroki; Miyashita, Sunao; Nakashima, Satoru, E-mail: snaka@hiroshima-u.ac.jp [Hiroshima University, Graduate School of Science (Japan)

2017-11-15

The present study applies all-electron relativistic DFT calculation with Douglas-Kroll-Hess (DKH) Hamiltonian to each ten sets of Ru and Os compounds. We perform the benchmark investigation of three density functionals (BP86, B3LYP and B2PLYP) using segmented all-electron relativistically contracted (SARC) basis set with the experimental Mössbauer isomer shifts for {sup 99}Ru and {sup 189}Os nuclides. Geometry optimizations at BP86 theory of level locate the structure in a local minimum. We calculate the contact density to the wavefunction obtained by a single point calculation. All functionals show the good linear correlation with experimental isomer shifts for both {sup 99}Ru and {sup 189}Os. Especially, B3LYP functional gives a stronger correlation compared to BP86 and B2PLYP functionals. The comparison of contact density between SARC and well-tempered basis set (WTBS) indicated that the numerical convergence of contact density cannot be obtained, but the reproducibility is less sensitive to the choice of basis set. We also estimate the values of ΔR/R, which is an important nuclear constant, for {sup 99}Ru and {sup 189}Os nuclides by using the benchmark results. The sign of the calculated ΔR/R values is consistent with the predicted data for {sup 99}Ru and {sup 189}Os. We obtain computationally the ΔR/R values of {sup 99}Ru and {sup 189}Os (36.2 keV) as 2.35×10{sup −4} and −0.20×10{sup −4}, respectively, at B3LYP level for SARC basis set.

Evaluation of severe accident safety system value based on averting financial risks

International Nuclear Information System (INIS)

Hatch, S.W.; Benjamin, A.S.; Bennett, P.R.

1983-01-01

The Severe Accident Risk Reduction Program is being performed to benchmark the risks from nuclear power plants and to assess the benefits and impacts of a set of severe accident safety features. This paper describes the program in general and presents some preliminary results. These results include estimates of the financial risks associated with the operation of six reference plants and the value of severe accident prevention and mitigation safety systems in averting these risks. The results represent initial calculations and will be iterated before being used to support NRC decisions

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.

Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste.

Science.gov (United States)

Heuel-Fabianek, Burkhard; Hille, Ralf

2005-01-01

During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.

Toxicological benchmarks for screening potential contaminants of concern for effects on aquatic biota: 1996 revision

Energy Technology Data Exchange (ETDEWEB)

Suter, G.W. II [Oak Ridge National Lab., TN (United States); Tsao, C.L. [Duke Univ., Durham, NC (United States). School of the Environment

1996-06-01

This report presents potential screening benchmarks for protection of aquatic life form contaminants in water. Because there is no guidance for screening for benchmarks, a set of alternative benchmarks is presented herein. This report presents the alternative benchmarks for chemicals that have been detected on the Oak Ridge Reservation. It also presents the data used to calculate the benchmarks and the sources of the data. It compares the benchmarks and discusses their relative conservatism and utility. Also included is the updates of benchmark values where appropriate, new benchmark values, secondary sources are replaced by primary sources, and a more complete documentation of the sources and derivation of all values are presented.

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

A review of the current state-of-the-art methodology for handling bias and uncertainty in performing criticality safety evaluations. Final report

International Nuclear Information System (INIS)

Disney, R.K.

1994-10-01

The methodology for handling bias and uncertainty when calculational methods are used in criticality safety evaluations (CSE's) is a rapidly evolving technology. The changes in the methodology are driven by a number of factors. One factor responsible for changes in the methodology for handling bias and uncertainty in CSE's within the overview of the US Department of Energy (DOE) is a shift in the overview function from a ''site'' perception to a more uniform or ''national'' perception. Other causes for change or improvement in the methodology for handling calculational bias and uncertainty are; (1) an increased demand for benchmark criticals data to expand the area (range) of applicability of existing data, (2) a demand for new data to supplement existing benchmark criticals data, (3) the increased reliance on (or need for) computational benchmarks which supplement (or replace) experimental measurements in critical assemblies, and (4) an increased demand for benchmark data applicable to the expanded range of conditions and configurations encountered in DOE site restoration and remediation

Slope Safety Calculation With A Non-Linear Mohr Criterion Using Finite Element Method

DEFF Research Database (Denmark)

Clausen, Johan; Damkilde, Lars

2005-01-01

Safety factors for soil slopes are calculated using a non-linear Mohr envelope. The often used linear Mohr-Coulomb envelope tends to overestimate the safety as the material parameters are usually determined at much higher stress levels, than those present at slope failure. Experimental data...

Performance of exchange-correlation functionals in density functional theory calculations for liquid metal: A benchmark test for sodium

Science.gov (United States)

Han, Jeong-Hwan; Oda, Takuji

2018-04-01

The performance of exchange-correlation functionals in density-functional theory (DFT) calculations for liquid metal has not been sufficiently examined. In the present study, benchmark tests of Perdew-Burke-Ernzerhof (PBE), Armiento-Mattsson 2005 (AM05), PBE re-parameterized for solids, and local density approximation (LDA) functionals are conducted for liquid sodium. The pair correlation function, equilibrium atomic volume, bulk modulus, and relative enthalpy are evaluated at 600 K and 1000 K. Compared with the available experimental data, the errors range from -11.2% to 0.0% for the atomic volume, from -5.2% to 22.0% for the bulk modulus, and from -3.5% to 2.5% for the relative enthalpy depending on the DFT functional. The generalized gradient approximation functionals are superior to the LDA functional, and the PBE and AM05 functionals exhibit the best performance. In addition, we assess whether the error tendency in liquid simulations is comparable to that in solid simulations, which would suggest that the atomic volume and relative enthalpy performances are comparable between solid and liquid states but that the bulk modulus performance is not. These benchmark test results indicate that the results of liquid simulations are significantly dependent on the exchange-correlation functional and that the DFT functional performance in solid simulations can be used to roughly estimate the performance in liquid simulations.

Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

Energy Technology Data Exchange (ETDEWEB)

Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-11

Training classes for nuclear criticality safety, MCNP documentation. The need for, and problems surrounding, validation of computer codes and data area considered first. Then some background for MCNP & Whisper is given--best practices for Monte Carlo criticality calculations, neutron spectra, S(α,β) thermal neutron scattering data, nuclear data sensitivities, covariance data, and correlation coefficients. Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the Monte Carlo radiation transport package MCNP. Whisper's methodology (benchmark selection – C_k's, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usage are discussed.

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

Guideline for benchmarking thermal treatment systems for low-level mixed waste

International Nuclear Information System (INIS)

Hoffman, D.P.; Gibson, L.V. Jr.; Hermes, W.H.; Bastian, R.E.; Davis, W.T.

1994-01-01

A process for benchmarking low-level mixed waste (LLMW) treatment technologies has been developed. When used in conjunction with the identification and preparation of surrogate waste mixtures, and with defined quality assurance and quality control procedures, the benchmarking process will effectively streamline the selection of treatment technologies being considered by the US Department of Energy (DOE) for LLMW cleanup and management. Following the quantitative template provided in the benchmarking process will greatly increase the technical information available for the decision-making process. The additional technical information will remove a large part of the uncertainty in the selection of treatment technologies. It is anticipated that the use of the benchmarking process will minimize technology development costs and overall treatment costs. In addition, the benchmarking process will enhance development of the most promising LLMW treatment processes and aid in transferring the technology to the private sector. To instill inherent quality, the benchmarking process is based on defined criteria and a structured evaluation format, which are independent of any specific conventional treatment or emerging process technology. Five categories of benchmarking criteria have been developed for the evaluation: operation/design; personnel health and safety; economics; product quality; and environmental quality. This benchmarking document gives specific guidance on what information should be included and how it should be presented. A standard format for reporting is included in Appendix A and B of this document. Special considerations for LLMW are presented and included in each of the benchmarking categories

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Calculation of partial derivatives of thermophysical properties of sodium for safety analysis

International Nuclear Information System (INIS)

Shan Jianqiang; Qiu Suizhang; Zhu Jizhou; Zhang Guiqin

1997-01-01

According to the characters of safety analysis of LMFBR, the partial derivatives formula of some special thermophysical properties of sodium, including single-and two-phase properties, are calculated based on the basic Maxwell equations, and on the formulae of basic thermophysical properties of sodium which were verified abroad. The present study can provide theoretical base for safety analysis of LMFBR

Verification of the code DYN3D/R with the help of international benchmarks

International Nuclear Information System (INIS)

Grundmann, U.; Rohde, U.

1997-10-01

Different benchmarks for reactors with quadratic fuel assemblies were calculated with the code DYN3D/R. In this report comparisons with the results of the reference solutions are carried out. The results of DYN3D/R and the reference calculation for the eigenvalue k eff and the power distribution are shown for the steady-state 3-dimensional IAEA-Benchmark. The results of NEACRP-Benchmarks on control rod ejections in a standard PWR were compared with the reference solutions published by the NEA Data Bank. For assessing the accuracy of DYN3D/R results in comparison to other codes the deviations to the reference solutions are considered. Detailed comparisons with the published reference solutions of the NEA-NSC Benchmarks on uncontrolled withdrawal of control rods are made. The influence of the axial nodalization is also investigated. All in all, a good agreement of the DYN3D/R results with the reference solutions can be seen for the considered benchmark problems. (orig.) [de

Benchmark experiment on vanadium assembly with D-T neutrons. In-situ measurement

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio; Kasugai, Yoshimi; Konno, Chikara; Wada, Masayuki; Oyama, Yukio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murata, Isao; Kokooo; Takahashi, Akito

1998-03-01

Fusion neutronics benchmark experimental data on vanadium were obtained for neutrons in almost entire energies as well as secondary gamma-rays. Benchmark calculations for the experiment were performed to investigate validity of recent nuclear data files, i.e., JENDL Fusion File, FENDL/E-1.0 and EFF-3. (author)

Static benchmarking of the NESTLE advanced nodal code

International Nuclear Information System (INIS)

Mosteller, R.D.

1997-01-01

Results from the NESTLE advanced nodal code are presented for multidimensional numerical benchmarks representing four different types of reactors, and predictions from NESTLE are compared with measured data from pressurized water reactors (PWRs). The numerical benchmarks include cases representative of PWRs, boiling water reactors (BWRs), CANDU heavy water reactors (HWRs), and high-temperature gas-cooled reactors (HTGRs). The measured PWR data include critical soluble boron concentrations and isothermal temperature coefficients of reactivity. The results demonstrate that NESTLE correctly solves the multigroup diffusion equations for both Cartesian and hexagonal geometries, that it reliably calculates k eff and reactivity coefficients for PWRs, and that--subsequent to the incorporation of additional thermal-hydraulic models--it will be able to perform accurate calculations for the corresponding parameters in BWRs, HWRs, and HTGRs as well

Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

International Nuclear Information System (INIS)

Bess, John D.; Marshall, Margaret A.; Gorham, Mackenzie L.; Christensen, Joseph; Turnbull, James C.; Clark, Kim

2011-01-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) (1) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) (2) were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

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.

Food safety performance indicators to benchmark food safety output of food safety management systems

NARCIS (Netherlands)

Jacxsens, L.; Uyttendaele, M.; Devlieghere, F.; Rovira, J.; Oses Gomez, S.; Luning, P.A.

2010-01-01

There is a need to measure the food safety performance in the agri-food chain without performing actual microbiological analysis. A food safety performance diagnosis, based on seven indicators and corresponding assessment grids have been developed and validated in nine European food businesses.

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

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.

Concrete reflected cylinders of highly enriched solutions of uranyl nitrate ICSBEP Benchmark: A re-evaluation by means of MCNPX using ENDF/B-VI cross section library

International Nuclear Information System (INIS)

Cruzate, J.A.; Carelli, J.L.

2011-01-01

This work presents a theoretical re-evaluation of a set of original experiments included in the 2009 issue of the International Handbook of Evaluated Criticality Safety Benchmark Experiments, as “Concrete Reflected Cylinders of Highly Enriched Solutions of Uranyl Nitrate” (identification number: HEU-SOL-THERM- 002) [4]. The present evaluation has been made according to benchmark specifications [4], and added data taken out of the original published report [3], but applying a different approach, resulting in a more realistic calculation model. In addition, calculations have been made using the latest version of MCNPX Monte Carlo code, combined with an updated set of cross section data, the continuous-energy ENDF/B-VI library. This has resulted in a comprehensive model for the given experimental situation. Uncertainties analysis has been made based on the evaluation of experimental data presented in the HEU-SOLTHERM-002 report. Resulting calculations with the present improved physical model have been able to reproduce the criticality of configurations within 0.5%, in good agreement with experimental data. Results obtained in the analysis of uncertainties are in general agreement with those at HEU-SOL-THERM-002 benchmark document. Qualitative results from analyses made in the present work can be extended to similar fissile systems: well moderated units of 235 U solutions, reflected with concrete from all directions. Results have confirmed that neutron absorbers, even as impurities, must be taken into account in calculations if at least approximate proportions were known. (authors)

Russian system of computerized analysis for licensing at atomic industry (SCALA) and its validation on ICSBEP handbook data and some burnup calculations

International Nuclear Information System (INIS)

Ivanova, T.; Nikolaev, M.; Polyakov, A.; Saraeva, T.; Tsiboulia, A.

2000-01-01

The System of Computerized Analysis for Licensing at Atomic industry (SCALA) is a Russian analogue of the well-known SCALE system. For criticality evaluations the ABBN-93 system is used with TWODANT and with joined American KENO and Russian MMK Monte-Carlo code MMKKENO. Using the same cross sections and input models, all these codes give results that coincide within the statistical uncertainties (for Monte-Carlo codes). Validation of criticality calculations using SCALA was performed using data presented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Another task of the work was to test the burnup capability of SCALA system in complex geometry in compare with other codes. Benchmark models of VVER type reactor assemblies with UO 2 and MOX fuel including the cases with burnable gadolinium absorbers were calculated. KENO-VI and MMK codes were used for power distribution calculations, ORIGEN code was used for the isotopic kinetics calculations. (authors)

Benchmarking NNWSI flow and transport codes: COVE 1 results

International Nuclear Information System (INIS)

Hayden, N.K.

1985-06-01

The code verification (COVE) activity of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project is the first step in certification of flow and transport codes used for NNWSI performance assessments of a geologic repository for disposing of high-level radioactive wastes. The goals of the COVE activity are (1) to demonstrate and compare the numerical accuracy and sensitivity of certain codes, (2) to identify and resolve problems in running typical NNWSI performance assessment calculations, and (3) to evaluate computer requirements for running the codes. This report describes the work done for COVE 1, the first step in benchmarking some of the codes. Isothermal calculations for the COVE 1 benchmarking have been completed using the hydrologic flow codes SAGUARO, TRUST, and GWVIP; the radionuclide transport codes FEMTRAN and TRUMP; and the coupled flow and transport code TRACR3D. This report presents the results of three cases of the benchmarking problem solved for COVE 1, a comparison of the results, questions raised regarding sensitivities to modeling techniques, and conclusions drawn regarding the status and numerical sensitivities of the codes. 30 refs

Technology relevance of the 'uncertainty analysis in modelling' project for nuclear reactor safety

International Nuclear Information System (INIS)

D'Auria, F.; Langenbuch, S.; Royer, E.; Del Nevo, A.; Parisi, C.; Petruzzi, A.

2007-01-01

The OECD/NEA Nuclear Science Committee (NSC) endorsed the setting up of an Expert Group on Uncertainty Analysis in Modelling (UAM) in June 2006. This Expert Group reports to the Working Party on Scientific issues in Reactor Systems (WPRS) and because it addresses multi-scale / multi-physics aspects of uncertainty analysis, it will work in close co-ordination with the benchmark groups on coupled neutronics-thermal-hydraulics and on coupled core-plant problems, and the CSNI Group on Analysis and Management of Accidents (GAMA). The NEA/NSC has endorsed that this activity be undertaken with Prof. K. Ivanov from the Pennsylvania State University (PSU) as the main coordinator and host with the assistance of the Scientific Board. The objective of the proposed work is to define, coordinate, conduct, and report an international benchmark for uncertainty analysis in best-estimate coupled code calculations for design, operation, and safety analysis of LWRs entitled 'OECD UAM LWR Benchmark'. At the First Benchmark Workshop (UAM-1) held from 10 to 11 May 2007 at the OECD/NEA, one action concerned the forming of a sub-group, led by F. D'Auria, member of CSNI, responsible for defining the objectives, the impact and benefit of the UAM for safety and licensing. This report is the result of this action by the subgroup. (authors)

Benchmarking: A tool for conducting self-assessment

International Nuclear Information System (INIS)

Perkey, D.N.

1992-01-01

There is more information on nuclear plant performance available than can reasonably be assimilated and used effectively by plant management or personnel responsible for self-assessment. Also, it is becoming increasingly more important that an effective self-assessment program uses internal parameters not only to evaluate performance, but to incorporate lessons learned from other plants. Because of the quantity of information available, it is important to focus efforts and resources in areas where safety or performance is a concern and where the most improvement can be realized. One of the techniques that is being used to effectively accomplish this is benchmarking. Benchmarking involves the use of various sources of information to self-identify a plant's strengths and weaknesses, identify which plants are strong performers in specific areas, evaluate what makes a top performer, and incorporate the success factors into existing programs. The formality with which benchmarking is being implemented varies widely depending on the objective. It can be as simple as looking at a single indicator, such as systematic assessment of licensee performance (SALP) in engineering and technical support, then surveying the top performers with specific questions. However, a more comprehensive approach may include the performance of a detailed benchmarking study. Both operational and economic indicators may be used in this type of evaluation. Some of the indicators that may be considered and the limitations of each are discussed

Benchmark calculations for critical experiments at FKBN-M facility with uranium-plutonium-polyethylene systems using JENDL-3.2 and MVP Monte-Carlo code

International Nuclear Information System (INIS)

Obara, Toru; Morozov, A.G.; Kevrolev, V.V.; Kuznetsov, V.V.; Treschalin, S.A.; Lukin, A.V.; Terekhin, V.A.; Sokolov, Yu.A.; Kravchenko, V.G.

2000-01-01

Benchmark calculations were performed for critical experiments at FKBN-M facility in RFNC-VNIITF, Russia using JENDL-3.2 nuclear data library and continuous energy Monte-Carlo code MVP. The fissile materials were high-enriched uranium and plutonium. Polyethylene was used as moderator. The neutron spectrum was changed by changing the geometry. Calculation results by MVP showed some errors. Discussion was made by reaction rates and η values obtained by MVP. It showed the possibility that cross sections of U-235 had different trend of error in fast and thermal energy region respectively. It also showed the possibility of some error of cross section of Pu-239 in high energy region. (author)

New Improved Nuclear Data for Nuclear Criticality and Safety

International Nuclear Information System (INIS)

Guber, Klaus H.; Leal, Luiz C.; Lampoudis, C.; Kopecky, S.; Schillebeeckx, P.; Emiliani, F.; Wynants, R.; Siegler, P.

2011-01-01

The Geel Electron Linear Accelerator (GELINA) was used to measure neutron total and capture cross sections of 182,183,184,186 W and 63,65 Cu in the energy range from 100 eV to ∼200 keV using the time-of-flight method. GELINA is the only high-power white neutron source with excellent timing resolution and ideally suited for these experiments. Concerns about the use of existing cross-section data in nuclear criticality calculations using Monte Carlo codes and benchmarks were a prime motivator for the new cross-section measurements. To support the Nuclear Criticality Safety Program, neutron cross-section measurements were initiated using GELINA at the EC-JRC-IRMM. Concerns about data deficiencies in some existing cross-section evaluations from libraries such as ENDF/B, JEFF, or JENDL for nuclear criticality calculations were the prime motivator for new cross-section measurements. Over the past years many troubles with existing nuclear data have emerged, such as problems related to proper normalization, neutron sensitivity backgrounds, poorly characterized samples, and use of improper pulse-height weighting functions. These deficiencies may occur in the resolved- and unresolved-resonance region and may lead to erroneous nuclear criticality calculations. An example is the use of the evaluated neutron cross-section data for tungsten in nuclear criticality safety calculations, which exhibit discrepancies in benchmark calculations and show the need for reliable covariance data. We measured the neutron total and capture cross sections of 182,183,184,186 W and 63,65 Cu in the neutron energy range from 100 eV to several hundred keV. This will help to improve the representation of the cross sections since most of the available evaluated data rely only on old measurements. Usually these measurements were done with poor experimental resolution or only over a very limited energy range, which is insufficient for the current application.