Problems of the power plant shield optimization

International Nuclear Information System (INIS)

Abagyan, A.A.; Dubinin, A.A.; Zhuravlev, V.I.; Kurachenko, Yu.A.; Petrov, Eh.E.

1981-01-01

General approaches to the solution of problems on the nuclear power plant radiation shield optimization are considered. The requirements to the shield parameters are formulated in a form of restrictions on a number of functionals, determined by the solution of γ quantum and neutron transport equations or dimensional and weight characteristics of shield components. Functional determined by weight-dimensional parameters (shield cost, mass and thickness) and functionals, determined by radiation fields (equivalent dose rate, produced by neutrons and γ quanta, activation functional, radiation functional, heat flux, integral heat flux in a particular part of the shield volume, total energy flux through a particular shield surface are considered. The following methods of numerical solution of simplified optimization problems are discussed: semiempirical methods using radiation transport physical leaks, numerical solution of approximate transport equations, numerical solution of transport equations for the simplest configurations making possible to decrease essentially a number of variables in the problem. The conclusion is drawn that the attained level of investigations on the problem of nuclear power plant shield optimization gives the possibility to pass on at present to the solution of problems with a more detailed account of the real shield operating conditions (shield temperature field account, its strength and other characteristics) [ru

Bench-mark experiments to study the neutron distribution in a heterogeneous reactor shielding

International Nuclear Information System (INIS)

Bolyatko, V.V.; Vyrskij, M.Yu.; Mashkovich, V.P.; Nagaev, R.Kh.; Prit'mov, A.P.; Sakharov, V.K.; Troshin, V.S.; Tikhonov, E.G.

1981-01-01

The bench-mark experiments performed at the B-2 facility of the BR-10 reactor to investigate the spatial and energy neutron distributions are described. The experimental facility includes the neutron beam channel with a slide, a mo shielding composition investigated consisted of sequential layers of steel (1KH18N9T) and graphite slabs. The neutron spectra were measured by activation method, a set of treshold and resonance detectors having been used. The detectors made it possible to obtain the absolute neutron spectra in the 1.4 eV-10 MeV range. The comparison of calculations with the results of the bench-mark experiments made it possible to prove the neutron transport calculational model realized in the ROZ-9 and ARAMAKO-2F computer codes and evaluate the validity of the ARAMAKO constants for the class of shielding compositions in question [ru

Several problems in accelerator shielding study

International Nuclear Information System (INIS)

Nakamura, Takashi; Hirayama, Hideo; Ban, Shuichi.

1980-01-01

Recently, the utilization of accelerators has increased rapidly, and the increase of accelerating energy and beam intensity is also remarkable. The studies on accelerator shielding have become important, because the amount of radiation emitted from accelerators increased, the regulation of the dose of environmental radiation was tightened, and the cost of constructing shielding rose. As the plans of constructing large accelerators have been made successively, the survey on the present state and the problems of the studies on accelerator shielding was carried out. Accelerators are classified into electron accelerators and proton accelerators in view of the studies on shielding. In order to start the studies on accelerator shielding, first, the preparation of the cross section data is indispensable. The cross sections for generating Bremsstrahlung, photonuclear reactions generating neutrons, generation of neutrons by hadrons, nuclear reaction of neutrons and generation of gamma-ray by hadrons are described. The generation of neutrons and gamma-ray as the problems of thick targets is explained. The shielding problems are complex and diversified, but in this paper, the studies on the shielding, by which basic data are obtainable, are taken up, such as beam damping and side wall shielding. As for residual radioactivity, main nuclides and the difference of residual radioactivity according to substances have been studied. (J.P.N.)

Dosimetry and shielding

International Nuclear Information System (INIS)

Farinelli, U.

1977-01-01

Today, reactor dosimetry and shielding have wide areas of overlap as concerns both problems and methods. Increased interchange of results and know-how would benefit both. The areas of common interest include calculational methods, sensitivity studies, theoretical and experimental benchmarks, cross sections and other nuclear data, multigroup libraries and procedures for their adjustment, experimental techniques and damage functions. This paper reviews the state-of-the-art and the latest development in each of these areas as far as shielding is concerned, and suggests a number of interactions that could be profitable for reactor dosimetry. Among them, re-evaluation of the potentialities of calculational methods (in view of the recent developments) in predicting radiation environments of interest; the application of sensitivity analysis to dosimetry problems; a common effort in the field of theoretical benchmarks; the use of the shielding one-material propagation experiments as reference spectra for detector cross sections; common standardization of the detector nuclear data used in both fields; the setting up of a common (or compatible) multigroup structure and library applicable to shielding, dosimetry and core physics; the exchange of information and experience in the fields of cross section errors, correlations and adjustment; and the intercomparison of experimental techniques

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

U.S. integral and benchmark experiments

International Nuclear Information System (INIS)

Maienschein, F.C.

1978-01-01

Verification of methods for analysis of radiation-transport (shielding) problems in Liquid-Metal Fast Breeder Reactors has required a series of experiments that can be classified as benchmark, parametric, or design-confirmation experiments. These experiments, performed at the Oak Ridge Tower Shielding Facility, have included measurements of neutron transport in bulk shields of sodium, steel, and inconel and in configurations that simulate lower axial shields, pipe chases, and top-head shields. They have also included measurements of the effects of fuel stored within the reactor vessel and of gamma-ray energy deposition (heating). The paper consists of brief comments on these experiments, and also on a recent experiment in which neutron streaming problems in a Gas-Cooled Fast Breeder Reactor were studied. The need for additional experiments for a few areas of LMFBR shielding is also cited

A simplified 2D HTTR benchmark problem

International Nuclear Information System (INIS)

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

2009-01-01

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

Benchmark problems for numerical implementations of phase field models

International Nuclear Information System (INIS)

Jokisaari, A. M.; Voorhees, P. W.; Guyer, J. E.; Warren, J.; Heinonen, O. G.

2016-01-01

Here, we present the first set of benchmark problems for phase field models that are being developed by the Center for Hierarchical Materials Design (CHiMaD) and the National Institute of Standards and Technology (NIST). While many scientific research areas use a limited set of well-established software, the growing phase field community continues to develop a wide variety of codes and lacks benchmark problems to consistently evaluate the numerical performance of new implementations. Phase field modeling has become significantly more popular as computational power has increased and is now becoming mainstream, driving the need for benchmark problems to validate and verify new implementations. We follow the example set by the micromagnetics community to develop an evolving set of benchmark problems that test the usability, computational resources, numerical capabilities and physical scope of phase field simulation codes. In this paper, we propose two benchmark problems that cover the physics of solute diffusion and growth and coarsening of a second phase via a simple spinodal decomposition model and a more complex Ostwald ripening model. We demonstrate the utility of benchmark problems by comparing the results of simulations performed with two different adaptive time stepping techniques, and we discuss the needs of future benchmark problems. The development of benchmark problems will enable the results of quantitative phase field models to be confidently incorporated into integrated computational materials science and engineering (ICME), an important goal of the Materials Genome Initiative.

A 3D stylized half-core CANDU benchmark problem

International Nuclear Information System (INIS)

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

2011-01-01

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

Simplified two and three dimensional HTTR benchmark problems

International Nuclear Information System (INIS)

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

2011-01-01

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

Methodology of shielding calculation for nuclear reactors

International Nuclear Information System (INIS)

Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo

1982-01-01

A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt

Argonne Code Center: benchmark problem book

International Nuclear Information System (INIS)

1977-06-01

This report is a supplement to the original report, published in 1968, as revised. The Benchmark Problem Book is intended to serve as a source book of solutions to mathematically well-defined problems for which either analytical or very accurate approximate solutions are known. This supplement contains problems in eight new areas: two-dimensional (R-z) reactor model; multidimensional (Hex-z) HTGR model; PWR thermal hydraulics--flow between two channels with different heat fluxes; multidimensional (x-y-z) LWR model; neutron transport in a cylindrical ''black'' rod; neutron transport in a BWR rod bundle; multidimensional (x-y-z) BWR model; and neutronic depletion benchmark problems. This supplement contains only the additional pages and those requiring modification

Measurements and FLUKA Simulations of Bismuth, Aluminium and Indium Activation at the upgraded CERN Shielding Benchmark Facility (CSBF)

Science.gov (United States)

Iliopoulou, E.; Bamidis, P.; Brugger, M.; Froeschl, R.; Infantino, A.; Kajimoto, T.; Nakao, N.; Roesler, S.; Sanami, T.; Siountas, A.; Yashima, H.

2018-06-01

The CERN High energy AcceleRator Mixed field (CHARM) facility is situated in the CERN Proton Synchrotron (PS) East Experimental Area. The facility receives a pulsed proton beam from the CERN PS with a beam momentum of 24 GeV/c with 5·1011 protons per pulse with a pulse length of 350 ms and with a maximum average beam intensity of 6.7·1010 protons per second. The extracted proton beam impacts on a cylindrical copper target. The shielding of the CHARM facility includes the CERN Shielding Benchmark Facility (CSBF) situated laterally above the target that allows deep shielding penetration benchmark studies of various shielding materials. This facility has been significantly upgraded during the extended technical stop at the beginning of 2016. It consists now of 40 cm of cast iron shielding, a 200 cm long removable sample holder concrete block with 3 inserts for activation samples, a material test location that is used for the measurement of the attenuation length for different shielding materials as well as for sample activation at different thicknesses of the shielding materials. Activation samples of bismuth, aluminium and indium were placed in the CSBF in September 2016 to characterize the upgraded version of the CSBF. Monte Carlo simulations with the FLUKA code have been performed to estimate the specific production yields of bismuth isotopes (206 Bi, 205 Bi, 204 Bi, 203 Bi, 202 Bi, 201 Bi) from 209 Bi, 24 Na from 27 Al and 115 m I from 115 I for these samples. The production yields estimated by FLUKA Monte Carlo simulations are compared to the production yields obtained from γ-spectroscopy measurements of the samples taking the beam intensity profile into account. The agreement between FLUKA predictions and γ-spectroscopy measurements for the production yields is at a level of a factor of 2.

Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

Directory of Open Access Journals (Sweden)

Kulesza Joel A.

2016-01-01

Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

Development and preliminary verification of 2-D transport module of radiation shielding code ARES

International Nuclear Information System (INIS)

Zhang Penghe; Chen Yixue; Zhang Bin; Zang Qiyong; Yuan Longjun; Chen Mengteng

2013-01-01

The 2-D transport module of radiation shielding code ARES is two-dimensional neutron and radiation shielding code. The theory model was based on the first-order steady state neutron transport equation, adopting the discrete ordinates method to disperse direction variables. Then a set of differential equations can be obtained and solved with the source iteration method. The 2-D transport module of ARES was capable of calculating k eff and fixed source problem with isotropic or anisotropic scattering in x-y geometry. The theoretical model was briefly introduced and series of benchmark problems were verified in this paper. Compared with the results given by the benchmark, the maximum relative deviation of k eff is 0.09% and the average relative deviation of flux density is about 0.60% in the BWR cells benchmark problem. As for the fixed source problem with isotropic and anisotropic scattering, the results of the 2-D transport module of ARES conform with DORT very well. These numerical results of benchmark problems preliminarily demonstrate that the development process of the 2-D transport module of ARES is right and it is able to provide high precision result. (authors)

The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

International Nuclear Information System (INIS)

Batistoni, P.; Angelone, M.; Martone, M.; Petrizzi, L.; Pillon, M.; Rado, V.; Santamarina, A.; Abidi, I.; Gastaldi, G.; Joyer, P.; Marquette, J.P.; Martini, M.

1994-01-01

In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat a l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The γ-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File). ((orig.))

Design experience: CRBRP radiation shielding

International Nuclear Information System (INIS)

Disney, R.K.; Chan, T.C.; Gallo, F.G.; Hedgecock, L.R.; McGinnis, C.A.; Wrights, G.N.

1978-11-01

The Clinch River Breeder Reactor Plant (CRBRP) is being designed as a fast breeder demonstration project in the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program. Radiation shielding design of the facility consists of a comprehensive design approach to assure compliance with design and government regulatory requirements. Studies conducted during the CRBRP design process involved the aspects of radiation shielding dealing with protection of components, systems, and personnel from radiation exposure. Achievement of feasible designs, while considering the mechanical, structural, nuclear, and thermal performance of the component or system, has required judicious trade-offs in radiation shielding performance. Specific design problems which have been addressed are in-vessel radial shielding to protect permanent core support structures, flux monitor system shielding to isolate flux monitoring systems for extraneous background sources, reactor vessel support shielding to allow personnel access to the closure head during full power operation, and primary heat transport system pipe chaseway shielding to limit intermediate heat transport system sodium system coolant activation. The shielding design solutions to these problems defined a need for prototypic or benchmark experiments to provide assurance of the predicted shielding performance of selected design solutions and the verification of design methodology. Design activities of CRBRP plant components an systems, which have the potential for radiation exposure of plant personnel during operation or maintenance, are controlled by a design review process related to radiation shielding. The program implements design objectives, design requirements, and cost/benefit guidelines to assure that radiation exposures will be ''as low as reasonably achievable''

Benchmark problems for repository siting models

International Nuclear Information System (INIS)

Ross, B.; Mercer, J.W.; Thomas, S.D.; Lester, B.H.

1982-12-01

This report describes benchmark problems to test computer codes used in siting nuclear waste repositories. Analytical solutions, field problems, and hypothetical problems are included. Problems are included for the following types of codes: ground-water flow in saturated porous media, heat transport in saturated media, ground-water flow in saturated fractured media, heat and solute transport in saturated porous media, solute transport in saturated porous media, solute transport in saturated fractured media, and solute transport in unsaturated porous media

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

Impact testing and analysis for structural code benchmarking

International Nuclear Information System (INIS)

Glass, R.E.

1989-01-01

Sandia National Laboratories, in cooperation with industry and other national laboratories, has been benchmarking computer codes used to predict the structural, thermal, criticality, and shielding behavior of radioactive materials packages. The first step in the benchmarking of the codes was to develop standard problem sets and to compare the results from several codes and users. This step for structural analysis codes has been completed as described in Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Casks, R.E. Glass, Sandia National Laboratories, 1985. The problem set is shown in Fig. 1. This problem set exercised the ability of the codes to predict the response to end (axisymmetric) and side (plane strain) impacts with both elastic and elastic/plastic materials. The results from these problems showed that there is good agreement in predicting elastic response. Significant differences occurred in predicting strains for the elastic/plastic models. An example of the variation in predicting plastic behavior is given, which shows the hoop strain as a function of time at the impacting end of Model B. These differences in predicting plastic strains demonstrated a need for benchmark data for a cask-like problem

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.

The Benchmark experiment on stainless steel bulk shielding at the Frascati neutron generator

International Nuclear Information System (INIS)

Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V.

1994-11-01

In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L'Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S N and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the ENEA Italian Agency for New Technologies, Energy and Environment) team

Automatic generation of 3D fine mesh geometries for the analysis of the venus-3 shielding benchmark experiment with the Tort code

International Nuclear Information System (INIS)

Pescarini, M.; Orsi, R.; Martinelli, T.

2003-01-01

In many practical radiation transport applications today the cost for solving refined, large size and complex multi-dimensional problems is not so much computing but is linked to the cumbersome effort required by an expert to prepare a detailed geometrical model, verify and validate that it is correct and represents, to a specified tolerance, the real design or facility. This situation is, in particular, relevant and frequent in reactor core criticality and shielding calculations, with three-dimensional (3D) general purpose radiation transport codes, requiring a very large number of meshes and high performance computers. The need for developing tools that make easier the task to the physicist or engineer, by reducing the time required, by facilitating through effective graphical display the verification of correctness and, finally, that help the interpretation of the results obtained, has clearly emerged. The paper shows the results of efforts in this field through detailed simulations of a complex shielding benchmark experiment. In the context of the activities proposed by the OECD/NEA Nuclear Science Committee (NSC) Task Force on Computing Radiation Dose and Modelling of Radiation-Induced Degradation of Reactor Components (TFRDD), the ENEA-Bologna Nuclear Data Centre contributed with an analysis of the VENUS-3 low-flux neutron shielding benchmark experiment (SCK/CEN-Mol, Belgium). One of the targets of the work was to test the BOT3P system, originally developed at the Nuclear Data Centre in ENEA-Bologna and actually released to OECD/NEA Data Bank for free distribution. BOT3P, ancillary system of the DORT (2D) and TORT (3D) SN codes, permits a flexible automatic generation of spatial mesh grids in Cartesian or cylindrical geometry, through combinatorial geometry algorithms, following a simplified user-friendly approach. This system demonstrated its validity also in core criticality analyses, as for example the Lewis MOX fuel benchmark, permitting to easily

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Activities of the Shielding Subcommittee of the ENDF/B Cross Section Evaluation Working Group

International Nuclear Information System (INIS)

Roussin, R.W.

1977-01-01

The Shielding Subcommittee of the Cross Section Evaluation Working Group (CSEWG) was established in 1967 to help ensure that the content of the ENDF/B cross section library was adequate for treating shielding problems. Early work of the subcommittee concentrated on devising formats for gamma-ray interaction and production data, as well as providing programs for testing the clerical and physics consistency of the files. The Radiation Shielding Information Center (RSIC) collaborated directly with evaluators on behalf of the National Neutron Cross Section Center (NNCSC) to begin testing and adding data sets to be fed into the official ENDF/B libraries. These efforts, which were sponsored by AEC-DRDT (now ERDA-DRDD), were augmented greatly through the Defense Nuclear Agency program of establishing a working cross section library in ENDF format. The effort concentrated on evaluation and testing of materials of interest to DNA programs and providing these for inclusion in the ENDF/B library. Shielding data testing efforts, as a part of the CSEWG Data Testing Program, are now also an integral part of the Shielding Subcommittee effort. Procedures for writing and approving the shielding benchmarks were devised by Shielding Subcommittee members. Data testing benchmark experiments have been documented and analyzed, and the most recent results for ENDF/B-IV are as reported as part of ENDF-230, ''Benchmark Testing of ENDF/B-IV.''

Impact testing and analysis for structural code benchmarking

International Nuclear Information System (INIS)

Glass, R.E.

1989-01-01

Sandia National Laboratories, in cooperation with industry and other national laboratories, has been benchmarking computer codes (''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Cask,'' R.E. Glass, Sandia National Laboratories, 1985; ''Sample Problem Manual for Benchmarking of Cask Analysis Codes,'' R.E. Glass, Sandia National Laboratories, 1988; ''Standard Thermal Problem Set for the Evaluation of Heat Transfer Codes Used in the Assessment of Transportation Packages, R.E. Glass, et al., Sandia National Laboratories, 1988) used to predict the structural, thermal, criticality, and shielding behavior of radioactive materials packages. The first step in the benchmarking of the codes was to develop standard problem sets and to compare the results from several codes and users. This step for structural analysis codes has been completed as described in ''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Casks,'' R.E. Glass, Sandia National Laboratories, 1985. The problem set is shown in Fig. 1. This problem set exercised the ability of the codes to predict the response to end (axisymmetric) and side (plane strain) impacts with both elastic and elastic/plastic materials. The results from these problems showed that there is good agreement in predicting elastic response. Significant differences occurred in predicting strains for the elastic/plastic models. An example of the variation in predicting plastic behavior is given, which shows the hoop strain as a function of time at the impacting end of Model B. These differences in predicting plastic strains demonstrated a need for benchmark data for a cask-like problem. 6 refs., 5 figs

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

International Nuclear Information System (INIS)

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

1994-07-01

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

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

μ-synthesis for the coupled mass benchmark problem

DEFF Research Database (Denmark)

Niemann, Hans Henrik; Stoustrup, J.; Tøffner-Clausen, S.

1997-01-01

A robust controller design for the coupled mass benchmark problem is presented in this paper. The applied design method is based on a modified D-K iteration, i.e. μ-synthesis which take care of mixed real and complex perturbations sets. This μ-synthesis method for mixed perturbation sets is a str......A robust controller design for the coupled mass benchmark problem is presented in this paper. The applied design method is based on a modified D-K iteration, i.e. μ-synthesis which take care of mixed real and complex perturbations sets. This μ-synthesis method for mixed perturbation sets...

Piping benchmark problems for the Westinghouse AP600 Standardized Plant

International Nuclear Information System (INIS)

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

1997-01-01

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

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

OpenAIRE

Pescarini Massimo; Orsi Roberto; Frisoni Manuela

2016-01-01

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

Self-shielding effect in unresolved resonance data in JENDL-4.0

International Nuclear Information System (INIS)

Konno, Chikara; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Kato, Yoshinari

2012-01-01

At International Conference on Nuclear Data for Science and Technology in 2007 we pointed out that most of unresolved resonance data in JENDL-3.3 have a problem related to self-shielding correction. Here with a simple calculation model we have investigated whether the latest JENDL, JENDL-4.0, was improved for the problem or not. The results suggest that unresolved resonance data in JENDL-4.0 have no problem, but it seems that self-shielding effects for the unresolved resonance data in JENDL-4.0 are too large. New benchmark experiments for unresolved resonance data are strongly recommended in order to verify unresolved resonance data. (author)

Radiation streaming: the continuing problem of shield design

International Nuclear Information System (INIS)

Avery, A.F.

1977-01-01

The practical problems of shield design are reviewed and the major difficulties are shown to be those associated with streaming problems. The situations in which streaming occurs in various types of reactor are described including LMFBR's and fusion devices, and examples are given of ways in which the problems have been solved

A HYDROCHEMICAL HYBRID CODE FOR ASTROPHYSICAL PROBLEMS. I. CODE VERIFICATION AND BENCHMARKS FOR A PHOTON-DOMINATED REGION (PDR)

International Nuclear Information System (INIS)

Motoyama, Kazutaka; Morata, Oscar; Hasegawa, Tatsuhiko; Shang, Hsien; Krasnopolsky, Ruben

2015-01-01

A two-dimensional hydrochemical hybrid code, KM2, is constructed to deal with astrophysical problems that would require coupled hydrodynamical and chemical evolution. The code assumes axisymmetry in a cylindrical coordinate system and consists of two modules: a hydrodynamics module and a chemistry module. The hydrodynamics module solves hydrodynamics using a Godunov-type finite volume scheme and treats included chemical species as passively advected scalars. The chemistry module implicitly solves nonequilibrium chemistry and change of energy due to thermal processes with transfer of external ultraviolet radiation. Self-shielding effects on photodissociation of CO and H 2 are included. In this introductory paper, the adopted numerical method is presented, along with code verifications using the hydrodynamics module and a benchmark on the chemistry module with reactions specific to a photon-dominated region (PDR). Finally, as an example of the expected capability, the hydrochemical evolution of a PDR is presented based on the PDR benchmark

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 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

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

Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system benchmark experiments at the CEA Valduc SILENE facility

International Nuclear Information System (INIS)

Miller, Thomas Martin; Celik, Cihangir; Dunn, Michael E; Wagner, John C; McMahan, Kimberly L; Authier, Nicolas; Jacquet, Xavier; Rousseau, Guillaume; Wolff, Herve; Savanier, Laurence; Baclet, Nathalie; Lee, Yi-kang; Trama, Jean-Christophe; Masse, Veronique; Gagnier, Emmanuel; Naury, Sylvie; Blanc-Tranchant, Patrick; Hunter, Richard; Kim, Soon; Dulik, George Michael; Reynolds, Kevin H.

2015-01-01

In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereas in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available

Criticality benchmarks for COG: A new point-wise Monte Carlo code

International Nuclear Information System (INIS)

Alesso, H.P.; Pearson, J.; Choi, J.S.

1989-01-01

COG is a new point-wise Monte Carlo code being developed and tested at LLNL for the Cray computer. It solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) charged particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems. However, its point-wise cross-sections also make it effective for a wide variety of criticality problems. COG has some similarities to a number of other computer codes used in the shielding and criticality community. These include the Lawrence Livermore National Laboratory (LLNL) codes TART and ALICE, the Los Alamos National Laboratory code MCNP, the Oak Ridge National Laboratory codes 05R, 06R, KENO, and MORSE, the SACLAY code TRIPOLI, and the MAGI code SAM. Each code is a little different in its geometry input and its random-walk modification options. Validating COG consists in part of running benchmark calculations against critical experiments as well as other codes. The objective of this paper is to present calculational results of a variety of critical benchmark experiments using COG, and to present the resulting code bias. Numerous benchmark calculations have been completed for a wide variety of critical experiments which generally involve both simple and complex physical problems. The COG results, which they report in this paper, have been excellent

Problems in radiation shielding calculations with Monte Carlo methods

International Nuclear Information System (INIS)

Ueki, Kohtaro

1985-01-01

The Monte Carlo method is a very useful tool for solving a large class of radiation transport problem. In contrast with deterministic method, geometric complexity is a much less significant problem for Monte Carlo calculations. However, the accuracy of Monte Carlo calculations is of course, limited by statistical error of the quantities to be estimated. In this report, we point out some typical problems to solve a large shielding system including radiation streaming. The Monte Carlo coupling technique was developed to settle such a shielding problem accurately. However, the variance of the Monte Carlo results using the coupling technique of which detectors were located outside the radiation streaming, was still not enough. So as to bring on more accurate results for the detectors located outside the streaming and also for a multi-legged-duct streaming problem, a practicable way of ''Prism Scattering technique'' is proposed in the study. (author)

Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

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

1993-08-01

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

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

Science.gov (United States)

Lovaglio, Pietro Giorgio

2012-01-01

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

A suite of benchmark and challenge problems for enhanced geothermal systems

Energy Technology Data Exchange (ETDEWEB)

White, Mark; Fu, Pengcheng; McClure, Mark; Danko, George; Elsworth, Derek; Sonnenthal, Eric; Kelkar, Sharad; Podgorney, Robert

2017-11-06

A diverse suite of numerical simulators is currently being applied to predict or understand the performance of enhanced geothermal systems (EGS). To build confidence and identify critical development needs for these analytical tools, the United States Department of Energy, Geothermal Technologies Office sponsored a Code Comparison Study (GTO-CCS), with participants from universities, industry, and national laboratories. A principal objective for the study was to create a community forum for improvement and verification of numerical simulators for EGS modeling. Teams participating in the study were those representing U.S. national laboratories, universities, and industries, and each team brought unique numerical simulation capabilities to bear on the problems. Two classes of problems were developed during the study, benchmark problems and challenge problems. The benchmark problems were structured to test the ability of the collection of numerical simulators to solve various combinations of coupled thermal, hydrologic, geomechanical, and geochemical processes. This class of problems was strictly defined in terms of properties, driving forces, initial conditions, and boundary conditions. The challenge problems were based on the enhanced geothermal systems research conducted at Fenton Hill, near Los Alamos, New Mexico, between 1974 and 1995. The problems involved two phases of research, stimulation, development, and circulation in two separate reservoirs. The challenge problems had specific questions to be answered via numerical simulation in three topical areas: 1) reservoir creation/stimulation, 2) reactive and passive transport, and 3) thermal recovery. Whereas the benchmark class of problems were designed to test capabilities for modeling coupled processes under strictly specified conditions, the stated objective for the challenge class of problems was to demonstrate what new understanding of the Fenton Hill experiments could be realized via the application of

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

Science.gov (United States)

Lovaglio, Pietro Giorgio

2012-01-01

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

Validation of NESTLE against static reactor benchmark problems

International Nuclear Information System (INIS)

Mosteller, R.D.

1996-01-01

The NESTLE advanced modal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE's geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs) and CANDU heavy- water reactors (HWRs)

Validation of NESTLE against static reactor benchmark problems

International Nuclear Information System (INIS)

Mosteller, R.D.

1996-01-01

The NESTLE advanced nodal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE's geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs), and Canada deuterium uranium (CANDU) heavy-water reactors (HWRs)

HEATING6 analysis of international thermal benchmark problem sets 1 and 2

International Nuclear Information System (INIS)

Childs, K.W.; Bryan, C.B.

1986-10-01

In order to assess the heat transfer computer codes used in the analysis of nuclear fuel shipping casks, the Nuclear Energy Agency Committee on Reactor Physics has defined seven problems for benchmarking thermal codes. All seven of these problems have been solved using the HEATING6 heat transfer code. This report presents the results of five of the problems. The remaining two problems were used in a previous benchmarking of thermal codes used in the United States, and their solutions have been previously published

Study and application of Dot 3.5 computer code in radiation shielding problems

International Nuclear Information System (INIS)

Otto, A.C.; Mendonca, A.G.; Maiorino, J.R.

1983-01-01

The application of nuclear transportation code S sub(N), Dot 3.5, to radiation shielding problems is revised. Aiming to study the better available option (convergence scheme, calculation mode), of DOT 3.5 computer code to be applied in radiation shielding problems, a standard model from 'Argonne Code Center' was selected and a combination of several calculation options to evaluate the accuracy of the results and the computational time was used, for then to select the more efficient option. To illustrate the versatility and efficacy in the application of the code for tipical shielding problems, the streaming neutrons calculation along a sodium coolant channel is ilustrated. (E.G.) [pt

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)

Benchmarks for evaluation of shielding calculations

International Nuclear Information System (INIS)

Coelho, P.R.P.; Maiorino, J.R.

1989-01-01

The spatial-energy neutron distribution emerging from a laminated shielding (stainless, polyethylene and lead) were measured by a fast neutron spectrometer and some experimental results were compared with those calculated by a network of codes. The source neutrons incident in the shielding were 14 MeV neutrons from a H-3(d,n)He-4 reaction coming from a Van de Graaff accelerator. Experimentally was verified a good radial symmetry of neutron energy-spectrum, and also a moderation and attenuation effect for points located out of the central axis of symmetry. These results indicate that the experiment can be well modelated by R-Z geometry. A neutron-energy spectra calculated by DOT 3.5 was compared with the measured spectra, showing a good agreement in the shape and value of the spectra (12% for an integrated spectrum from 2 to 16 MeV). (author) [pt

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)

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

Validation of Shielding Analysis Capability of SuperMC with SINBAD

Directory of Open Access Journals (Sweden)

Chen Chaobin

2017-01-01

Full Text Available Abstract: The shielding analysis capability of SuperMC was validated with the Shielding Integral Benchmark Archive Database (SINBAD. The SINBAD was compiled by RSICC and NEA, it includes numerous benchmark experiments performed with the D-T fusion neutron source facilities of OKTAVIAN, FNS, IPPE, etc. The results from SuperMC simulation were compared with experimental data and MCNP results. Very good agreement with deviation lower than 1% was achieved and it suggests that SuperMC is reliable in shielding calculation.

Benchmark analyses for the ITER bulk shield experiment with EFF-3.0, -3.1 and FENDL-1, -2 nuclear cross-section data

International Nuclear Information System (INIS)

Fischer, U.; Wu, Y.; Hansen, W.; Richter, D.; Seidel, K.; Unholzer, S.

1999-01-01

The present article is part of the summary report on the Consultants' Meeting on the transport sublibrary of the Fusion Evaluated Data Library version 2.0. It reports on the comparison between benchmark experiments on a mock-up of the ITER inboard shield system at FNG, Frascati and Monte Carlo calculations, using different versions of the FENDL and EFF libraries

Workshops and problems for benchmarking eddy current codes

International Nuclear Information System (INIS)

Turner, L.R.; Davey, K.; Ida, N.; Rodger, D.; Kameari, A.; Bossavit, A.; Emson, C.R.I.

1988-02-01

A series of six workshops was held to compare eddy current codes, using six benchmark problems. The problems include transient and steady-state ac magnetic fields, close and far boundary conditions, magnetic and non-magnetic materials. All the problems are based either on experiments or on geometries that can be solved analytically. The workshops and solutions to the problems are described. Results show that many different methods and formulations give satisfactory solutions, and that in many cases reduced dimensionality or coarse discretization can give acceptable results while reducing the computer time required. 13 refs., 1 tab

Supply network configuration—A benchmarking problem

Science.gov (United States)

Brandenburg, Marcus

2018-03-01

Managing supply networks is a highly relevant task that strongly influences the competitiveness of firms from various industries. Designing supply networks is a strategic process that considerably affects the structure of the whole network. In contrast, supply networks for new products are configured without major adaptations of the existing structure, but the network has to be configured before the new product is actually launched in the marketplace. Due to dynamics and uncertainties, the resulting planning problem is highly complex. However, formal models and solution approaches that support supply network configuration decisions for new products are scant. The paper at hand aims at stimulating related model-based research. To formulate mathematical models and solution procedures, a benchmarking problem is introduced which is derived from a case study of a cosmetics manufacturer. Tasks, objectives, and constraints of the problem are described in great detail and numerical values and ranges of all problem parameters are given. In addition, several directions for future research are suggested.

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

International Nuclear Information System (INIS)

Lago, Daniel; Rahnema, Farzad

2017-01-01

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

Measurements and FLUKA simulations of bismuth and aluminium activation at the CERN Shielding Benchmark Facility (CSBF)

Science.gov (United States)

Iliopoulou, E.; Bamidis, P.; Brugger, M.; Froeschl, R.; Infantino, A.; Kajimoto, T.; Nakao, N.; Roesler, S.; Sanami, T.; Siountas, A.

2018-03-01

The CERN High Energy AcceleRator Mixed field facility (CHARM) is located in the CERN Proton Synchrotron (PS) East Experimental Area. The facility receives a pulsed proton beam from the CERN PS with a beam momentum of 24 GeV/c with 5 ṡ1011 protons per pulse with a pulse length of 350 ms and with a maximum average beam intensity of 6.7 ṡ1010 p/s that then impacts on the CHARM target. The shielding of the CHARM facility also includes the CERN Shielding Benchmark Facility (CSBF) situated laterally above the target. This facility consists of 80 cm of cast iron and 360 cm of concrete with barite concrete in some places. Activation samples of bismuth and aluminium were placed in the CSBF and in the CHARM access corridor in July 2015. Monte Carlo simulations with the FLUKA code have been performed to estimate the specific production yields for these samples. The results estimated by FLUKA Monte Carlo simulations are compared to activation measurements of these samples. The comparison between FLUKA simulations and the measured values from γ-spectrometry gives an agreement better than a factor of 2.

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)

Radiation shielding quality assurance

Science.gov (United States)

Um, Dallsun

For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

Single pin BWR benchmark problem for coupled Monte Carlo - Thermal hydraulics analysis

International Nuclear Information System (INIS)

Ivanov, A.; Sanchez, V.; Hoogenboom, J. E.

2012-01-01

As part of the European NURISP research project, a single pin BWR benchmark problem was defined. The aim of this initiative is to test the coupling strategies between Monte Carlo and subchannel codes developed by different project participants. In this paper the results obtained by the Delft Univ. of Technology and Karlsruhe Inst. of Technology will be presented. The benchmark problem was simulated with the following coupled codes: TRIPOLI-SUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHANFLOW, and KENO-SUBCHANFLOW. (authors)

Single pin BWR benchmark problem for coupled Monte Carlo - Thermal hydraulics analysis

Energy Technology Data Exchange (ETDEWEB)

Ivanov, A.; Sanchez, V. [Karlsruhe Inst. of Technology, Inst. for Neutron Physics and Reactor Technology, Herman-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Hoogenboom, J. E. [Delft Univ. of Technology, Faculty of Applied Sciences, Mekelweg 15, 2629 JB Delft (Netherlands)

2012-07-01

As part of the European NURISP research project, a single pin BWR benchmark problem was defined. The aim of this initiative is to test the coupling strategies between Monte Carlo and subchannel codes developed by different project participants. In this paper the results obtained by the Delft Univ. of Technology and Karlsruhe Inst. of Technology will be presented. The benchmark problem was simulated with the following coupled codes: TRIPOLI-SUBCHANFLOW, MCNP-FLICA, MCNP-SUBCHANFLOW, and KENO-SUBCHANFLOW. (authors)

Benchmark problems for radiological assessment codes. Final report

International Nuclear Information System (INIS)

Mills, M.; Vogt, D.; Mann, B.

1983-09-01

This report describes benchmark problems to test computer codes used in the radiological assessment of high-level waste repositories. The problems presented in this report will test two types of codes. The first type of code calculates the time-dependent heat generation and radionuclide inventory associated with a high-level waste package. Five problems have been specified for this code type. The second code type addressed in this report involves the calculation of radionuclide transport and dose-to-man. For these codes, a comprehensive problem and two subproblems have been designed to test the relevant capabilities of these codes for assessing a high-level waste repository setting

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

Workshops and problems for benchmarking eddy current codes

Energy Technology Data Exchange (ETDEWEB)

Turner, L.R.; Davey, K.; Ida, N.; Rodger, D.; Kameari, A.; Bossavit, A.; Emson, C.R.I.

1988-08-01

A series of six workshops was held in 1986 and 1987 to compare eddy current codes, using six benchmark problems. The problems included transient and steady-state ac magnetic fields, close and far boundary conditions, magnetic and non-magnetic materials. All the problems were based either on experiments or on geometries that can be solved analytically. The workshops and solutions to the problems are described. Results show that many different methods and formulations give satisfactory solutions, and that in many cases reduced dimensionality or coarse discretization can give acceptable results while reducing the computer time required. A second two-year series of TEAM (Testing Electromagnetic Analysis Methods) workshops, using six more problems, is underway. 12 refs., 15 figs., 4 tabs.

Workshops and problems for benchmarking eddy current codes

International Nuclear Information System (INIS)

Turner, L.R.; Davey, K.; Ida, N.; Rodger, D.; Kameari, A.; Bossavit, A.; Emson, C.R.I.

1988-08-01

A series of six workshops was held in 1986 and 1987 to compare eddy current codes, using six benchmark problems. The problems included transient and steady-state ac magnetic fields, close and far boundary conditions, magnetic and non-magnetic materials. All the problems were based either on experiments or on geometries that can be solved analytically. The workshops and solutions to the problems are described. Results show that many different methods and formulations give satisfactory solutions, and that in many cases reduced dimensionality or coarse discretization can give acceptable results while reducing the computer time required. A second two-year series of TEAM (Testing Electromagnetic Analysis Methods) workshops, using six more problems, is underway. 12 refs., 15 figs., 4 tabs

Three anisotropic benchmark problems for adaptive finite element methods

Czech Academy of Sciences Publication Activity Database

Šolín, Pavel; Čertík, O.; Korous, L.

2013-01-01

Roč. 219, č. 13 (2013), s. 7286-7295 ISSN 0096-3003 R&D Projects: GA AV ČR IAA100760702 Institutional support: RVO:61388998 Keywords : benchmark problem * anisotropic solution * boundary layer Subject RIV: BA - General Mathematics Impact factor: 1.600, year: 2013

Benchmark Problems of the Geothermal Technologies Office Code Comparison Study

Energy Technology Data Exchange (ETDEWEB)

White, Mark D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Podgorney, Robert [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kelkar, Sharad M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); McClure, Mark W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Danko, George [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ghassemi, Ahmad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fu, Pengcheng [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bahrami, Davood [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Barbier, Charlotte [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cheng, Qinglu [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chiu, Kit-Kwan [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Detournay, Christine [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Elsworth, Derek [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fang, Yi [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Furtney, Jason K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gan, Quan [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gao, Qian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Guo, Bin [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hao, Yue [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Horne, Roland N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Huang, Kai [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Im, Kyungjae [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Norbeck, Jack [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rutqvist, Jonny [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Safari, M. R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sesetty, Varahanaresh [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sonnenthal, Eric [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tao, Qingfeng [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); White, Signe K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wong, Yang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Xia, Yidong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-12-02

A diverse suite of numerical simulators is currently being applied to predict or understand the performance of enhanced geothermal systems (EGS). To build confidence and identify critical development needs for these analytical tools, the United States Department of Energy, Geothermal Technologies Office has sponsored a Code Comparison Study (GTO-CCS), with participants from universities, industry, and national laboratories. A principal objective for the study was to create a community forum for improvement and verification of numerical simulators for EGS modeling. Teams participating in the study were those representing U.S. national laboratories, universities, and industries, and each team brought unique numerical simulation capabilities to bear on the problems. Two classes of problems were developed during the study, benchmark problems and challenge problems. The benchmark problems were structured to test the ability of the collection of numerical simulators to solve various combinations of coupled thermal, hydrologic, geomechanical, and geochemical processes. This class of problems was strictly defined in terms of properties, driving forces, initial conditions, and boundary conditions. Study participants submitted solutions to problems for which their simulation tools were deemed capable or nearly capable. Some participating codes were originally developed for EGS applications whereas some others were designed for different applications but can simulate processes similar to those in EGS. Solution submissions from both were encouraged. In some cases, participants made small incremental changes to their numerical simulation codes to address specific elements of the problem, and in other cases participants submitted solutions with existing simulation tools, acknowledging the limitations of the code. The challenge problems were based on the enhanced geothermal systems research conducted at Fenton Hill, near Los Alamos, New Mexico, between 1974 and 1995. The problems

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

Energy Technology Data Exchange (ETDEWEB)

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

1979-06-01

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

Parameters calculation of shielding experiment

International Nuclear Information System (INIS)

Gavazza, S.

1986-02-01

The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author) [pt

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

International Nuclear Information System (INIS)

Pescarini, M.; Borgia, M. G.

1997-03-01

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

Development of Neutron Shielding Material for Cask and Accelerator

International Nuclear Information System (INIS)

Kang, Hee Young; Seo, Ki Seog; Lee, Byung Chul; Park, Chang Jae; Kim, Ho Dong

2008-01-01

The neutron shielding materials are used as a neutron shield for spent fuel shipping cask, beam accelerators and neutron generators. At early stage, the neutron attenuations of materials were evaluated with the cross sections. After that, benchmark or mock-up experiments on the multi-layer problem to confirm the shielding characteristics or to evaluate analysis accuracy were reported. Recently, the need to transport spent nuclear fuels is increasing due to the current limited storage capacity. The on-site storage capacity at some of nuclear power plants is expected to be full in near future. With a growing inventory of spent fuels at power plants, these spent fuels need to be transported to other storage facilities. Shipping casks have been developed to safely transport spent fuels that emit high neutrons and gamma-ray radiation. The external radiation level of the shipping cask from the spent fuel must be limited to meet the standards specified by the IAEA radioactive material package regulation, so it is important to develop a proper neutron shielding material for a shipping cask. Neutron shielding experiments and analyses on the shielding effects of materials have been conducted, and some experiments have been performed to examine the shielding effects of selected materials. The shielding experiments consist of evaluating not only the shielding effects of a material alone but also the effects of the material thickness. The experimental results were compared with those obtained by using the MCNP-5c code

Status of reactor-shielding research in the US

International Nuclear Information System (INIS)

Maienshein, F.C.

1980-01-01

While reactor programs change, shielding analysis methods are improved slowly. Version-V of ENDF/B provides improved data and Version-VI will be cost effective in advanced fission reactors are to be developed in the US. Benchmarks for data and methods validation are collected and distributed in the US in two series, one primarily for FBR-related experiments and one for LWR calculational methods. For LWR design, cavity streaming is now handled adequately, if with varying degrees of elegance. Investigations of improved detector response for LWRs rely upon transport methods. The great potential importance of pressure-vessel damage is dreflected in widespread studies to aid in the prediction of neutron fluences in vessels. For LMFBRS, the FFTF should give attenuation results on an operating reactor. For larger power reactors, the advantages of alternate shield materials appear compelling. A few other shielding studies appear to require experimental confirmation if LMFBRs are to be economically competitive. A coherent shielding program for the GCFR is nearing completion. For the fusion-reactor program, methods verification is under way, practical calculations are well advanced for test devices such as the TFTR and FMIT, and consideration is now given to shielding problems of large reactors, as in the ETF study

Benchmark shielding calculations for the NEACRP [Nuclear Energy Agency-Committee on Reactor Physics] Working Group on shielding assessment of transportation packages

International Nuclear Information System (INIS)

Broadhead, B.L.; Brady, M.C.; Parks, C.V.

1990-11-01

In 1985, the Nuclear Energy Agency-Committee on Reactor Physics (NEACRP) established a working group on shielding assessment of transportation packages. Following the initial distribution of a set of six problems, discussions were held at the Organization for Economic Cooperation and Development (OECD) Headquarters in Paris, France, in June/July 1986, May 1988, and February/March 1990. The US contribution to the working group is documented in this report. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask. Among the effects reported here are the performance of multiple cross-section sets, the comparison of several source generation codes, and multidimensional versus one-dimensional (1-D) analyses. 18 refs., 16 figs., 33 tabs

Survey of the results of a two- and three-dimensional kinetics benchmark problem typical for a thermal reactor

International Nuclear Information System (INIS)

Werner, W.

1975-01-01

In 1973, NEACRP and CSNI posed a number of kinetic benchmark problems intended to be solved by different groups. Comparison of the submitted results should lead to estimates on the accuracy and efficiency of the employed codes. This was felt to be of great value since the codes involved become more and more important in the field of reactor safety. In this paper the results of the 2d and 3d benchmark problem for a BWR are presented. The specification of the problem is included in the appendix of this survey. For the 2d benchmark problem, 5 contributions have been obtained, while for the 3d benchmark problem 2 contributions have been submitted. (orig./RW) [de

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-05-01

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

MARMER, a flexible point-kernel shielding code

International Nuclear Information System (INIS)

Kloosterman, J.L.; Hoogenboom, J.E.

1990-01-01

A point-kernel shielding code entitled MARMER is described. It has several options with respect to geometry input, source description and detector point description which extend the flexibility and usefulness of the code, and which are especially useful in spent fuel shielding. MARMER has been validated using the TN12 spent fuel shipping cask benchmark. (author)

MARMER, a flexible point-kernel shielding code

Energy Technology Data Exchange (ETDEWEB)

Kloosterman, J.L.; Hoogenboom, J.E. (Interuniversitair Reactor Inst., Delft (Netherlands))

1990-01-01

A point-kernel shielding code entitled MARMER is described. It has several options with respect to geometry input, source description and detector point description which extend the flexibility and usefulness of the code, and which are especially useful in spent fuel shielding. MARMER has been validated using the TN12 spent fuel shipping cask benchmark. (author).

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

Development of solutions to benchmark piping problems

Energy Technology Data Exchange (ETDEWEB)

Reich, M; Chang, T Y; Prachuktam, S; Hartzman, M

1977-12-01

Benchmark problems and their solutions are presented. The problems consist in calculating the static and dynamic response of selected piping structures subjected to a variety of loading conditions. The structures range from simple pipe geometries to a representative full scale primary nuclear piping system, which includes the various components and their supports. These structures are assumed to behave in a linear elastic fashion only, i.e., they experience small deformations and small displacements with no existing gaps, and remain elastic through their entire response. The solutions were obtained by using the program EPIPE, which is a modification of the widely available program SAP IV. A brief outline of the theoretical background of this program and its verification is also included.

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

Validation of flexible multibody dynamics beam formulations using benchmark problems

Energy Technology Data Exchange (ETDEWEB)

Bauchau, Olivier A., E-mail: obauchau@umd.edu [University of Maryland (United States); Betsch, Peter [Karlsruhe Institute of Technology (Germany); Cardona, Alberto [CIMEC (UNL/Conicet) (Argentina); Gerstmayr, Johannes [Leopold-Franzens Universität Innsbruck (Austria); Jonker, Ben [University of Twente (Netherlands); Masarati, Pierangelo [Politecnico di Milano (Italy); Sonneville, Valentin [Université de Liège (Belgium)

2016-05-15

As the need to model flexibility arose in multibody dynamics, the floating frame of reference formulation was developed, but this approach can yield inaccurate results when elastic displacements becomes large. While the use of three-dimensional finite element formulations overcomes this problem, the associated computational cost is overwhelming. Consequently, beam models, which are one-dimensional approximations of three-dimensional elasticity, have become the workhorse of many flexible multibody dynamics codes. Numerous beam formulations have been proposed, such as the geometrically exact beam formulation or the absolute nodal coordinate formulation, to name just two. New solution strategies have been investigated as well, including the intrinsic beam formulation or the DAE approach. This paper provides a systematic comparison of these various approaches, which will be assessed by comparing their predictions for four benchmark problems. The first problem is the Princeton beam experiment, a study of the static large displacement and rotation behavior of a simple cantilevered beam under a gravity tip load. The second problem, the four-bar mechanism, focuses on a flexible mechanism involving beams and revolute joints. The third problem investigates the behavior of a beam bent in its plane of greatest flexural rigidity, resulting in lateral buckling when a critical value of the transverse load is reached. The last problem investigates the dynamic stability of a rotating shaft. The predictions of eight independent codes are compared for these four benchmark problems and are found to be in close agreement with each other and with experimental measurements, when available.

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Automatic generation of biasing parameters for MORSE shielding problems

International Nuclear Information System (INIS)

Hoogenboom, J.E.

1995-01-01

It would be favourable if the biasing functions could be obtained from the Monte Carlo calculation itself. This is discussed in this paper as well as the way to obtain biasing parameters from it for splitting, Russian roulette and path length stretching. The method is demonstrated for a shielding problem solved with the MORSE-SGC/S Monte Carlo code of the SCALE-system. (K.A.)

Merton's problem for an investor with a benchmark in a Barndorff-Nielsen and Shephard market.

Science.gov (United States)

Lennartsson, Jan; Lindberg, Carl

2015-01-01

To try to outperform an externally given benchmark with known weights is the most common equity mandate in the financial industry. For quantitative investors, this task is predominantly approached by optimizing their portfolios consecutively over short time horizons with one-period models. We seek in this paper to provide a theoretical justification to this practice when the underlying market is of Barndorff-Nielsen and Shephard type. This is done by verifying that an investor who seeks to maximize her expected terminal exponential utility of wealth in excess of her benchmark will in fact use an optimal portfolio equivalent to the one-period Markowitz mean-variance problem in continuum under the corresponding Black-Scholes market. Further, we can represent the solution to the optimization problem as in Feynman-Kac form. Hence, the problem, and its solution, is analogous to Merton's classical portfolio problem, with the main difference that Merton maximizes expected utility of terminal wealth, not wealth in excess of a benchmark.

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-10-01

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

Proceedings of a meeting on radiation shielding and related topics

International Nuclear Information System (INIS)

1978-01-01

This is a proceedings of a meeting on radiation shielding and related topics held on Feb. 22 and 23 in 1978 at Nuclear Engineering Research Laboratory of University of Tokyo. The reports includes the following items (1) studies on neutronics with accelerators (2) radiation damage (3) shielding design (4) radiation streaming (5) shielding experiments from a point of view of radiation measurements (6) shielding benchmark experiments (7) prospects on the study of neutronics. All items are written in Japanese. (auth.)

Adaptive unified continuum FEM modeling of a 3D FSI benchmark problem.

Science.gov (United States)

Jansson, Johan; Degirmenci, Niyazi Cem; Hoffman, Johan

2017-09-01

In this paper, we address a 3D fluid-structure interaction benchmark problem that represents important characteristics of biomedical modeling. We present a goal-oriented adaptive finite element methodology for incompressible fluid-structure interaction based on a streamline diffusion-type stabilization of the balance equations for mass and momentum for the entire continuum in the domain, which is implemented in the Unicorn/FEniCS software framework. A phase marker function and its corresponding transport equation are introduced to select the constitutive law, where the mesh tracks the discontinuous fluid-structure interface. This results in a unified simulation method for fluids and structures. We present detailed results for the benchmark problem compared with experiments, together with a mesh convergence study. Copyright © 2016 John Wiley & Sons, Ltd.

Effect of interpolation error in pre-processing codes on calculations of self-shielding factors and their temperature derivatives

International Nuclear Information System (INIS)

Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullan, D.E.

1986-01-01

We investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. We consider the 2.0347 to 3.3546 keV energy region for 238 U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems. (author)

Effect of interpolation error in pre-processing codes on calculations of self-shielding factors and their temperature derivatives

International Nuclear Information System (INIS)

Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullen, D.E.

1985-01-01

The authors investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. They consider the 2.0347 to 3.3546 keV energy region for /sup 238/U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems

A highly simplified 3D BWR benchmark problem

International Nuclear Information System (INIS)

Douglass, Steven; Rahnema, Farzad

2010-01-01

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

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

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

International Nuclear Information System (INIS)

Downar, T.

2009-01-01

The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multidimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. Specifically, the methods here utilize the existing continuous energy SCALE5 module, CENTRM, and the multi-dimensional discrete ordinates solver, NEWT to develop a new code, CENTRM( ) NEWT. The work here addresses specific theoretical limitations in existing CENTRM resonance treatment, as well as investigates advanced numerical and parallel computing algorithms for CENTRM and NEWT in order to reduce the computational burden. The result of the work here will be a new computer code capable of performing problem dependent self-shielding analysis for both existing and proposed GENIV fuel designs. The objective of the work was to have an immediate impact on the safety analysis of existing reactors through improvements in the calculation of fuel temperature effects, as well as on the analysis of more sophisticated GENIV/NGNP systems through improvements in the depletion/transmutation of actinides for Advanced Fuel Cycle Initiatives.

Radiation shielding activities at the OECD/Nuclear Energy Agency

International Nuclear Information System (INIS)

Sartori, Enrico; Vaz, Pedro

2000-01-01

The OECD Nuclear Energy Agency (NEA) has devoted considerable effort over the years to radiation shielding issues. The issues are addressed through international working groups. These activities are carried out in close co-ordination and co-operation with the Radiation Safety Information Computational Center (RSICC). The areas of work include: basic nuclear data activities in support of radiation shielding, computer codes, shipping cask shielding applications, reactor pressure vessel dosimetry, shielding experiments database. The method of work includes organising international code comparison exercises and benchmark studies. Training courses on radiation shielding computer codes are organised regularly including hands-on experience in modelling skills. The scope of the activity covers mainly reactor shields and spent fuel transportation packages, but also fusion neutronics and in particular shielding of accelerators and irradiation facilities. (author)

Two-dimensional shielding benchmarks for iron at YAYOI, (1)

International Nuclear Information System (INIS)

Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.

The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)

Optimization of multi-layered metallic shield

International Nuclear Information System (INIS)

Ben-Dor, G.; Dubinsky, A.; Elperin, T.

2011-01-01

Research highlights: → We investigated the problem of optimization of a multi-layered metallic shield. → The maximum ballistic limit velocity is a criterion of optimization. → The sequence of materials and the thicknesses of layers in the shield are varied. → The general problem is reduced to the problem of Geometric Programming. → Analytical solutions are obtained for two- and three-layered shields. - Abstract: We investigate the problem of optimization of multi-layered metallic shield whereby the goal is to determine the sequence of materials and the thicknesses of the layers that provide the maximum ballistic limit velocity of the shield. Optimization is performed under the following constraints: fixed areal density of the shield, the upper bound on the total thickness of the shield and the bounds on the thicknesses of the plates manufactured from every material. The problem is reduced to the problem of Geometric Programming which can be solved numerically using known methods. For the most interesting in practice cases of two-layered and three-layered shields the solution is obtained in the explicit analytical form.

Benchmarking time-dependent neutron problems with Monte Carlo codes

International Nuclear Information System (INIS)

Couet, B.; Loomis, W.A.

1990-01-01

Many nuclear logging tools measure the time dependence of a neutron flux in a geological formation to infer important properties of the formation. The complex geometry of the tool and the borehole within the formation does not permit an exact deterministic modelling of the neutron flux behaviour. While this exact simulation is possible with Monte Carlo methods the computation time does not facilitate quick turnaround of results useful for design and diagnostic purposes. Nonetheless a simple model based on the diffusion-decay equation for the flux of neutrons of a single energy group can be useful in this situation. A combination approach where a Monte Carlo calculation benchmarks a deterministic model in terms of the diffusion constants of the neutrons propagating in the media and their flux depletion rates thus offers the possibility of quick calculation with assurance as to accuracy. We exemplify this approach with the Monte Carlo benchmarking of a logging tool problem, showing standoff and bedding response. (author)

Survivor shielding. Part C. Improvements in terrain shielding

International Nuclear Information System (INIS)

Egbert, Stephen D.; Kaul, Dean C.; Roberts, James A.; Kerr, George D.

2005-01-01

A number of atomic-bomb survivors were affected by shielding provided by terrain features. These terrain features can be a small hill, affecting one or two houses, or a high mountain that shields large neighborhoods. In the survivor dosimetry system, terrain shielding can be described by a transmission factor (TF), which is the ratio between the dose with and without the terrain present. The terrain TF typically ranges between 0.1 and 1.0. After DS86 was implemented at RERF, the terrain shielding categories were examined and found to either have a bias or an excessive uncertainty that could readily be removed. In 1989, an improvement in the terrain model was implemented at RERF in the revised DS86 code, but the documentation was not published. It is now presented in this section. The solution to the terrain shielding in front of a house is described in this section. The problem of terrain shielding of survivors behind Hijiyama mountain at Hiroshima and Konpirasan mountain at Nagasaki has also been recognized, and a solution to this problem has been included in DS02. (author)

Shielding practice

International Nuclear Information System (INIS)

Sauermann, P.F.

1985-08-01

The basis of shielding practice against external irradiation is shown in a simple way. For most sources of radiation (point sources) occurring in shielding practice, the basic data are given, mainly in the form of tables, which are required to solve the shielding problems. The application of these data is explained and discussed using practical examples. Thickness of shielding panes of glove boxes for α and β radiation; shielding of sealed γ-radiography sources; shielding of a Co-60 radiation source, and of the manipulator panels for hot cells; damping factors for γ radiation and neutrons; shielding of fast and thermal neutrons, and of bremsstrahlung (X-ray tubes, Kr-85 pressure gas cylinders, 42 MeV betatrons, 20 MeV linacs); two-fold shielding (lead glass windows for hot cells, 14 MeV neutron generators); shielding against scattered radiation. (orig./HP) [de

Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

International Nuclear Information System (INIS)

Bozic, M.; Zagar, T.; Ravnik, M.

2001-01-01

The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

Generation and testing of the shielding data library EURLIB for fission and fusion technology

International Nuclear Information System (INIS)

Caglioti, E.; Hehn, G.; Herrnberger, V.; Mattes, M.; Nicks, R.; Penkuhn, H.

1977-01-01

For the common field of core physics and shielding, the CSEWG group structure of 239 fast neutron groups had been proposed, of which the 100 neutron groups of the EURLIB Library is a sub-set for shielding. This standard group Library EURLIB had been initiated by the NEA-specialist group on shielding benchmarks in 1974. The wide acceptance of the Library for interpretation of benchmarks in the NEA program represents an important step forward in the standardization of group data which is the basic requirement for a useful collaboration. On the other side the interpretation of a series of different benchmark experiments with the EURLIB Library provides the best check of the cross section data for neutron and gamma-rays showing the needs for further improvements. The paper describes the joint work of IKE, Stuttgart and EURATOM, Ispra in generating multigroup libraries for neutron and gamma-rays. Special effort has been devoted to improve the flux weighting for both types of radiation and proper treatment of thermal neutrons. The coupled multigroup Library of 100 neutron and 20 gamma groups is collapsed into few group structures for typical designs of LWR, LMFBR, gas cooled and thermonuclear reactors. The work for optimal few group representation is done in cooperation with EIR, Wurenlingen. The testing of the EURLIB Library is a common effort of several institutions participating in the NEA shielding benchmark program

Self-shielding models of MICROX-2 code: Review and updates

International Nuclear Information System (INIS)

Hou, J.; Choi, H.; Ivanov, K.N.

2014-01-01

Highlights: • The MICROX-2 code has been improved to expand its application to advanced reactors. • New fine-group cross section libraries based on ENDF/B-VII have been generated. • Resonance self-shielding and spatial self-shielding models have been improved. • The improvements were assessed by a series of benchmark calculations against MCNPX. - Abstract: The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. The MICROX-2 code has been updated to expand its application to advanced reactor concepts and fuel cycle simulations, including generation of new fine-group cross section libraries based on ENDF/B-VII. In continuation of previous work, the MICROX-2 methods are reviewed and updated in this study, focusing on its resonance self-shielding and spatial self-shielding models for neutron spectrum calculations. The improvement of self-shielding method was assessed by a series of benchmark calculations against the Monte Carlo code, using homogeneous and heterogeneous pin cell models. The results have shown that the implementation of the updated self-shielding models is correct and the accuracy of physics calculation is improved. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by ∼0.1% and ∼0.2% for the homogeneous and heterogeneous pin cell models, respectively, considered in this study

Benchmarking shielding simulations for an accelerator-driven spallation neutron source

Directory of Open Access Journals (Sweden)

Nataliia Cherkashyna

2015-08-01

Full Text Available The shielding at an accelerator-driven spallation neutron facility plays a critical role in the performance of the neutron scattering instruments, the overall safety, and the total cost of the facility. Accurate simulation of shielding components is thus key for the design of upcoming facilities, such as the European Spallation Source (ESS, currently in construction in Lund, Sweden. In this paper, we present a comparative study between the measured and the simulated neutron background at the Swiss Spallation Neutron Source (SINQ, at the Paul Scherrer Institute (PSI, Villigen, Switzerland. The measurements were carried out at several positions along the SINQ monolith wall with the neutron dosimeter WENDI-2, which has a well-characterized response up to 5 GeV. The simulations were performed using the Monte-Carlo radiation transport code geant4, and include a complete transport from the proton beam to the measurement locations in a single calculation. An agreement between measurements and simulations is about a factor of 2 for the points where the measured radiation dose is above the background level, which is a satisfactory result for such simulations spanning many energy regimes, different physics processes and transport through several meters of shielding materials. The neutrons contributing to the radiation field emanating from the monolith were confirmed to originate from neutrons with energies above 1 MeV in the target region. The current work validates geant4 as being well suited for deep-shielding calculations at accelerator-based spallation sources. We also extrapolate what the simulated flux levels might imply for short (several tens of meters instruments at ESS.

Double-layer neutron shield design as neutron shielding application

Science.gov (United States)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

The shield design in particle accelerators and other high energy facilities are mainly connected to the high-energy neutrons. The deep penetration of neutrons through massive shield has become a very serious problem. For shielding to be efficient, most of these neutrons should be confined to the shielding volume. If the interior space will become limited, the sufficient thickness of multilayer shield must be used. Concrete and iron are widely used as a multilayer shield material. Two layers shield material was selected to guarantee radiation safety outside of the shield against neutrons generated in the interaction of the different proton energies. One of them was one meter of concrete, the other was iron-contained material (FeB, Fe2B and stainless-steel) to be determined shield thicknesses. FLUKA Monte Carlo code was used for shield design geometry and required neutron dose distributions. The resulting two layered shields are shown better performance than single used concrete, thus the shield design could leave more space in the interior shielded areas.

Using FLUKA to Study Concrete Square Shield Performance in Attenuation of Neutron Radiation Produced by APF Plasma Focus Neutron Source

Science.gov (United States)

Nemati, M. J.; Habibi, M.; Amrollahi, R.

2013-04-01

In 2010, representatives from the Nuclear Engineering and physics Department of Amirkabir University of Technology (AUT) requested development of a project with the objective of determining the performance of a concrete shield for their Plasma Focus as neutron source. The project team in Laboratory of Nuclear Engineering and physics department of Amirkabir University of Technology choose some shape of shield to study on their performance with Monte Carlo code. In the present work, the capability of Monte Carlo code FLUKA will be explored to model the APF Plasma Focus, and investigating the neutron fluence on the square concrete shield in each region of problem. The physical models embedded in FLUKA are mentioned, as well as examples of benchmarking against future experimental data. As a result of this study suitable thickness of concrete for shielding APF will be considered.

Detailed benchmark test of JENDL-4.0 iron data for fusion applications

Energy Technology Data Exchange (ETDEWEB)

Konno, Chikara, E-mail: konno.chikara@jaea.go.jp [Japan Atomic Energy Agency, Tokai-Mura, Ibaraki-ken, 319-1195 (Japan); Wada, Masayuki [Japan Computer System, Mito, 310-0805 (Japan); Kondo, Keitaro; Ohnishi, Seiki; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi [Japan Atomic Energy Agency, Tokai-Mura, Ibaraki-ken, 319-1195 (Japan)

2011-10-15

The major revised version of Japanese Evaluated Nuclear Data Library (JENDL), JENDL-4.0, was released in May, 2010. As one of benchmark tests, we have carried out a benchmark test of JENDL-4.0 iron data, which are very important for radiation shielding in fusion reactors, by analyzing the iron fusion neutronics integral experiments (in situ and Time-of-Flight (TOF) experiments) at JAEA/FNS. It is demonstrated that the problems of the iron data in the previous version of JENDL, JENDL-3.3, are solved in JENDL-4.0; the first inelastic scattering cross section data of {sup 57}Fe and the angular distribution of the elastic scattering of {sup 56}Fe. The iron data in JENDL-4.0 are comparable to or are partly better than those in ENDF/B-VII.0 and JEFF-3.1.

Calculation of parameters for an iron shield experiment

International Nuclear Information System (INIS)

Gavazza, S.

1986-01-01

In this text is carreid out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gama-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The tranpsort calculations were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reaction and doses rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented reasonable concordance with the experimental measurements. Finally, is presented a proposal for setting up of an experimental arrangement, using the IEA-R1 reactor, with the purpose of lay down a shielding benchmark. (Author) [pt

Nodal deterministic simulation for problems of neutron shielding in multigroup formulation

International Nuclear Information System (INIS)

Baptista, Josue Costa; Heringer, Juan Diego dos Santos; Santos, Luiz Fernando Trindade; Alves Filho, Hermes

2013-01-01

In this paper, we propose the use of some computational tools, with the implementation of numerical methods SGF (Spectral Green's Function), making use of a deterministic model of transport of neutral particles in the study and analysis of a known and simplified problem of nuclear engineering, known in the literature as a problem of neutron shielding, considering the model with two energy groups. These simulations are performed in MatLab platform, version 7.0, and are presented and developed with the help of a Computer Simulator providing a friendly computer application for their utilities

Benchmark analysis of MCNP trademark ENDF/B-VI iron

International Nuclear Information System (INIS)

Court, J.D.; Hendricks, J.S.

1994-12-01

The MCNP ENDF/B-VI iron cross-section data was subjected to four benchmark studies as part of the Hiroshima/Nagasaki dose re-evaluation for the National Academy of Science and the Defense Nuclear Agency. The four benchmark studies were: (1) the iron sphere benchmarks from the Lawrence Livermore Pulsed Spheres; (2) the Oak Ridge National Laboratory Fusion Reactor Shielding Benchmark; (3) a 76-cm diameter iron sphere benchmark done at the University of Illinois; (4) the Oak Ridge National Laboratory Benchmark for Neutron Transport through Iron. MCNP4A was used to model each benchmark and computational results from the ENDF/B-VI iron evaluations were compared to ENDF/B-IV, ENDF/B-V, the MCNP Recommended Data Set (which includes Los Alamos National Laboratory Group T-2 evaluations), and experimental data. The results show that the ENDF/B-VI iron evaluations are as good as, or better than, previous data sets

Mesoscale Benchmark Demonstration Problem 1: Mesoscale Simulations of Intra-granular Fission Gas Bubbles in UO2 under Post-irradiation Thermal Annealing

Energy Technology Data Exchange (ETDEWEB)

Li, Yulan; Hu, Shenyang Y.; Montgomery, Robert; Gao, Fei; Sun, Xin; Tonks, Michael; Biner, Bullent; Millet, Paul; Tikare, Veena; Radhakrishnan, Balasubramaniam; Andersson , David

2012-04-11

A study was conducted to evaluate the capabilities of different numerical methods used to represent microstructure behavior at the mesoscale for irradiated material using an idealized benchmark problem. The purpose of the mesoscale benchmark problem was to provide a common basis to assess several mesoscale methods with the objective of identifying the strengths and areas of improvement in the predictive modeling of microstructure evolution. In this work, mesoscale models (phase-field, Potts, and kinetic Monte Carlo) developed by PNNL, INL, SNL, and ORNL were used to calculate the evolution kinetics of intra-granular fission gas bubbles in UO2 fuel under post-irradiation thermal annealing conditions. The benchmark problem was constructed to include important microstructural evolution mechanisms on the kinetics of intra-granular fission gas bubble behavior such as the atomic diffusion of Xe atoms, U vacancies, and O vacancies, the effect of vacancy capture and emission from defects, and the elastic interaction of non-equilibrium gas bubbles. An idealized set of assumptions was imposed on the benchmark problem to simplify the mechanisms considered. The capability and numerical efficiency of different models are compared against selected experimental and simulation results. These comparisons find that the phase-field methods, by the nature of the free energy formulation, are able to represent a larger subset of the mechanisms influencing the intra-granular bubble growth and coarsening mechanisms in the idealized benchmark problem as compared to the Potts and kinetic Monte Carlo methods. It is recognized that the mesoscale benchmark problem as formulated does not specifically highlight the strengths of the discrete particle modeling used in the Potts and kinetic Monte Carlo methods. Future efforts are recommended to construct increasingly more complex mesoscale benchmark problems to further verify and validate the predictive capabilities of the mesoscale modeling

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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.

Using the MCNP Taylor series perturbation feature (efficiently) for shielding problems

Science.gov (United States)

Favorite, Jeffrey

2017-09-01

The Taylor series or differential operator perturbation method, implemented in MCNP and invoked using the PERT card, can be used for efficient parameter studies in shielding problems. This paper shows how only two PERT cards are needed to generate an entire parameter study, including statistical uncertainty estimates (an additional three PERT cards can be used to give exact statistical uncertainties). One realistic example problem involves a detailed helium-3 neutron detector model and its efficiency as a function of the density of its high-density polyethylene moderator. The MCNP differential operator perturbation capability is extremely accurate for this problem. A second problem involves the density of the polyethylene reflector of the BeRP ball and is an example of first-order sensitivity analysis using the PERT capability. A third problem is an analytic verification of the PERT capability.

Problems related to the definition of the shielding of a large fast power reactor

International Nuclear Information System (INIS)

Moreau, J.

Solutions for the shielding of a 1000 MW(e) power plant in the same technological line as Phenix are given. They have been evaluated with a monodimensional transport code. The choice is based on the comparison of their efficiency towards neutrons and on the consequences of their characteristics on the conception of the reactor tank. A few economical considerations give an idea of the influence of the choice in shielding on the cost of the power plant. At last the problem of the optimization possibilities is approached from the designer's point of view

A Comparative Study of Differential Evolution, Particle Swarm Optimization, and Evolutionary Algorithms on Numerical Benchmark Problems

DEFF Research Database (Denmark)

Vesterstrøm, Jacob Svaneborg; Thomsen, Rene

2004-01-01

Several extensions to evolutionary algorithms (EAs) and particle swarm optimization (PSO) have been suggested during the last decades offering improved performance on selected benchmark problems. Recently, another search heuristic termed differential evolution (DE) has shown superior performance...... in several real-world applications. In this paper, we evaluate the performance of DE, PSO, and EAs regarding their general applicability as numerical optimization techniques. The comparison is performed on a suite of 34 widely used benchmark problems. The results from our study show that DE generally...... outperforms the other algorithms. However, on two noisy functions, both DE and PSO were outperformed by the EA....

Review of international solutions to NEACRP benchmark BWR lattice cell problems

International Nuclear Information System (INIS)

Halsall, M.J.

1977-12-01

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

Validation of the 3D finite element transport theory code EVENT for shielding applications

International Nuclear Information System (INIS)

Warner, Paul; Oliveira, R.E. de

2000-01-01

This paper is concerned with the validation of the 3D deterministic neutral-particle transport theory code EVENT for shielding applications. The code is based on the finite element-spherical harmonics (FE-P N ) method which has been extensively developed over the last decade. A general multi-group, anisotropic scattering formalism enables the code to address realistic steady state and time dependent, multi-dimensional coupled neutron/gamma radiation transport problems involving high scattering and deep penetration alike. The powerful geometrical flexibility and competitive computational effort makes the code an attractive tool for shielding applications. In recognition of this, EVENT is currently in the process of being adopted by the UK nuclear industry. The theory behind EVENT is described and its numerical implementation is outlined. Numerical results obtained by the code are compared with predictions of the Monte Carlo code MCBEND and also with the results from benchmark shielding experiments. In particular, results are presented for the ASPIS experimental configuration for both neutron and gamma ray calculations using the BUGLE 96 nuclear data library. (author)

Shielding from cosmic radiation for interplanetary missions Active and passive methods

CERN Document Server

Spillantini, P; Durante, M; Müller-Mellin, R; Reitz, G; Rossi, L; Shurshakov, V; Sorbi, M

2007-01-01

Shielding is arguably the main countermeasure for the exposure to cosmic radiation during interplanetary exploratory missions. However, shielding of cosmic rays, both of galactic or solar origin, is problematic, because of the high energy of the charged particles involved and the nuclear fragmentation occurring in shielding materials. Although computer codes can predict the shield performance in space, there is a lack of biological and physical measurements to benchmark the codes. An attractive alternative to passive, bulk material shielding is the use of electromagnetic fields to deflect the charged particles from the spacecraft target. Active shielding concepts based on electrostatic fields, plasma, or magnetic fields have been proposed in the past years, and should be revised based on recent technological improvements. To address these issues, the European Space Agency (ESA) established a Topical Team (TT) in 2002 including European experts in the field of space radiation shielding and superconducting magn...

Benchmarking of EPRI-cell epithermal methods with the point-energy discrete-ordinates code (OZMA)

International Nuclear Information System (INIS)

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

1982-01-01

The purpose of the present study is to benchmark E-C resonance-shielding and cell-averaging methods against a rigorous deterministic solution on a fine-group level (approx. 30 groups between 1 eV and 5.5 keV). The benchmark code used is OZMA, which solves the space-dependent slowing-down equations using continuous-energy discrete ordinates or integral transport theory to produce fine-group cross sections. Results are given for three water-moderated lattices - a mixed oxide, a uranium method, and a tight-pitch high-conversion uranium oxide configuration. The latter two lattices were chosen because of the strong self shielding of the 238 U resonances

Solution of the CSEWG - STD-9 shielding standard problem using the coupled systems NJOY/AMPX-II and the JENDL-3 library

International Nuclear Information System (INIS)

Mendoncca, A.G.; Santos, A dos

1993-01-01

The CSEWG STD-9 benchmark has been analyzed to evaluate the shielding and out-of-core methodologies using the coupled systems NJOY/AMPX-II and the JENDL-3 library as well as the DOT 3.5 for the transport calculations. Good agreement between calculated and experimental values for the spatial and energetic neutron flux distributions have been found. However, the results for the reaction rates of the activation foils were discrepant even though for same cases they come into a good agreement. (author)

Mathematics and computational methods development in U.S. department of energy-sponsored research (nuclear energy research initiative and nuclear engineering education research). 5. Analysis of Angular V-Cycle Multigrid Formulation for Three-Dimensional Discrete Ordinates Shielding Problems

International Nuclear Information System (INIS)

Kucukboyaci, Vefa; Haghighat, Alireza

2001-01-01

We have developed new angular multigrid formulations, including the Simplified Angular Multigrid (SAM), Nested Iteration (NI), and V-Cycle schemes, that are compatible with the parallel environment and the adaptive differencing strategy of the PENTRAN three-dimensional parallel S N code. Using the Fourier analysis method for an infinite, homogenous medium, we have investigated the effectiveness of the V-Cycle scheme for different problem parameters including scattering ratio, spatial differencing weights, quadrature order, and mesh size. We have further investigated the effectiveness of the new schemes for practical shielding applications such as the Kobayashi benchmark problem and the boiling water reactor core shroud problem. In this paper, we summarize the angular V-Cycle scheme implemented in the PENTRAN code, the Fourier Analysis of the V-Cycle scheme, and results of convergence analysis of the V-Cycle scheme using different problem parameters. The theoretical analysis reveals that the V-Cycle scheme is effective for a large range of scattering ratios and is insensitive to mesh size. Besides the theoretical analysis, we have applied the new angular multigrid schemes to shielding problems. In comparison to the standard PCR formulation, combinations of the new angular multigrid schemes and PCR (e.g., SAM+V-Cycle+PCR) have proved to be very effective for scattering ratios in a range of 0.6 to 0.9. (authors)

ANS shielding standards for light-water reactors

International Nuclear Information System (INIS)

Trubey, D.K.

1982-01-01

The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions

Generating Importance Map for Geometry Splitting using Discrete Ordinates Code in Deep Shielding Problem

International Nuclear Information System (INIS)

Kim, Jong Woon; Lee, Young Ouk

2016-01-01

When we use MCNP code for a deep shielding problem, we prefer to use variance reduction technique such as geometry splitting, or weight window, or source biasing to have relative error within reliable confidence interval. To generate importance map for geometry splitting in MCNP calculation, we should know the track entering number and previous importance on each cells since a new importance is calculated based on these information. If a problem is deep shielding problem such that we have zero tracks entering on a cell, we cannot generate new importance map. In this case, discrete ordinates code can provide information to generate importance map easily. In this paper, we use AETIUS code as a discrete ordinates code. Importance map for MCNP is generated based on a zone average flux of AETIUS calculation. The discretization of space, angle, and energy is not necessary for MCNP calculation. This is the big merit of MCNP code compared to the deterministic code. However, deterministic code (i.e., AETIUS) can provide a rough estimate of the flux throughout a problem relatively quickly. This can help MCNP by providing variance reduction parameters. Recently, ADVANTG code is released. This is an automated tool for generating variance reduction parameters for fixed-source continuous-energy Monte Carlo simulations with MCNP5 v1.60

Benchmarking semantic web technology

CERN Document Server

García-Castro, R

2009-01-01

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

Implementation and verification of global optimization benchmark problems

Science.gov (United States)

Posypkin, Mikhail; Usov, Alexander

2017-12-01

The paper considers the implementation and verification of a test suite containing 150 benchmarks for global deterministic box-constrained optimization. A C++ library for describing standard mathematical expressions was developed for this purpose. The library automate the process of generating the value of a function and its' gradient at a given point and the interval estimates of a function and its' gradient on a given box using a single description. Based on this functionality, we have developed a collection of tests for an automatic verification of the proposed benchmarks. The verification has shown that literary sources contain mistakes in the benchmarks description. The library and the test suite are available for download and can be used freely.

Generalizable open source urban water portfolio simulation framework demonstrated using a multi-objective risk-based planning benchmark problem.

Science.gov (United States)

Trindade, B. C.; Reed, P. M.

2017-12-01

The growing access and reduced cost for computing power in recent years has promoted rapid development and application of multi-objective water supply portfolio planning. As this trend continues there is a pressing need for flexible risk-based simulation frameworks and improved algorithm benchmarking for emerging classes of water supply planning and management problems. This work contributes the Water Utilities Management and Planning (WUMP) model: a generalizable and open source simulation framework designed to capture how water utilities can minimize operational and financial risks by regionally coordinating planning and management choices, i.e. making more efficient and coordinated use of restrictions, water transfers and financial hedging combined with possible construction of new infrastructure. We introduce the WUMP simulation framework as part of a new multi-objective benchmark problem for planning and management of regionally integrated water utility companies. In this problem, a group of fictitious water utilities seek to balance the use of the mentioned reliability driven actions (e.g., restrictions, water transfers and infrastructure pathways) and their inherent financial risks. Several traits of this problem make it ideal for a benchmark problem, namely the presence of (1) strong non-linearities and discontinuities in the Pareto front caused by the step-wise nature of the decision making formulation and by the abrupt addition of storage through infrastructure construction, (2) noise due to the stochastic nature of the streamflows and water demands, and (3) non-separability resulting from the cooperative formulation of the problem, in which decisions made by stakeholder may substantially impact others. Both the open source WUMP simulation framework and its demonstration in a challenging benchmarking example hold value for promoting broader advances in urban water supply portfolio planning for regions confronting change.

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

OpenAIRE

Pescarini Massimo; Orsi Roberto; Frisoni Manuela

2016-01-01

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

Computational benchmark problems: a review of recent work within the American Nuclear Society Mathematics and Computation Division

International Nuclear Information System (INIS)

Dodds, H.L. Jr.

1977-01-01

An overview of the recent accomplishments of the Computational Benchmark Problems Committee of the American Nuclear Society Mathematics and Computation Division is presented. Solutions of computational benchmark problems in the following eight areas are presented and discussed: (a) high-temperature gas-cooled reactor neutronics, (b) pressurized water reactor (PWR) thermal hydraulics, (c) PWR neutronics, (d) neutron transport in a cylindrical ''black'' rod, (e) neutron transport in a boiling water reactor (BWR) rod bundle, (f) BWR transient neutronics with thermal feedback, (g) neutron depletion in a heavy water reactor, and (h) heavy water reactor transient neutronics. It is concluded that these problems and solutions are of considerable value to the nuclear industry because they have been and will continue to be useful in the development, evaluation, and verification of computer codes and numerical-solution methods

Implementation and verification of global optimization benchmark problems

Directory of Open Access Journals (Sweden)

Posypkin Mikhail

2017-12-01

Full Text Available The paper considers the implementation and verification of a test suite containing 150 benchmarks for global deterministic box-constrained optimization. A C++ library for describing standard mathematical expressions was developed for this purpose. The library automate the process of generating the value of a function and its’ gradient at a given point and the interval estimates of a function and its’ gradient on a given box using a single description. Based on this functionality, we have developed a collection of tests for an automatic verification of the proposed benchmarks. The verification has shown that literary sources contain mistakes in the benchmarks description. The library and the test suite are available for download and can be used freely.

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

Radiation protection/shield design

International Nuclear Information System (INIS)

Disney, R.K.

1977-01-01

Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. In the following discussion, the system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection will be described, and the program developed to implement this approach will be defined. In addition, the principal shielding design problems of LMFBR nuclear reactor systems will be discussed in realtion to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods will be discussed. (orig.) [de

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

Directory of Open Access Journals (Sweden)

Daniel J. Kelly, III

2017-09-01

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

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Benchmark problem for IAEA coordinated research program (CRP-3) on GCR afterheat removal. 1

International Nuclear Information System (INIS)

Takada, Shoji; Shiina, Yasuaki; Inagaki, Yoshiyuki; Hishida, Makoto; Sudo, Yukio

1995-08-01

In this report, detailed data which are necessary for the benchmark analysis of International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP-3) on 'Heat Transport and Afterheat Removal for Gas-cooled Reactors under Accident Conditions' are described concerning about the configuration and sizes of the cooling panel test apparatus, experimental data and thermal properties. The test section of the test apparatus is composed of pressure vessel (max. 450degC) containing an electric heater (max. 100kW, 600degC) and cooling panels surrounding the pressure vessel. Gas pressure is varied from vacuum to 1.0MPa in the pressure vessel. Two experimental cases are selected as benchmark problems about afterheat removal of HTGR, described as follows, The experimental conditions are vacuum inside the pressure vessel and heater output 13.14kW, and helium gas pressure 0.73MPa inside the pressure vessel and heater output 28.79kW. Benchmark problems are to calculate temperature distributions on the outer surface of pressure vessel and heat transferred to the cooling panel using the experimental data. The analytical result of temperature distribution on the pressure vessel was estimated +38degC, -29degC compared with the experimental data, and analytical result of heat transferred from the surface of pressure vessel to the cooling panel was estimated max. -11.4% compared with the experimental result by using the computational code -THANPACST2- of JAERI. (author)

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

International Nuclear Information System (INIS)

Kliem, S.

1998-01-01

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

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

Application of variance reduction techniques of Monte-Carlo method to deep penetration shielding problems

International Nuclear Information System (INIS)

Rawat, K.K.; Subbaiah, K.V.

1996-01-01

General purpose Monte Carlo code MCNP is being widely employed for solving deep penetration problems by applying variance reduction techniques. These techniques depend on the nature and type of the problem being solved. Application of geometry splitting and implicit capture method are examined to study the deep penetration problems of neutron, gamma and coupled neutron-gamma in thick shielding materials. The typical problems chosen are: i) point isotropic monoenergetic gamma ray source of 1 MeV energy in nearly infinite water medium, ii) 252 Cf spontaneous source at the centre of 140 cm thick water and concrete and iii) 14 MeV fast neutrons incident on the axis of 100 cm thick concrete disk. (author). 7 refs., 5 figs

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

International Nuclear Information System (INIS)

Smith, H.J.; Santamarina, A.

1991-01-01

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

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

International Nuclear Information System (INIS)

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

2014-03-01

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

Comparisons of the MCNP criticality benchmark suite with ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0

International Nuclear Information System (INIS)

Kim, Do Heon; Gil, Choong-Sup; Kim, Jung-Do; Chang, Jonghwa

2003-01-01

A comparative study has been performed with the latest evaluated nuclear data libraries ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0. The study has been conducted through the benchmark calculations for 91 criticality problems with the libraries processed for MCNP4C. The calculation results have been compared with those of the ENDF60 library. The self-shielding effects of the unresolved-resonance (UR) probability tables have also been estimated for each library. The χ 2 differences between the MCNP results and experimental data were calculated for the libraries. (author)

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

Technology development for radiation shielding analysis

International Nuclear Information System (INIS)

Ha, Jung Woo; Lee, Jae Kee; Kim, Jong Kyung

1986-12-01

Radiation shielding analysis in nuclear engineering fields is an important technology which is needed for the calculation of reactor shielding as well as radiation related safety problems in nuclear facilities. Moreover, the design technology required in high level radioactive waste management and disposal facilities is faced on serious problems with rapidly glowing nuclear industry development, and more advanced technology has to be developed for tomorrow. The main purpose of this study is therefore to build up the self supporting ability of technology development for the radiation shielding analysis in order to achieve successive development of nuclear industry. It is concluded that basic shielding calculations are possible to handle and analyze by using our current technology, but more advanced technology is still needed and has to be learned for the degree of accuracy in two-dimensional shielding calculation. (Author)

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.

Comparison of typical inelastic analysis predictions with benchmark problem experimental results

International Nuclear Information System (INIS)

Clinard, J.A.; Corum, J.M.; Sartory, W.K.

1975-01-01

The results of exemplary inelastic analyses for experimental benchmark problems on reactor components are presented. Consistent analytical procedures and constitutive relations were used in each of the analyses, and the material behavior data presented in the Appendix were used in all cases. Two finite-element inelastic computer programs were employed. These programs implement the analysis procedures and constitutive equations for type 304 stainless steel that are currently used in many analyses of elevated-temperature nuclear reactor system components. The analysis procedures and constitutive relations are briefly discussed, and representative analytical results are presented and compared to the test data. The results that are presented demonstrate the feasibility of performing inelastic analyses for the types of problems discussed, and they are indicative of the general level of agreement that the analyst might expect when using conventional inelastic analysis procedures. (U.S.)

Contaminant deposition building shielding factors for US residential structures.

Science.gov (United States)

Dickson, Elijah; Hamby, David; Eckerman, Keith

2017-10-10

This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario from contaminant deposition on the roof and surrounding surfaces. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations from contaminant deposition on the roof and surrounding ground as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement as well for single-wide manufactured housing-unit. © 2017 IOP Publishing Ltd.

Contaminant deposition building shielding factors for US residential structures

International Nuclear Information System (INIS)

Dickson, E D; Hamby, D M; Eckerman, K F

2015-01-01

This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario from contaminant deposition on the roof and surrounding surfaces. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations from contaminant deposition on the roof and surrounding ground as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement as well for single-wide manufactured housing-unit. (paper)

Cloud immersion building shielding factors for US residential structures

International Nuclear Information System (INIS)

Dickson, E D; Hamby, D M

2014-01-01

This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario within a semi-infinite cloud of radioactive material. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement, as well for single-wide manufactured housing-units. (paper)

Using glass as a shielding material

International Nuclear Information System (INIS)

Yousef, S.

2002-04-01

Different theoretical and technological concepts and problems in using glass as a shielding material was discussed, some primarily designs for different types of radiation shielding windows were illustrated. (author)

Using glass as a shielding material

International Nuclear Information System (INIS)

Yousef, S.

2003-01-01

Different theoretical and technological concepts and problems in using glass as a shielding material was discussed, some primarily designs for different types of radiation shielding windows were illustrated. (author)

A perturbation technique for shield weight minimization

International Nuclear Information System (INIS)

Watkins, E.F.; Greenspan, E.

1993-01-01

The radiation shield optimization code SWAN (Ref. 1) was originally developed for minimizing the thickness of a shield that will meet a given dose (or another) constraint or for extremizing a performance parameter of interest (e.g., maximizing energy multiplication or minimizing dose) while maintaining the shield volume constraint. The SWAN optimization process proved to be highly effective (e.g., see Refs. 2, 3, and 4). The purpose of this work is to investigate the applicability of the SWAN methodology to problems in which the weight rather than the volume is the relevant shield characteristic. Such problems are encountered in shield design for space nuclear power systems. The investigation is carried out using SWAN with the coupled neutron-photon cross-section library FLUNG (Ref. 5)

Comparison of three-dimensional ocean general circulation models on a benchmark problem

International Nuclear Information System (INIS)

Chartier, M.

1990-12-01

A french and an american Ocean General Circulation Models for deep-sea disposal of radioactive wastes are compared on a benchmark test problem. Both models are three-dimensional. They solve the hydrostatic primitive equations of the ocean with two different finite difference techniques. Results show that the dynamics simulated by both models are consistent. Several methods for the running of a model from a known state are tested in the French model: the diagnostic method, the prognostic method, the acceleration of convergence and the robust-diagnostic method

C5 Benchmark Problem with Discrete Ordinate Radiation Transport Code DENOVO

Energy Technology Data Exchange (ETDEWEB)

Yesilyurt, Gokhan [ORNL; Clarno, Kevin T [ORNL; Evans, Thomas M [ORNL; Davidson, Gregory G [ORNL; Fox, Patricia B [ORNL

2011-01-01

The C5 benchmark problem proposed by the Organisation for Economic Co-operation and Development/Nuclear Energy Agency was modeled to examine the capabilities of Denovo, a three-dimensional (3-D) parallel discrete ordinates (S{sub N}) radiation transport code, for problems with no spatial homogenization. Denovo uses state-of-the-art numerical methods to obtain accurate solutions to the Boltzmann transport equation. Problems were run in parallel on Jaguar, a high-performance supercomputer located at Oak Ridge National Laboratory. Both the two-dimensional (2-D) and 3-D configurations were analyzed, and the results were compared with the reference MCNP Monte Carlo calculations. For an additional comparison, SCALE/KENO-V.a Monte Carlo solutions were also included. In addition, a sensitivity analysis was performed for the optimal angular quadrature and mesh resolution for both the 2-D and 3-D infinite lattices of UO{sub 2} fuel pin cells. Denovo was verified with the C5 problem. The effective multiplication factors, pin powers, and assembly powers were found to be in good agreement with the reference MCNP and SCALE/KENO-V.a Monte Carlo calculations.

Irrigoscopy - irrigography method, dosimetry and radiation shielding

International Nuclear Information System (INIS)

Zubanov, Z.; Kolarevic, G.

1999-01-01

Use of patient's radiation shielding during radiology diagnostic procedures in our country is insufficiently represent, so patients needlessly receive very high entrance skin doses in body areas which are not in direct x-ray beam. During irrigoscopy, patient's radiation shielding is very complex problem, because of the organs position. In the future that problem must be solved. We hope that some of our suggestions about patient's radiation shielding during irrigoscopy, can be a small step in that way. (author)

Radiation Shielding Information Center: a source of computer codes and data for fusion neutronics studies

International Nuclear Information System (INIS)

McGill, B.L.; Roussin, R.W.; Trubey, D.K.; Maskewitz, B.F.

1980-01-01

The Radiation Shielding Information Center (RSIC), established in 1962 to collect, package, analyze, and disseminate information, computer codes, and data in the area of radiation transport related to fission, is now being utilized to support fusion neutronics technology. The major activities include: (1) answering technical inquiries on radiation transport problems, (2) collecting, packaging, testing, and disseminating computing technology and data libraries, and (3) reviewing literature and operating a computer-based information retrieval system containing material pertinent to radiation transport analysis. The computer codes emphasize methods for solving the Boltzmann equation such as the discrete ordinates and Monte Carlo techniques, both of which are widely used in fusion neutronics. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

Antiproton annihilation physics annihilation physics in the Monte Carlo particle transport code particle transport code SHIELD-HIT12A

DEFF Research Database (Denmark)

Taasti, Vicki Trier; Knudsen, Helge; Holzscheiter, Michael

2015-01-01

The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An...

Measuring space radiation shielding effectiveness

OpenAIRE

Bahadori Amir; Semones Edward; Ewert Michael; Broyan James; Walker Steven

2017-01-01

Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles ...

Transmission test of the polyethylene shield against 40 and 65 MeV quasi monochrome neutron

International Nuclear Information System (INIS)

Nakao, Makoto; Nakamura, Takashi; Sakuya, Yoshimasa; Nauchi, Yasushi; Nakao, Noriaki; Tanaka, Susumu; Sakamoto, Yukio; Nakajima, Hiroshi; Nakane, Yoshihiro.

1996-01-01

Using 40 and 65 MeV quasi monochrome neutron of the AVF cyclotron installed at Takasaki Laboratory, Japan Atomic Energy Research Institute, the neutron energy spectra were measured after transmitting the polyethylene shield. Results of the shielding experiments using concrete and iron recognized as main shielding material were proposed previously. As data obtained in the experiments were useful for a bench-mark experiment to investigate for shielding calculation and sectional data set, a shielding calculation simulated with new experiment to compare with and investigate for the previous experimental data. As a result, it was found that calculation result of neutron flux transmitting through the polyethylene shield showed difference with increase of the shield thickness. And, reducing distance of the peak neutron was also found to be over-estimated in its calculation value, such as three and five times on 43 MeV at 120 and 180 cm thick, respectively. (G.K.)

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

DOT 3.5-E (DOT 3.5-E/JEF-1) analysis of the PCA-Replica (H2O/FE) shielding benchmark for the LWR-PV damage prediction

International Nuclear Information System (INIS)

Pescarini, M.

1991-01-01

The results of a DOT 3.5-E/JEF-1 validation on the (H2O/Fr) PCA-REPLICA (UKAEA-Winfith) low-flux shielding benchmark are presented. The PCA-REPLICA experiments reproduces the excore radial geometry of a PWR and is closely related to LWR safety since it is dedicated to test the accuracy of the calculated neutron exposure parameters (fast fluence and iron displacement rates) in a pressure vessel simulator. The NJOY/THEMIS data processing system is employed to obtain the neutron damage-energy cross sections for the JEF-1 iron file. The SN 1-D ANISN code is used to collapse cross sections from the VITAMIN-J (175 n) shielding library, based on the JEF-1 data, to a 28 group working library for 2-D calculations. A 3-D-equivalent synthesis (X,Y,Z) of 2-D and 1-D DOT 3.5-E SN calculations in a plane geometry, gives the integral and spectral results for comparison with the respective experimental data. The underprediction of the in-vessel dosimeter experimental activities depends probably on an overestimation of the iron inelastic scattering cross section of the JEF-1 file

PID controller tuning using metaheuristic optimization algorithms for benchmark problems

Science.gov (United States)

Gholap, Vishal; Naik Dessai, Chaitali; Bagyaveereswaran, V.

2017-11-01

This paper contributes to find the optimal PID controller parameters using particle swarm optimization (PSO), Genetic Algorithm (GA) and Simulated Annealing (SA) algorithm. The algorithms were developed through simulation of chemical process and electrical system and the PID controller is tuned. Here, two different fitness functions such as Integral Time Absolute Error and Time domain Specifications were chosen and applied on PSO, GA and SA while tuning the controller. The proposed Algorithms are implemented on two benchmark problems of coupled tank system and DC motor. Finally, comparative study has been done with different algorithms based on best cost, number of iterations and different objective functions. The closed loop process response for each set of tuned parameters is plotted for each system with each fitness function.

Shield design development of nuclear propulsion merchant ship

International Nuclear Information System (INIS)

Tanaka, Yoshihisa

1975-01-01

Shielding design both in Japan and abroad for nuclear propulsion merchant ships is explained, with emphasis on the various technological problems having occurred in the shield design for one-body type and separate type LWRs as conceptual design. The following matters are described: the peculiarities of the design as compared with the case of land-based nuclear reactors, problems in the design standards of shielding, the present status and development of the design methods, and the instances of the design; thereby, the trends of shielding design are disclosed. The following matters are pointed out: Importance of the optimum design, of shielding, significance of radiation streaming through large voids, activation of the secondary water in built-in type steam generators, and the need of the guides for shield design. (Mori, K.)

Demonstration study on shielding safety analysis code (VI)

Energy Technology Data Exchange (ETDEWEB)

Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

1999-03-01

Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this steady is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) Construction and improvement of a pulsed radiation measurement system due to the gated counting method. (2) Using the system, carried out the radiation monitoring near and in the facility of 45 MeV Linear accelerator installed at Hokkaido University. (3) Simulation analysis of the photo-neutron production and the transport by using the EGS4 and MCNP code. (author)

The OECD/NEA Data Bank, its computer program services and benchmarking activities

International Nuclear Information System (INIS)

Sartori, E.; Galan, J.M.

1998-01-01

The OECD/NEA Data Bank collects, tests and distributes computer programs and numerical data in the field of nuclear energy applications. This activity is coordinated with several similar centres in the United States (ESTSC, NNDC, RSIC) and outside the OECD area through an arrangement with the IAEA. This information is shared worldwide for the benefit of scientists and engineers working on the safe and economic use of nuclear energy. The OECD/NEA Nuclear Science Committee the supervising body of the Data Bank has conducted a series of international computer code benchmark exercises with the aim of verifying the correctness of codes, of building confidence in models used for predicting macroscopic behaviour of nuclear systems and to drive towards refinement of models where necessary. Exercises involving nuclear cross section predictions, in-core reactor physics issues, such as pin cells for different type of reactors, plutonium recycling, reconstruction of pin power within assemblies, core transients, reactor shielding and dosimetry, away from reactor issues such as criticality safety for transport and storage of spent fuel, shielding of radioactive material packages and other problems connected with the back end of the fuel cycle, are listed and the relevant references provided. (author)

Solution of 2D and 3D hexagonal geometry benchmark problems by using the finite element diffusion code DIFGEN

International Nuclear Information System (INIS)

Gado, J.

1986-02-01

The four group, 2D and 3D hexagonal geometry HTGR benchmark problems and a 2D hexagonal geometry PWR (WWER) benchmark problem have been solved by using the finite element diffusion code DIFGEN. The hexagons (or hexagonal prisms) were subdivided into first order or second order triangles or quadrilaterals (or triangular or quadrilateral prisms). In the 2D HTGR case of the number of the inserted absorber rods was also varied (7, 6, 0 or 37 rods). The calculational results are in a good agreement with the results of other calculations. The larger systematic series of DIFGEN calculations have given a quantitative picture on the convergence properties of various finite element modellings of hexagonal grids in DIFGEN. (orig.)

Validation of the DRAGON/DONJON code package for MNR using the IAEA 10 MW benchmark problem

International Nuclear Information System (INIS)

Day, S.E.; Garland, W.J.

2000-01-01

The first step in developing a framework for reactor physics analysis is to establish the appropriate and proven reactor physics codes. The chosen code package is tested, by executing a benchmark problem and comparing the results to the accepted standards. The IAEA 10 MW Benchmark problem is suitable for static reactor physics calculations on plate-fueled research reactor systems and has been used previously to validate codes for the McMaster Nuclear (MNR). The flexible and advanced geometry capabilities of the DRAGON transport theory code make it a desirable tool, and the accompanying DONJON diffusion theory code also has useful features applicable to safety analysis work at MNR. This paper describes the methodology used to benchmark the DRAGON/DONJON code package against this problem and the results herein extend the domain of validation of this code package. The results are directly applicable to MNR and are relevant to a reduced-enrichment fuel program. The DRAGON transport code models, used in this study, are based on the 1-D infinite slab approximation whereas the DONJON diffusion code models are defined in 3-D Cartesian geometry. The cores under consideration are composed of HEU (93% enrichment), MEU (45% enrichment) and LEU (20% enrichment) fuel and are examined in a fresh state, as well as at beginning-of-life (BOL) and end-of-life (EOL) exposures. The required flux plots and flux-ratio plots are included, as are transport theory code k∞and diffusion theory code k eff results. In addition to this, selected isotope atom densities are charted as a function of fuel burnup. Results from this analysis are compared to and are in good agreement with previously published results. (author)

MHD and heat transfer benchmark problems for liquid metal flow in rectangular ducts

International Nuclear Information System (INIS)

Sidorenkov, S.I.; Hua, T.Q.; Araseki, H.

1994-01-01

Liquid metal cooling systems of a self-cooled blanket in a tokamak reactor will likely include channels of rectangular cross section where liquid metal is circulated in the presence of strong magnetic fields. MHD pressure drop, velocity distribution and heat transfer characteristics are important issues in the engineering design considerations. Computer codes for the reliable solution of three-dimensional MHD flow problems are needed for fusion relevant conditions. Argonne National Laboratory and The Efremov Institute have jointly defined several benchmark problems for code validation. The problems, described in this paper, are based on two series of rectangular duct experiments conducted at ANL; one of the series is a joint ANL/Efremov experiment. The geometries consist of variation of aspect ratio and wall thickness (thus wall conductance ratio). The transverse magnetic fields are uniform and nonuniform in the axial direction

Error Analysis of Variations on Larsen's Benchmark Problem

International Nuclear Information System (INIS)

Azmy, YY

2001-01-01

Error norms for three variants of Larsen's benchmark problem are evaluated using three numerical methods for solving the discrete ordinates approximation of the neutron transport equation in multidimensional Cartesian geometry. The three variants of Larsen's test problem are concerned with the incoming flux boundary conditions: unit incoming flux on the left and bottom edges (Larsen's configuration); unit, incoming flux only on the left edge; unit incoming flux only on the bottom edge. The three methods considered are the Diamond Difference (DD) method, and the constant-approximation versions of the Arbitrarily High Order Transport method of the Nodal type (AHOT-N), and of the Characteristic (AHOT-C) type. The cell-wise error is computed as the difference between the cell-averaged flux computed by each method and the exact value, then the L 1 , L 2 , and L ∞ error norms are calculated. The results of this study demonstrate that while integral error norms, i.e. L 1 , L 2 , converge to zero with mesh refinement, the pointwise L ∞ norm does not due to solution discontinuity across the singular characteristic. Little difference is observed between the error norm behavior of the three methods considered in spite of the fact that AHOT-C is locally exact, suggesting that numerical diffusion across the singular characteristic as the major source of error on the global scale. However, AHOT-C possesses a given accuracy in a larger fraction of computational cells than DD

3D Space Radiation Transport in a Shielded ICRU Tissue Sphere

Science.gov (United States)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2014-01-01

A computationally efficient 3DHZETRN code capable of simulating High Charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation was recently developed for a simple homogeneous shield object. Monte Carlo benchmarks were used to verify the methodology in slab and spherical geometry, and the 3D corrections were shown to provide significant improvement over the straight-ahead approximation in some cases. In the present report, the new algorithms with well-defined convergence criteria are extended to inhomogeneous media within a shielded tissue slab and a shielded tissue sphere and tested against Monte Carlo simulation to verify the solution methods. The 3D corrections are again found to more accurately describe the neutron and light ion fluence spectra as compared to the straight-ahead approximation. These computationally efficient methods provide a basis for software capable of space shield analysis and optimization.

Performance of the improved version of Monte Carlo code A 3MCNP for large-scale shielding problems

International Nuclear Information System (INIS)

Omura, M.; Miyake, Y.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G. E.

2005-01-01

A 3MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, which automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3MCNP uses the three-dimensional (3-D) Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3MCNP (referred to as A 3MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3MCNPV for a concrete cask neutron and gamma-ray shielding problem, and a PWR dosimetry problem. (authors)

OECD/NRC Benchmark Based on NUPEC PWR Sub-channel and Bundle Test (PSBT). Volume I: Experimental Database and Final Problem Specifications

International Nuclear Information System (INIS)

Rubin, A.; Schoedel, A.; Avramova, M.; Utsuno, H.; Bajorek, S.; Velazquez-Lozada, A.

2012-01-01

The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermal-hydraulics modelling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan, which includes sub-channel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurised Water Reactor (PWR) fuel assembly. Part of this database has been made available for this international benchmark activity entitled 'NUPEC PWR Sub-channel and Bundle Tests (PSBT) benchmark'. This international project has been officially approved by the Japanese Ministry of Economy, Trade, and Industry (METI), the US Nuclear Regulatory Commission (NRC) and endorsed by the OECD/NEA. The benchmark team has been organised based on the collaboration between Japan and the USA. A large number of international experts have agreed to participate in this programme. The fine-mesh high-quality sub-channel void fraction and departure from nucleate boiling data encourages advancement in understanding and modelling complex flow behaviour in real bundles. Considering that the present theoretical approach is relatively immature, the benchmark specification is designed so that it will systematically assess and compare the participants' analytical models on the prediction of detailed void distributions and DNB. The development of truly mechanistic models for DNB prediction is currently underway. The benchmark problem includes both macroscopic and microscopic measurement data. In this context, the sub-channel grade void fraction data are regarded as the macroscopic data and the digitised computer graphic images are the

Radiation shielding lead shield

International Nuclear Information System (INIS)

Dei, Shoichi.

1991-01-01

The present invention concerns lead shields for radiation shielding. Shield boxes are disposed so as to surround a pipeline through which radioactive liquids, mists or like other objects are passed. Flanges are formed to each of the end edges of the shield boxes and the shield boxes are connected to each other by the flanges. Upon installation, empty shield boxes not charged with lead particles and iron plate shields are secured at first at the periphery of the pipeline. Then, lead particles are charged into the shield boxes. This attains a state as if lead plate corresponding to the depth of the box is disposed. Accordingly, operations for installation, dismantling and restoration can be conducted in an empty state with reduced weight to facilitate the operations. (I.S.)

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

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.

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

ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport

International Nuclear Information System (INIS)

1993-01-01

1 - Nature of physical problem solved: ASOP is a shield optimization calculational system based on the one-dimensional discrete ordinates transport program ANISN. It has been used to design optimum shields for space applications of SNAP zirconium-hydride-uranium- fueled reactors and uranium-oxide fueled thermionic reactors and to design beam stops for the ORELA facility. 2 - Method of solution: ASOP generates coefficients of linear equations describing the logarithm of the dose and dose-weight derivatives as functions of position from data obtained in an automated sequence of ANISN calculations. With the dose constrained to a design value and all dose-weight derivatives required to be equal, the linear equations may be solved for a new set of shield dimensions. Since changes in the shield dimensions may cause the linear functions to change, the entire procedure is repeated until convergence is obtained. The detailed calculations of the radiation transport through shield configurations for every step in the procedure distinguish ASOP from other shield optimization computer code systems which rely on multiple component sources and attenuation coefficients to describe the transport. 3 - Restrictions on the complexity of the problem: Problem size is limited only by machine size

OECD/NEZ Main Steam Line Break Benchmark Problem Exercise I Simulation Using the SPACE Code with the Point Kinetics Model

International Nuclear Information System (INIS)

Kim, Yohan; Kim, Seyun; Ha, Sangjun

2014-01-01

The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Nuclear Hydro and Nuclear Power Co. (KHNP) through collaborative works with other Korean nuclear industries. The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient features to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the development, the 2.14 version of the code was released through the successive various V and V works. The topical reports on the code and related safety analysis methodologies have been prepared for license works. In this study, the OECD/NEA Main Steam Line Break (MSLB) Benchmark Problem Exercise I was simulated as a V and V work. The results were compared with those of the participants in the benchmark project. The OECD/NEA MSLB Benchmark Problem Exercise I was simulated using the SPACE code. The results were compared with those of the participants in the benchmark project. Through the simulation, it was concluded that the SPACE code can effectively simulate PWR MSLB accidents

A Study of Fixed-Order Mixed Norm Designs for a Benchmark Problem in Structural Control

Science.gov (United States)

Whorton, Mark S.; Calise, Anthony J.; Hsu, C. C.

1998-01-01

This study investigates the use of H2, p-synthesis, and mixed H2/mu methods to construct full-order controllers and optimized controllers of fixed dimensions. The benchmark problem definition is first extended to include uncertainty within the controller bandwidth in the form of parametric uncertainty representative of uncertainty in the natural frequencies of the design model. The sensitivity of H2 design to unmodelled dynamics and parametric uncertainty is evaluated for a range of controller levels of authority. Next, mu-synthesis methods are applied to design full-order compensators that are robust to both unmodelled dynamics and to parametric uncertainty. Finally, a set of mixed H2/mu compensators are designed which are optimized for a fixed compensator dimension. These mixed norm designs recover the H, design performance levels while providing the same levels of robust stability as the u designs. It is shown that designing with the mixed norm approach permits higher levels of controller authority for which the H, designs are destabilizing. The benchmark problem is that of an active tendon system. The controller designs are all based on the use of acceleration feedback.

Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors

International Nuclear Information System (INIS)

Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L.; Xolocostli M, J. V.; Gomez T, A. M.

2016-09-01

The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S_N, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO_2 cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors

International Nuclear Information System (INIS)

Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M.; Reyes F, M. del C.; Del Valle G, E.

2014-10-01

In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

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

Directory of Open Access Journals (Sweden)

Tagor Malem Sembiring

2015-10-01

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

The problem of resonance self-shielding effect in neutron multigroup calculations

International Nuclear Information System (INIS)

Wang Qingming; Huang Jinghua

1991-01-01

It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations

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 radiation shielding

International Nuclear Information System (INIS)

Kaplan, M.F.

1989-01-01

The increased use of nuclear energy has given rise to a growth in the amount of artificially produced radiation and radioactive materials. The design and construction of shielding to protect people, equipment and structures from the effects of radiation has never been more important. Experience has shown that concrete is an effective, versatile and economical material for the construction of radiation shielding. This book provides information on the principles governing the interaction of radiation with matter and on relevant nuclear physics to give the engineer an understanding of the design and construction of concrete shielding. It covers the physical, mechanical and nuclear properties of concrete; the effects of elevated temperatures and possible damage to concrete due to radiation; basic procedures for the design of concrete radiation shields and finally the special problems associated with their construction and cost. Although written primarily for engineers concerned with the design and construction of concrete shielding, the book also reviews the widely scattered data and information available on this subject and should therefore be of interest to students and those wishing to research further in this field. (author)

A Monte-Carlo Benchmark of TRIPOLI-4® and MCNP on ITER neutronics

Science.gov (United States)

Blanchet, David; Pénéliau, Yannick; Eschbach, Romain; Fontaine, Bruno; Cantone, Bruno; Ferlet, Marc; Gauthier, Eric; Guillon, Christophe; Letellier, Laurent; Proust, Maxime; Mota, Fernando; Palermo, Iole; Rios, Luis; Guern, Frédéric Le; Kocan, Martin; Reichle, Roger

2017-09-01

Radiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the `C-lite', is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results.

Evaluation of the shielding integrity of end-shields in PHWR type NPPs

International Nuclear Information System (INIS)

Sah, B.M.L.; Ramamirtham, B.; Kutty, B.S.

1996-01-01

In the new plants (Narora Atomic Power Plants (NAPP) onwards) relatively higher radiation fields exist on the north and south fuelling machine (FM) vault walls of the E1 100m accessible area passages. These fields were first noticed at NAPS-1 and subsequently at NAPS-2 and KAPS-1. Such surveys done at RAPS have indicated that the fields on these walls would come out to be quite low (only 1-2 mR/h) from sources other than that arising from 41 Ar contamination. RAPS/MAPS experience pointed to adequacy of shielding of the FM vault walls and sufficient overall shielding thickness of the end-shields. Further, radiometry tests of end-shields carried out at Kaiga and RAPP 3 and 4 indicated fairly satisfactory and uniform filling of balls. Hence, incomplete filling of water column of the end-shields due to any venting problem was suspected to be one possible reason for the observed high fields in NAPS and Kakrapar Atomic Power Station (KAPS). Since the presence of high radiation fields, both neutron and gamma, is of long-term concern, a special study/measurement of radiation levels on reactor face during high power operation was undertaken. In order to compare the shielding integrity of the older (RAPS/MAPS solid plate type shielding) and newer (NAPS/KAPS steel ball-filled type) end shields, these experiments were done at MAPS-2 and NAPS-2. (author). 2 refs., 2 tabs

Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

Directory of Open Access Journals (Sweden)

Ahmed Amin El Said Abd El Hameed

2015-08-01

Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code.  We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon

Benchmarking: A Process for Improvement.

Science.gov (United States)

Peischl, Thomas M.

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

Concrete shielding exterior to iron

International Nuclear Information System (INIS)

Yurista, P.; Cossairt, D.

1983-08-01

A rule of thumb at Fermilab has been to use 3 feet of concrete exterior to iron shielding. A recent design of a shield with a severe dimensional constraint has prompted a re-evaluation of this rule of thumb and has led to the following calculations of the concrete thickness required to nullify this problem. 4 references, 4 figures

Computing Moment-Based Probability Tables for Self-Shielding Calculations in Lattice Codes

International Nuclear Information System (INIS)

Hebert, Alain; Coste, Mireille

2002-01-01

As part of the self-shielding model used in the APOLLO2 lattice code, probability tables are required to compute self-shielded cross sections for coarse energy groups (typically with 99 or 172 groups). This paper describes the replacement of the multiband tables (typically with 51 subgroups) with moment-based tables in release 2.5 of APOLLO2. An improved Ribon method is proposed to compute moment-based probability tables, allowing important savings in CPU resources while maintaining the accuracy of the self-shielding algorithm. Finally, a validation is presented where the absorption rates obtained with each of these techniques are compared with exact values obtained using a fine-group elastic slowing-down calculation in the resolved energy domain. Other results, relative to the Rowland's benchmark and to three assembly production cases, are also presented

Shielding analysis method applied to nuclear ship 'MUTSU' and its evaluation based on experimental analyses

International Nuclear Information System (INIS)

Yamaji, Akio; Miyakoshi, Jun-ichi; Iwao, Yoshiaki; Tsubosaka, Akira; Saito, Tetsuo; Fujii, Takayoshi; Okumura, Yoshihiro; Suzuoki, Zenro; Kawakita, Takashi.

1984-01-01

Procedures of shielding analysis are described which were used for the shielding modification design of the Nuclear Ship ''MUTSU''. The calculations of the radiation distribution on board were made using Sn codes ANISN and TWOTRAN, a point kernel code QAD and a Monte Carlo code MORSE. The accuracies of these calculations were investigated through the analysis of various shielding experiments: the shield tank experiment of the Nuclear Ship ''Otto Hahn'', the shielding mock-up experiment for ''MUTSU'' performed in JRR-4, the shielding benchmark experiment using the 16 N radiation facility of AERE Harwell and the shielding effect experiment of the ship structure performed in the training ship ''Shintoku-Maru''. The values calculated by the ANISN agree with the data measured at ''Otto Hahn'' within a factor of 2 for fast neutrons and within a factor of 3 for epithermal and thermal neutrons. The γ-ray dose rates calculated by the QAD agree with the measured values within 30% for the analysis of the experiment in JRR-4. The design values for ''MUTSU'' were determined in consequence of these experimental analyses. (author)

Calculation of neutron shielding for a real loaded C-30 cask by code DORT

International Nuclear Information System (INIS)

Lacina, J.

1999-01-01

Measured neutron dose rates of real loaded C-30 casks for WWER spent fuel assemblies are compared with calculated values in the frame of benchmark calculation task. The part of this benchmark task concerning neutron shielding was calculated. Neutron sources values were taken from data presented by V. Chrapciak during the eighth symposium Atomic Energy Research, Bystrice pod Perstejnem in 1998 and the data about cask from the article of the same author from the Atomic Energy Research working group E meeting at Stolpen in 1998. (Author)

TRIPOLI 01, a three-dimensional polykinetic Monte Carlo program. Pt.2. Constant data input and results obtained for a complex bench-mark example

International Nuclear Information System (INIS)

Baur, A.; Bourdet, L.; Gonnord, J.; Nimal, J.C.; Vergnaud, T.

1977-01-01

Some properties of the sub-routines included in TRIPOLI for the constant data input are briefly reviewed. The bench-mark example then proposed is a problem of the contour lines of a neutron-opaque shield. The example is derived from activation calculations for the secondary sodium in a fast reactor provided with an integrated shield of Phenix type. The dimensions have been reduced so as the geometry to be simplified, and the constant data load reduced. Besides the rate of 24 Na formation by capture in the exchanger region, the attenuation of various detector responses all along the neutron trajectory was taken into account. The detectors chiefly used were: 103 Rh activation from (n,n') reactions for fast neutron detection, the activation Mn detectors, either bare or under cadmium, for detecting thermal and epithermal neutrons respectively [fr

Neutron streaming studies along JET shielding penetrations

Science.gov (United States)

Stamatelatos, Ion E.; Vasilopoulou, Theodora; Batistoni, Paola; Obryk, Barbara; Popovichev, Sergey; Naish, Jonathan

2017-09-01

Neutronic benchmark experiments are carried out at JET aiming to assess the neutronic codes and data used in ITER analysis. Among other activities, experiments are performed in order to validate neutron streaming simulations along long penetrations in the JET shielding configuration. In this work, neutron streaming calculations along the JET personnel entrance maze are presented. Simulations were performed using the MCNP code for Deuterium-Deuterium and Deuterium- Tritium plasma sources. The results of the simulations were compared against experimental data obtained using thermoluminescence detectors and activation foils.

Issues in Benchmark Metric Selection

Science.gov (United States)

Crolotte, Alain

It is true that a metric can influence a benchmark but will esoteric metrics create more problems than they will solve? We answer this question affirmatively by examining the case of the TPC-D metric which used the much debated geometric mean for the single-stream test. We will show how a simple choice influenced the benchmark and its conduct and, to some extent, DBMS development. After examining other alternatives our conclusion is that the “real” measure for a decision-support benchmark is the arithmetic mean.

Status report of shielding investigation in Japan

International Nuclear Information System (INIS)

Shindo, M.

1964-01-01

The Japan Atomic Energy Research Institute (JAERI) was established in 1954, and immediately proceeded with the construction of a research reactor. The first symposium in Japan on nuclear energy was held in 1957. Most of the papers presented in the field of reactor shielding were limited to shielding materials and their fabrication. In the first stage of our investigations, our efforts were devoted to practical design studies of reactor shielding. As a result of these studies, it was found that the formulae at hand for calculations were inadequate, but at that time no electronic computer was available in Japan nor were theoretical calculations very actively undertaken. Problems on nuclear ship shielding had been investigated at the Ship Research Institute, since 1956 and many fruitful results had been obtained. About that time the Japan Atomic Industry Forum started activities and took the initiative in organizing shielding research. Research workers in the shipbuilding industry in particular have been seriously studying shielding problems. Few years after the first symposium, problems concerning more fundamental studies were treated by many research workers. Shielding experiments using radioisotopes were carried out and many fruitful results were obtained. They are described in the this paper. Medium size electronic computers became available in Japan, permitting a theoretical study group to make an active contribution. They produced some codes, and their results are also described in the following sections. This constituted the second stage of our investigations. A swimming-pool reactor, JRR-4 (Japan Research Reactor-4), has been under construction at JAERI since 1962 and will become critical in autumn 1964. After characteristic tests it will be a very powerful tool for the shielding investigations. This id the beginning of the third stage of investigations

INTEGRAL BENCHMARK DATA FOR NUCLEAR DATA TESTING THROUGH THE ICSBEP AND THE NEWLY ORGANIZED IRPHEP

Energy Technology Data Exchange (ETDEWEB)

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

2007-04-01

The status of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) was last reported in a nuclear data conference at the International Conference on Nuclear Data for Science and Technology, ND-2004, in Santa Fe, New Mexico. Since that time the number and type of integral benchmarks have increased significantly. Included in the ICSBEP Handbook are criticality-alarm / shielding and fundamental physic benchmarks in addition to the traditional critical / subcritical benchmark data. Since ND 2004, a reactor physics counterpart to the ICSBEP, the International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. The IRPhEP is patterned after the ICSBEP, but focuses on other integral measurements, such as buckling, spectral characteristics, reactivity effects, reactivity coefficients, kinetics measurements, reaction-rate and power distributions, nuclide compositions, and other miscellaneous-type measurements in addition to the critical configuration. The status of these two projects is discussed and selected benchmarks highlighted in this paper.

Magnetic shielding for MRI superconducting magnets

International Nuclear Information System (INIS)

Ishiyama, A.; Hirooka, H.

1991-01-01

This paper describes an optimal design of a highly homogeneous superconducting coil system with magnetic shielding for Magnetic Resonance Imaging (MRI). The presented optimal design method; which is originally proposed in our earlier papers, is a combination of the hybrid finite element and boundary element method for analysis of an axially symmetric nonlinear open boundary magnetic field problem, and the mathematical programming method for solving the corresponding optimization problem. In this paper, the multi-objective goal programming method and the nonlinear least squares method have been adopted. The optimal design results of 1.5- and 4.7-Tesla-magnet systems with different types of magnetic shielding for whole-body imaging are compared and the advantages of a combination of active and yoke shields are shown

Highly heat removing radiation shielding material

International Nuclear Information System (INIS)

Asano, Norio; Hozumi, Masahiro.

1990-01-01

Organic materials, inorganic materials or metals having excellent radiation shielding performance are impregnated into expanded metal materials, such as Al, Cu or Mg, having high heat conductivity. Further, the porosity of the expanded metals and combination of the expanded metals and the materials to be impregnated are changed depending on the purpose. Further, a plurality of shielding materials are impregnated into the expanded metal of the same kind, to constitute shielding materials. In such shielding materials, impregnated materials provide shielding performance against radiation rays such as neutrons and gamma rays, the expanded metals provide heat removing performance respectively and they act as shielding materials having heat removing performance as a whole. Accordingly, problems of non-informity and discontinuity in the prior art can be dissolved be provide materials having flexibility in view of fabrication work. (T.M.)

EURLIB-LWR-45/16 and - 15/5. Two board group libraries for LWR-shielding problems

Energy Technology Data Exchange (ETDEWEB)

Herrnberger, V

1982-04-01

Specifications of the broad group cross section libraries EURLIB-LWR-45/16 and -15/5 are given. They are based on EURLIB-III data and produced for LWR shielding problems. The elements considered are H, C{sub 12}, O, Na, Al, Si, Ca, Cr, Mn, Fe, Ni, Zr, U{sub 235}, U{sub 238}. The cross section libraries are available upon request from EIR, RSIC, NEA-CPL and IAEA-NDS. (author) Refs, figs, tabs

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...

Study on the mechanism and efficiency of simulated annealing using an LP optimization benchmark problem - 113

International Nuclear Information System (INIS)

Qianqian, Li; Xiaofeng, Jiang; Shaohong, Zhang

2010-01-01

Simulated Annealing Algorithm (SAA) for solving combinatorial optimization problems is a popular method for loading pattern optimization. The main purpose of this paper is to understand the underlying search mechanism of SAA and to study its efficiency. In this study, a general SAA that employs random pair exchange of fuel assemblies to search for the optimum fuel Loading Pattern (LP) is applied to an exhaustively searched LP optimization benchmark problem. All the possible LPs of the benchmark problem have been enumerated and evaluated via the use of the very fast and accurate Hybrid Harmonics and Linear Perturbation (HHLP) method, such that the mechanism of SA for LP optimization can be explicitly analyzed and its search efficiency evaluated. The generic core geometry itself dictates that only a small number LPs can be generated by performing random single pair exchanges and that the LPs are necessarily mostly similar to the initial LP. This phase space effect turns out to be the basic mechanism in SAA that can explain its efficiency and good local search ability. A measure of search efficiency is introduced which shows that the stochastic nature of SAA greatly influences the variability of its search efficiency. It is also found that using fuel assembly k-infinity distribution as a technique to filter the LPs can significantly enhance the SAA search efficiency. (authors)

BIGHORN Computational Fluid Dynamics Theory, Methodology, and Code Verification & Validation Benchmark Problems

Energy Technology Data Exchange (ETDEWEB)

Xia, Yidong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andrs, David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

This document presents the theoretical background for a hybrid finite-element / finite-volume fluid flow solver, namely BIGHORN, based on the Multiphysics Object Oriented Simulation Environment (MOOSE) computational framework developed at the Idaho National Laboratory (INL). An overview of the numerical methods used in BIGHORN are discussed and followed by a presentation of the formulation details. The document begins with the governing equations for the compressible fluid flow, with an outline of the requisite constitutive relations. A second-order finite volume method used for solving the compressible fluid flow problems is presented next. A Pressure-Corrected Implicit Continuous-fluid Eulerian (PCICE) formulation for time integration is also presented. The multi-fluid formulation is being developed. Although multi-fluid is not fully-developed, BIGHORN has been designed to handle multi-fluid problems. Due to the flexibility in the underlying MOOSE framework, BIGHORN is quite extensible, and can accommodate both multi-species and multi-phase formulations. This document also presents a suite of verification & validation benchmark test problems for BIGHORN. The intent for this suite of problems is to provide baseline comparison data that demonstrates the performance of the BIGHORN solution methods on problems that vary in complexity from laminar to turbulent flows. Wherever possible, some form of solution verification has been attempted to identify sensitivities in the solution methods, and suggest best practices when using BIGHORN.

Radiation shielding for fusion reactors

International Nuclear Information System (INIS)

Santoro, R.T.

2000-01-01

Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. (author)

Radiation shielding techniques and applications. 3. Analysis of Photon Streaming Through and Around Shield Doors

International Nuclear Information System (INIS)

Barnett, Marvin; Hack, Joe; Nathan, Steve; White, Travis

2001-01-01

Westinghouse Safety Management Solutions (Westinghouse SMS) has been tasked with providing radiological engineering design support for the new Commercial Light Water Reactor Tritium Extraction Facility (CLWR-TEF) being constructed at the Savannah River Site (SRS). The Remote Handling Building (RHB) of the CLWR-TEF will act as the receiving facility for irradiated targets used in the production of tritium for the U.S. Department of Energy (DOE). Because of the high dose rates, approaching 50 000 rads/h (500 Gy/h) from the irradiated target bundles, significant attention has been made to shielding structures within the facility. One aspect of the design that has undergone intense review is the shield doors. The RHB has six shield doors that needed to be studied with respect to photon streaming. Several aspects had to be examined to ensure that the design meets the radiation dose levels. Both the thickness and streaming issues around the door edges were designed and examined. Photon streaming through and around a shield door is a complicated problem, creating a reliance on computer modeling to perform the analyses. The computer code typically used by the Westinghouse SMS in the evaluation of photon transport through complex geometries is the MCNP Monte Carlo computer code. The complexity of the geometry within the problem can cause problems even with the Monte Carlo codes. Striking a balance between how the code handles transport through the shield door with transport through the streaming paths, particularly with the use of typical variance reduction methods, is difficult when trying to ensure that all important regions of the model are sampled appropriately. The thickness determination used a simple variance reduction technique. In construction, the shield door will not be flush against the wall, so a solid rectangular slab leaves streaming paths around the edges. Administrative controls could be used to control dose to workers; however, 10 CFR 835.1001 states

Measuring space radiation shielding effectiveness

Directory of Open Access Journals (Sweden)

Bahadori Amir

2017-01-01

Full Text Available Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles is described. Using accelerated alpha particles at the National Aeronautics and Space Administration Space Radiation Laboratory at Brookhaven National Laboratory, the method is applied to sample tiles from the Heat Melt Compactor, which were created by melting material from a simulated astronaut waste stream, consisting of materials such as trash and unconsumed food. The shielding effectiveness calculated from measurements of the Heat Melt Compactor sample tiles is about 10% less than the shielding effectiveness of polyethylene. Shielding material produced from the astronaut waste stream in the form of Heat Melt Compactor tiles is therefore found to be an attractive solution for protection against space radiation.

Measuring space radiation shielding effectiveness

Science.gov (United States)

Bahadori, Amir; Semones, Edward; Ewert, Michael; Broyan, James; Walker, Steven

2017-09-01

Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles is described. Using accelerated alpha particles at the National Aeronautics and Space Administration Space Radiation Laboratory at Brookhaven National Laboratory, the method is applied to sample tiles from the Heat Melt Compactor, which were created by melting material from a simulated astronaut waste stream, consisting of materials such as trash and unconsumed food. The shielding effectiveness calculated from measurements of the Heat Melt Compactor sample tiles is about 10% less than the shielding effectiveness of polyethylene. Shielding material produced from the astronaut waste stream in the form of Heat Melt Compactor tiles is therefore found to be an attractive solution for protection against space radiation.

The radiation shielding potential of CI and CM chondrites

Science.gov (United States)

Pohl, Leos; Britt, Daniel T.

2017-03-01

Galactic Cosmic Rays (GCRs) and Solar Energetic Particles (SEPs) pose a serious limit on the duration of deep space human missions. A shield composed of a bulk mass of material in which the incident particles deposit their energy is the simplest way to attenuate the radiation. The cost of bringing the sufficient mass from the Earth's surface is prohibitive. The shielding properties of asteroidal material, which is readily available in space, are investigated. Solution of Bethe's equation is implemented for incident protons and the application in composite materials and the significance of various correction terms are discussed; the density correction is implemented. The solution is benchmarked and shows good agreement with the results in literature which implement more correction terms within the energy ranges considered. The shielding properties of CI and CM asteroidal taxonomy groups and major asteroidal minerals are presented in terms of stopping force. The results show that CI and CM chondrites have better stopping properties than Aluminium. Beneficiation is discussed and is shown to have a significant effect on the stopping power.

Improvement of Monte Carlo code A3MCNP for large-scale shielding problems

International Nuclear Information System (INIS)

Miyake, Y.; Ohmura, M.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G.E.

2004-01-01

A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3 MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for a concrete cask streaming problem and a PWR dosimetry problem. (author)

Comparison of typical inelastic analysis predictions with benchmark problem experimental results

International Nuclear Information System (INIS)

Clinard, J.A.; Corum, J.M.; Sartory, W.K.

1975-01-01

The results of exemplary inelastic analyses are presented for a series of experimental benchmark problems. Consistent analytical procedures and constitutive relations were used in each of the analyses, and published material behavior data were used in all cases. Two finite-element inelastic computer programs were employed. These programs implement the analysis procedures and constitutive equations for Type 304 stainless steel that are currently used in many analyses of elevated-temperature nuclear reactor system components. The analysis procedures and constitutive relations are briefly discussed, and representative analytical results are presented and compared to the test data. The results that are presented demonstrate the feasibility of performing inelastic analyses, and they are indicative of the general level of agreement that the analyst might expect when using conventional inelastic analysis procedures. (U.S.)

New Improvements in Mixture Self-Shielding Treatment with APOLLO2 Code

International Nuclear Information System (INIS)

Coste-Delclaux, M.

2006-01-01

Full text of the presentation follows: APOLLO2 is a modular multigroup transport code developed at the CEA in Saclay (France). Previously, the self-shielding module could only treat one resonant isotope mixed with moderator isotopes. Consequently, the resonant mixture self-shielding treatment was an iterative one. Each resonant isotope of the mixture was treated separately, the other resonant isotopes of the mixture being then considered as moderator isotopes, that is to say non-resonant isotopes. This treatment could be iterated. Recently, we have developed a new method that consists in treating the resonant mixture as a unique entity. A main feature of APOLLO2 self-shielding module is that some implemented models are very general and therefore very powerful and versatile. We can give, as examples, the use of probability tables in order to describe the microscopic cross-section fluctuations or the TR slowing-down model that can deal with any resonance shape. The self-shielding treatment of a resonant mixture was developed essentially thanks to these two models. The goal of this paper is to describe the improvements on the self-shielding treatment of a resonant mixture and to present, as an application, the calculation of the ATRIUM-10 BWR benchmark. We will conclude by some prospects on remaining work in the self-shielding domain. (author)

Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks

International Nuclear Information System (INIS)

Mosteller, R.D.; Little, R.C.

1998-01-01

Versions of MCNP up through and including 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. This paper presents MCNP results for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment

Shielding concerns at a spallation source

International Nuclear Information System (INIS)

Russell, G.J.; Robinson, H.; Legate, G.L.; Woods, R.

1989-01-01

Neutrons produced by 800-MeV proton reactions at the Los Alamos Neutron Scattering Center spallation neutron source cause a variety of challenging shielding problems. We identify several characteristics distinctly different from reactor shielding and compute the dose attenuation through an infinite slab/shield composed of iron (100 cm) and borated polyethylene (15 cm). Our calculations show that (for an incident spallation spectrum characteristic of neutrons leaking from a tungsten target at 90/degree/) the dose through the shield is a complex mixture of neutrons and gamma rays. High-energy (> 20 MeV) neutron production from the target is ≅5% of the total, yet causes ≅68% of the dose at the shield surface. Primary low-energy (< 20 MeV) neutrons from the target contribute negligibly (≅0.5%) to the dose at the shield surface yet cause gamma rays, which contribute ≅31% to the total dose at the shield surface. Low-energy neutrons from spallation reactions behave similarly to neutrons with a fission spectrum distribution. 6 refs., 8 figs., 1 tab

Interior beam searchlight semi-analytical benchmark

International Nuclear Information System (INIS)

Ganapol, Barry D.; Kornreich, Drew E.

2008-01-01

Multidimensional semi-analytical benchmarks to provide highly accurate standards to assess routine numerical particle transport algorithms are few and far between. Because of the well-established 1D theory for the analytical solution of the transport equation, it is sometimes possible to 'bootstrap' a 1D solution to generate a more comprehensive solution representation. Here, we consider the searchlight problem (SLP) as a multidimensional benchmark. A variation of the usual SLP is the interior beam SLP (IBSLP) where a beam source lies beneath the surface of a half space and emits directly towards the free surface. We consider the establishment of a new semi-analytical benchmark based on a new FN formulation. This problem is important in radiative transfer experimental analysis to determine cloud absorption and scattering properties. (authors)

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

Science.gov (United States)

Mitchell, L

1996-01-01

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

Scale-4 shipping cask shielding applications

International Nuclear Information System (INIS)

Broadhead, B.L.; Parks, C.V.

1991-01-01

This paper reports the application of the SCALE-4 shielding sequences SAS1 and SAS4 to the problem set distributed by the Organization for Economic Cooperation and Development (OECD) Working Group on Shielding Assessment of Transportation Packages. In many cases, additional comparison are made with MCNP and QADS solutions to provide a complete cross-check of methods, cross sections, geometry, etc. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask

Benchmarking multi-dimensional large strain consolidation analyses

International Nuclear Information System (INIS)

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

2010-01-01

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

Analysis of a multigroup stylized CANDU half-core benchmark

International Nuclear Information System (INIS)

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

2011-01-01

Highlights: → This paper provides a benchmark that is a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. → An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core CANDU benchmark problem. → Reference eigenvalues and selected pin and bundle fission rates are included. → 2-, 4- and 47-group Monte Carlo solutions are compared to analyze homogenization-free transport approximations that result from energy condensation. - Abstract: An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem. Reference eigenvalues and selected pin and bundle fission rates are also included. This benchmark is intended to provide computational reactor physicists and methods developers with a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. In addition to transport theory code verification, the 8-group energy structure provides reactor physicist with an ideal problem for examining cross section homogenization and collapsing effects in a full-core environment. To this end, additional 2-, 4- and 47-group full-core Monte Carlo benchmark solutions are compared to analyze homogenization-free transport approximations incurred as a result of energy group condensation.

Resonance Self-Shielding Methodologies in SCALE 6

International Nuclear Information System (INIS)

Williams, Mark L.

2011-01-01

SCALE 6 includes several problem-independent multigroup (MG) libraries that were processed from the evaluated nuclear data file ENDF/B using a generic flux spectrum. The library data must be self-shielded and corrected for problem-specific spectral effects for use in MG neutron transport calculations. SCALE 6 computes problem-dependent MG cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic continuous-energy (CE) calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The CE calculation can be performed using an infinite medium approximation, a simplified two-region method for lattices, or a one-dimensional discrete ordinates transport calculation with pointwise (PW) cross-section data. This paper describes the SCALE-resonance self-shielding methodologies, including the deterministic calculation of the CE flux spectra using PW nuclear data and the method for using CE spectra to produce problem-specific MG cross sections for various configurations (including doubly heterogeneous lattices). It also presents results of verification and validation studies.

Benchmark experiments of dose distributions in phantom placed behind iron and concrete shields at the TIARA facility

International Nuclear Information System (INIS)

Nakane, Yoshihiro; Sakamoto, Yukio; Tsuda, Shuichi

2004-01-01

To verify the calculation methods used for the evaluations of neutron dose at the radiation shielding design of the high-intensity proton accelerator facility (J-PARC), dose distributions in a plastic phantom of 30x30x30 cm 3 slab placed behind iron and concrete test shields were measured by using a tissue equivalent proportional counter for 65-MeV quasi-monoenergetic neutrons generated from the 7 Li(p,n) reactions with 68-MeV protons at the TIARA facility. Dose distributions in the phantom were calculated by using the MCNPX and the NMTC/JAM-MCNP codes with the flux-to-dose conversion coefficients prepared for the shielding design of the facility. The comparison results show the calculated results were in good agreement with the measured ones within 20%. (author)

SKaMPI: A Comprehensive Benchmark for Public Benchmarking of MPI

Directory of Open Access Journals (Sweden)

Ralf Reussner

2002-01-01

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

Shielding research in France

Energy Technology Data Exchange (ETDEWEB)

Lafore, P

1964-10-01

Shielding research as an independent subject in France dates from 1956. The importance of these studies has been reflected in the contribution which they have made to power reactor design and in the resultant savings in expenditure for civil engineering and machinery for the removal of mobile shields. The Reactor Shielding Research Division numbers approximately 60 persons and uses several experimental facilities. These include: NAIADE I, installed near the ZOE reactor and operating with a natural uranium slab 2 cm thick (an effective diameter of 60 cm is the one most commonly used); the TRITON pool-type reactor, mainly used in shielding studies, includes an active-water loop, by means of which the secondary shields required for light-water reactors can be studied; core, NEREIDE, which is situated near a 2 m x 2 m aluminium window enables a large neutron source to be placed in a compartment without water in which large-scale mock-ups can be mounted for the study, in particular, of neutron diffusion in large cavities, and of reactor shielding of greater thickness than that in NAIADE I; SAMES 600 keV accelerator is used for monoenergetic neutron studies. Instrumentation studies are an important part of the work, mainly in the measurement of fast neutrons and their spectra by activation detectors. Of late, attention has been directed towards the use of (n, n') (rhodium) reactions and of heavy detectors for low-flux measurements. The simultaneous use of a large number of detectors poses automation problems. With our installation we can count 16 detectors simultaneously. Neutron spectrum studies are conducted with nuclear emulsions and a lithium-6 semiconductor spectrometer. As to the materials used, the research carried out in France involves chiefly graphite, iron and concrete at various temperatures up to 800 deg C. Different compounds, borated and non-borated and of densities up to between 1 and 9 are under consideration. Problems connected with applications are

Muon shielding for PEP

International Nuclear Information System (INIS)

Jenkins, T.M.; Thomas, R.H.

1974-01-01

The first stage of construction of PEP will consist of electron and positron storage rings. At a later date a 200 GeV proton storage ring may be added. It is judicious therefore, to ensure that the first and second phases of construction are compatible with each other. One of several factors determining the elevation at which the storage rings will be constructed is the necessity to provide adequate radiation shielding. The overhead shielding of PEP is determined by the reproduction of neutrons in the hadron cascade generated by primary protons lost from the storage ring. The minimum overburden planned for PEP is 5.5 meters of earth (1100 gm cm/sup /minus/2/). To obtain a rough estimate of the magnitude of the muon radiation problem this note presents some preliminary calculations. Their purpose is intended merely to show that the presently proposed design for PEP will present no major shielding problems should the protons storage ring be installed. More detailed calculations will be made using muon yield computer codes developed at CERN and NAL and muon transport codes developed at SLAC, when details of the proton storage ring become settled. 9 refs., 4 figs

Secondary gamma-ray data for shielding calculation

International Nuclear Information System (INIS)

Miyasaka, Sunichi

1979-01-01

In deep penetration transport calculations, the integral design parameters is determined mainly by secondary particles which are produced by interactions of the primary radiation with materials. The shield thickness and the biological dose rate at a given point of a bulk shield are determined from the contribution from secondary gamma rays. The heat generation and the radiation damage in the structural and shield materials depend strongly on the secondary gamma rays. In this paper, the status of the secondary gamma ray data and its further problems are described from the viewpoint of shield design. The secondary gamma-ray data in ENDF/B-IV and POPOP4 are also discussed based on the test calculations made for several shield assemblies. (author)

Fast reactor shield sensitivity studies for steel--sodium--iron systems

International Nuclear Information System (INIS)

Oblow, E.M.; Weisbin, C.R.

1977-01-01

A study was made of the adequacy of the current ENDF/B-IV sodium and iron neutron cross section data files for fast reactor shield design work. Experimental data from 21 fast reactor shield configurations containing large thicknesses of steel, sodium, and iron were analyzed with discrete ordinates calculations and sensitivity methods to assess the data files. This study represents the largest full-scale sensitivity analysis of benchmark quality experimental data to date. Included in the sensitivity studies were the results of the new cross section adjustment algorithms added to the FORSS code system. Conclusions were drawn about the need for more accurate data for sodium and iron elastic and discrete inelastic cross sections above 1 MeV and the values of the total cross section in the vicinity of important minima

Methods for calculating radiation attenuation in shields

Energy Technology Data Exchange (ETDEWEB)

Butler, J; Bueneman, D; Etemad, A; Lafore, P; Moncassoli, A M; Penkuhn, H; Shindo, M; Stoces, B

1964-10-01

In recent years the development of high-speed digital computers of large capacity has revolutionized the field of reactor shield design. For compact special-purpose reactor shields, Monte-Carlo codes in two- and three dimensional geometries are now available for the proper treatment of both the neutron and gamma- ray problems. Furthermore, techniques are being developed for the theoretical optimization of minimum-weight shield configurations for this type of reactor system. In the design of land-based power reactors, on the other hand, there is a strong incentive to reduce the capital cost of the plant, and economic considerations are also relevant to reactors designed for merchant ship propulsion. In this context simple methods are needed which are economic in their data input and computing time requirements and which, at the same time, are sufficiently accurate for design work. In general the computing time required for Monte-Carlo calculations in complex geometry is excessive for routine design calculations and the capacity of the present codes is inadequate for the proper treatment of large reactor shield systems in three dimensions. In these circumstances a wide range of simpler techniques are currently being employed for design calculations. The methods of calculation for neutrons in reactor shields fall naturally into four categories: Multigroup diffusion theory; Multigroup diffusion with removal sources; Transport codes; and Monte Carlo methods. In spite of the numerous Monte- Carlo techniques which are available for penetration and back scattering, serious problems are still encountered in practice with the scattering of gamma rays from walls of buildings which contain critical facilities and also concrete-lined discharge shafts containing irradiated fuel elements. The considerable volume of data in the unclassified literature on the solution of problems of this type in civil defence work appears not to have been evaluated for reactor shield design. In

IAEA sodium void reactivity benchmark calculations

International Nuclear Information System (INIS)

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

1992-01-01

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

Accuracy evaluation of the current data and method applied to shielding design of the Fusion Experimental Reactor (FER)

International Nuclear Information System (INIS)

Mori, Seiji; Kobayashi, Takeshi; Seki, Yasushi

1988-06-01

Shielding benchmarking study of the current data and method applied to the Fusion Experimental Reactor (FER) was performed. First, neutron and gamma ray fluxes were calculated by the one-dimensional S N code using various cross section libraries and the continuous energy Monte Carlo code. The results were compared in terms of the S N /MC ratio. The worst ratios are about 0.5 and 0.25 for neutron flux and gamma ray flux, respectively. Next, the analytical calculations of the iron sphere transmission experiment of 14 MeV neutrons were performed to examine the accuracy of cross section data of iron, which is the most important material of shield. The E/C ratio is larger than 2 even if the continuous energy Monte Carlo code was used. Thirdly, the influence of geometrical representation of the shield was investigated by comparing the homogeneous model and the heterogeneous model (alternating layers of SS316 and water). As a result, it was made clear that the homogeneous model underestimates neutron flux by a factor of 2. Finally, the necessity of benchmark experiment and improvement of cross section library was pointed out as the further R and D issues. (author)

A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

OpenAIRE

Hoff, Gabriela; Costa, Paulo Roberto

2013-01-01

The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and com...

Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

Energy Technology Data Exchange (ETDEWEB)

Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

Analytical theory of coherent synchrotron radiation wakefield of short bunches shielded by conducting parallel plates

Energy Technology Data Exchange (ETDEWEB)

Stupakov, Gennady; Zhou, Demin

2016-04-21

We develop a general model of coherent synchrotron radiation (CSR) impedance with shielding provided by two parallel conducting plates. This model allows us to easily reproduce all previously known analytical CSR wakes and to expand the analysis to situations not explored before. It reduces calculations of the impedance to taking integrals along the trajectory of the beam. New analytical results are derived for the radiation impedance with shielding for the following orbits: a kink, a bending magnet, a wiggler of finite length, and an infinitely long wiggler. All our formulas are benchmarked against numerical simulations with the CSRZ computer code.

The development of code benchmarks

International Nuclear Information System (INIS)

Glass, R.E.

1986-01-01

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

Sensitivity Analysis of OECD Benchmark Tests in BISON

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-01

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

Benchmarks for dynamic multi-objective optimisation

CSIR Research Space (South Africa)

Helbig, M

2013-06-01

Full Text Available When algorithms solve dynamic multi-objective optimisation problems (DMOOPs), benchmark functions should be used to determine whether the algorithm can overcome specific difficulties that can occur in real-world problems. However, for dynamic multi...

PWR upper/lower internals shield

Energy Technology Data Exchange (ETDEWEB)

Homyk, W.A. [Indian Point Station, Buchanan, NY (United States)

1995-03-01

During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use of lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.

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.

Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

International Nuclear Information System (INIS)

Moukhamadeev, R.; Suvorov, A.

2000-01-01

This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

Development of radiation shielding standards in the American Nuclear Society

International Nuclear Information System (INIS)

Trubey, D.K.

1975-11-01

The American Nuclear Society (ANS) is a standards-writing organization-member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Shielding, whose charge is to establish standards in connection with radiation protection and shielding, to provide shielding information to other standards writing groups, and to prepare recommended sets of shielding data and test problems. This paper is a progress report of this subcommittee

WWER in-core fuel management benchmark definition

International Nuclear Information System (INIS)

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

1994-01-01

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

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.

Radiation protection and shielding standards for the 1980s

International Nuclear Information System (INIS)

Trubey, D.K.

1982-01-01

The American Nuclear Society (ANS) is a standards-writing organization member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Radiation Protection and Shielding, whose charge is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. This paper is a progress report of this subcommittee. Significant progress has been made since the last comprehensive report to the Society

Resonance shielding in thermal reactor lattices

International Nuclear Information System (INIS)

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

1982-01-01

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

International Handbook of Evaluated Criticality Safety Benchmark Experiments - ICSBEP (DVD), Version 2013

International Nuclear Information System (INIS)

2013-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical experiment facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span nearly 66,000 pages and contain 558 evaluations with benchmark specifications for 4,798 critical, near critical or subcritical configurations, 24 criticality alarm placement/shielding configurations with multiple dose points for each and 200 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the Handbook are benchmark specifications for Critical, Bare, HEU(93.2)- Metal Sphere experiments referred to as ORSphere that were performed by a team of experimenters at Oak Ridge National Laboratory in the early 1970's. A photograph of this assembly is shown on the front cover

The extent of benchmarking in the South African financial sector

OpenAIRE

W Vermeulen

2014-01-01

Benchmarking is the process of identifying, understanding and adapting outstanding practices from within the organisation or from other businesses, to help improve performance. The importance of benchmarking as an enabler of business excellence has necessitated an in-depth investigation into the current state of benchmarking in South Africa. This research project highlights the fact that respondents realise the importance of benchmarking, but that various problems hinder the effective impleme...

Radiation protection/shield design: a need for a systems approach

International Nuclear Information System (INIS)

Disney, R.K.

1977-01-01

Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. The system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection is described, and the program developed to implement this approach is defined. In addition, the principal shielding design problems for LMFBR nuclear reactor systems are discussed in relation to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods are discussed

PMLB: a large benchmark suite for machine learning evaluation and comparison.

Science.gov (United States)

Olson, Randal S; La Cava, William; Orzechowski, Patryk; Urbanowicz, Ryan J; Moore, Jason H

2017-01-01

The selection, development, or comparison of machine learning methods in data mining can be a difficult task based on the target problem and goals of a particular study. Numerous publicly available real-world and simulated benchmark datasets have emerged from different sources, but their organization and adoption as standards have been inconsistent. As such, selecting and curating specific benchmarks remains an unnecessary burden on machine learning practitioners and data scientists. The present study introduces an accessible, curated, and developing public benchmark resource to facilitate identification of the strengths and weaknesses of different machine learning methodologies. We compare meta-features among the current set of benchmark datasets in this resource to characterize the diversity of available data. Finally, we apply a number of established machine learning methods to the entire benchmark suite and analyze how datasets and algorithms cluster in terms of performance. From this study, we find that existing benchmarks lack the diversity to properly benchmark machine learning algorithms, and there are several gaps in benchmarking problems that still need to be considered. This work represents another important step towards understanding the limitations of popular benchmarking suites and developing a resource that connects existing benchmarking standards to more diverse and efficient standards in the future.

The extent of benchmarking in the South African financial sector

Directory of Open Access Journals (Sweden)

W Vermeulen

2014-06-01

Full Text Available Benchmarking is the process of identifying, understanding and adapting outstanding practices from within the organisation or from other businesses, to help improve performance. The importance of benchmarking as an enabler of business excellence has necessitated an in-depth investigation into the current state of benchmarking in South Africa. This research project highlights the fact that respondents realise the importance of benchmarking, but that various problems hinder the effective implementation of benchmarking. Based on the research findings, recommendations for achieving success are suggested.

2010 Criticality Accident Alarm System Benchmark Experiments At The CEA Valduc SILENE Facility

International Nuclear Information System (INIS)

Miller, Thomas Martin; Dunn, Michael E.; Wagner, John C.; McMahan, Kimberly L.; Authier, Nicolas; Jacquet, Xavier; Rousseau, Guillaume; Wolff, Herve; Piot, Jerome; Savanier, Laurence; Baclet, Nathalie; Lee, Yi-kang; Masse, Veronique; Trama, Jean-Christophe; Gagnier, Emmanuel; Naury, Sylvie; Lenain, Richard; Hunter, Richard; Kim, Soon; Dulik, George Michael; Reynolds, Kevin H.

2011-01-01

Several experiments were performed at the CEA Valduc SILENE reactor facility, which are intended to be published as evaluated benchmark experiments in the ICSBEP Handbook. These evaluated benchmarks will be useful for the verification and validation of radiation transport codes and evaluated nuclear data, particularly those that are used in the analysis of CAASs. During these experiments SILENE was operated in pulsed mode in order to be representative of a criticality accident, which is rare among shielding benchmarks. Measurements of the neutron flux were made with neutron activation foils and measurements of photon doses were made with TLDs. Also unique to these experiments was the presence of several detectors used in actual CAASs, which allowed for the observation of their behavior during an actual critical pulse. This paper presents the preliminary measurement data currently available from these experiments. Also presented are comparisons of preliminary computational results with Scale and TRIPOLI-4 to the preliminary measurement data.

Shielding calculational system for plutonium

International Nuclear Information System (INIS)

Zimmerman, M.G.; Thomsen, D.H.

1975-08-01

A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

Comparison of calculational methods for liquid metal reactor shields

International Nuclear Information System (INIS)

Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.

1985-09-01

A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs

Analyzing the BBOB results by means of benchmarking concepts.

Science.gov (United States)

Mersmann, O; Preuss, M; Trautmann, H; Bischl, B; Weihs, C

2015-01-01

We present methods to answer two basic questions that arise when benchmarking optimization algorithms. The first one is: which algorithm is the "best" one? and the second one is: which algorithm should I use for my real-world problem? Both are connected and neither is easy to answer. We present a theoretical framework for designing and analyzing the raw data of such benchmark experiments. This represents a first step in answering the aforementioned questions. The 2009 and 2010 BBOB benchmark results are analyzed by means of this framework and we derive insight regarding the answers to the two questions. Furthermore, we discuss how to properly aggregate rankings from algorithm evaluations on individual problems into a consensus, its theoretical background and which common pitfalls should be avoided. Finally, we address the grouping of test problems into sets with similar optimizer rankings and investigate whether these are reflected by already proposed test problem characteristics, finding that this is not always the case.

PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

International Nuclear Information System (INIS)

Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

1990-01-01

1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

2014-10-15

In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

Performances of Kevlar and Polyethylene as radiation shielding on-board the International Space Station in high latitude radiation environment.

Science.gov (United States)

Narici, Livio; Casolino, Marco; Di Fino, Luca; Larosa, Marianna; Picozza, Piergiorgio; Rizzo, Alessandro; Zaconte, Veronica

2017-05-10

Passive radiation shielding is a mandatory element in the design of an integrated solution to mitigate the effects of radiation during long deep space voyages for human exploration. Understanding and exploiting the characteristics of materials suitable for radiation shielding in space flights is, therefore, of primary importance. We present here the results of the first space-test on Kevlar and Polyethylene radiation shielding capabilities including direct measurements of the background baseline (no shield). Measurements are performed on-board of the International Space Station (Columbus modulus) during the ALTEA-shield ESA sponsored program. For the first time the shielding capability of such materials has been tested in a radiation environment similar to the deep-space one, thanks to the feature of the ALTEA system, which allows to select only high latitude orbital tracts of the International Space Station. Polyethylene is widely used for radiation shielding in space and therefore it is an excellent benchmark material to be used in comparative investigations. In this work we show that Kevlar has radiation shielding performances comparable to the Polyethylene ones, reaching a dose rate reduction of 32 ± 2% and a dose equivalent rate reduction of 55 ± 4% (for a shield of 10 g/cm 2 ).

Storage-Intensive Supercomputing Benchmark Study

Energy Technology Data Exchange (ETDEWEB)

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

2007-10-30

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

Performance evaluation of firefly algorithm with variation in sorting for non-linear benchmark problems

Science.gov (United States)

Umbarkar, A. J.; Balande, U. T.; Seth, P. D.

2017-06-01

The field of nature inspired computing and optimization techniques have evolved to solve difficult optimization problems in diverse fields of engineering, science and technology. The firefly attraction process is mimicked in the algorithm for solving optimization problems. In Firefly Algorithm (FA) sorting of fireflies is done by using sorting algorithm. The original FA is proposed with bubble sort for ranking the fireflies. In this paper, the quick sort replaces bubble sort to decrease the time complexity of FA. The dataset used is unconstrained benchmark functions from CEC 2005 [22]. The comparison of FA using bubble sort and FA using quick sort is performed with respect to best, worst, mean, standard deviation, number of comparisons and execution time. The experimental result shows that FA using quick sort requires less number of comparisons but requires more execution time. The increased number of fireflies helps to converge into optimal solution whereas by varying dimension for algorithm performed better at a lower dimension than higher dimension.

DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program

International Nuclear Information System (INIS)

Courtney, J. C.

1987-01-01

1 - Description of problem or function: DEMONR treats the behavior of neutrons in a slab shield. It is frequently used as a teaching tool. 2 - Method of solution: An unbiased Monte Carlo code calculates the number, energy, and direction of neutrons that penetrate or are reflected from a shield. 3 - Restrictions on the complexity of the problem: Only one shield may be used in each problem. The shield material may be a single element or a homogeneous mixture of elements with a single effective atomic weight. Only elastic scattering and neutron capture processes are allowed. The source is a point located on one face of the slab. It provides a cosine distribution of current. Monoenergetic or fission spectrum neutrons may be selected

Piping structural design for the ITER thermal shield manifold

Energy Technology Data Exchange (ETDEWEB)

Noh, Chang Hyun, E-mail: chnoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Chung, Wooho, E-mail: whchung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo; Kang, Kyoung-O. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bae, Jing Do; Cha, Jong Kook [Korea Marine Equipment Research Institute, Busan 606-806 (Korea, Republic of); Kim, Kyoung-Kyu [Mecha T& S, Jinju-si 660-843 (Korea, Republic of); Hamlyn-Harris, Craig; Hicks, Robby; Her, Namil; Jun, Chang-Hoon [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

2015-10-15

Highlights: • We finalized piping design of ITER thermal shield manifold for procurement. • Support span is determined by stress and deflection limitation. • SQP, which is design optimization method, is used for the pipe design. • Benchmark analysis is performed to verify the analysis software. • Pipe design is verified by structural analyses. - Abstract: The thermal shield (TS) provides the thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2 K. The TS is actively cooled by 80 K pressurized helium gas which flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the manifold piping design and analysis for the ITER thermal shield. First, maximum allowable span for the manifold support is calculated based on the simple beam theory. In order to accommodate the thermal contraction in the manifold feeder, a contraction loop is designed and applied. Sequential Quadratic Programming (SQP) method is used to determine the optimized dimensions of the contraction loop to ensure adequate flexibility of manifold pipe. Global structural behavior of the manifold is investigated when the thermal movement of the redundant (un-cooled) pipe is large.

Thick Galactic Cosmic Radiation Shielding Using Atmospheric Data

Science.gov (United States)

Youngquist, Robert C.; Nurge, Mark A.; Starr, Stanley O.; Koontz, Steven L.

2013-01-01

NASA is concerned with protecting astronauts from the effects of galactic cosmic radiation and has expended substantial effort in the development of computer models to predict the shielding obtained from various materials. However, these models were only developed for shields up to about 120 g!cm2 in thickness and have predicted that shields of this thickness are insufficient to provide adequate protection for extended deep space flights. Consequently, effort is underway to extend the range of these models to thicker shields and experimental data is required to help confirm the resulting code. In this paper empirically obtained effective dose measurements from aircraft flights in the atmosphere are used to obtain the radiation shielding function of the earth's atmosphere, a very thick shield. Obtaining this result required solving an inverse problem and the method for solving it is presented. The results are shown to be in agreement with current code in the ranges where they overlap. These results are then checked and used to predict the radiation dosage under thick shields such as planetary regolith and the atmosphere of Venus.

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)

Application of a nodal collocation approximation for the multidimensional PL equations to the 3D Takeda benchmark problems

International Nuclear Information System (INIS)

Capilla, M.; Talavera, C.F.; Ginestar, D.; Verdú, G.

2012-01-01

Highlights: ► The multidimensional P L approximation to the nuclear transport equation is reviewed. ► A nodal collocation method is developed for the spatial discretization of P L equations. ► Advantages of the method are lower dimension and good characterists of the associated algebraic eigenvalue problem. ► The P L nodal collocation method is implemented into the computer code SHNC. ► The SHNC code is verified with 2D and 3D benchmark eigenvalue problems from Takeda and Ikeda, giving satisfactory results. - Abstract: P L equations are classical approximations to the neutron transport equations, which are obtained expanding the angular neutron flux in terms of spherical harmonics. These approximations are useful to study the behavior of reactor cores with complex fuel assemblies, for the homogenization of nuclear cross-sections, etc., and most of these applications are in three-dimensional (3D) geometries. In this work, we review the multi-dimensional P L equations and describe a nodal collocation method for the spatial discretization of these equations for arbitrary odd order L, which is based on the expansion of the spatial dependence of the fields in terms of orthonormal Legendre polynomials. The performance of the nodal collocation method is studied by means of obtaining the k eff and the stationary power distribution of several 3D benchmark problems. The solutions are obtained are compared with a finite element method and a Monte Carlo method.

Insufficient self-shielding correction in VITAMIN-B6

International Nuclear Information System (INIS)

Konno, Chikara; Ochiai, Kentaro; Ohnishi, Seiki

2011-01-01

We carried out a simple benchmark calculation test with a multigroup cross-section library VITAMIN-B6 generated from ENDF/B-VI. The model of this test consisted of an iron sphere of 1 m in radius with an isotropic 20 MeV neutron source in the center. Neutron spectra in the sphere were calculated with an Sn code ANISN and VITAMIN-B6 or FENDL/MG-1.1. A calculation with MCNP and ENDF/B-VI was carried out as a reference. The neutron spectra with ANISN and FENDL/MG-1.1 agreed with those with MCNP, while those with ANISN and VITAMIN-B6 were at most 50% different from those with MCNP. We uncovered that the discrepancy came from insufficient self-shielding correction due to the followings; 1) The smallest background cross section of 56 Fe in VITAMIN-B6 is 1. 2) The weighting flux used in generating VITAMIN-B6 is not adequate. VITAMIN-B6 should be revised for adequate self-shielding correction. (author)

Analysis of shield for the nuclear ship MUTSU

International Nuclear Information System (INIS)

Fuse, Takayoshi; Takeuchi, Kiyoshi; Yamaji, Akio

1975-01-01

On the nuclear ship MUTSU, a higher-than-expected level of radiation was found, with output raised to 1.4 per cent. To investigate the radiation leakage, the analysis of the shielding problem utilized a four-step sequence of PALLAS-2DCY cylindrical r-z calculations with fixed sources distributions in the core. The neutron dose contours show the importance of streaming in the gap between the reactor vessel and the primary shield. Dominant consideration of thermal insulation exclude shielding from this area resulting in an imbalance in the shielding effectiveness. The neutron dose rate at the upper part of the reactor vessel is increased by neutrons incident on the head from cavity scattering. The calculation indicates that the neutron dose rate at the top of the primary shield is 5 rem/hr at 100 per cent output. (auth.)

Radiation shielding device

International Nuclear Information System (INIS)

Nakagawa, Takahiro; Yamagami, Makoto.

1996-01-01

A fixed shielding member made of a radiation shielding material is constituted in perpendicular to an opening formed on radiation shielding walls. The fixed shielding member has one side opened and has other side, the upper portion and the lower portion disposed in close contact with the radiation shielding walls. Movable shielding members made of a radiation shielding material are each disposed openably on both side of the fixed shielding member. The movable shielding member has a shaft as a fulcrum on one side thereof for connecting it to the radiation shielding walls. The other side has a handle attached for opening/closing the movable shielding member. Upon access of an operator, when each one of the movable shielding members is opened/closed on every time, leakage of linear or scattered radiation can be prevented. Even when both of the movable shielding members are opened simultaneously, the fixed shielding member and the movable shielding members form labyrinth to prevent leakage of linear radioactivity. (I.N.)

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

Directory of Open Access Journals (Sweden)

Pescarini Massimo

2016-01-01

Full Text Available The PCA-Replica 12/13 (H2O/Fe neutron shielding benchmark experiment was analysed using the TORT-3.2 3D SN code. PCA-Replica reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1 and UGENDF70.BOLIB (ENDF/B-VII.0 libraries and the ORNL BUGLE-B7 (ENDF/B-VII.0 library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n′Rh-103m, In-115(n,n′In-115m and S-32(n,pP-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.

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

Science.gov (United States)

Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela

2016-03-01

The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the TORT-3.2 3D SN code. PCA-Replica reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and UGENDF70.BOLIB (ENDF/B-VII.0) libraries and the ORNL BUGLE-B7 (ENDF/B-VII.0) library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n')Rh-103m, In-115(n,n')In-115m and S-32(n,p)P-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.

Comparison of neutron fluxes obtained by 2-D and 3-D geometry with different shielding libraries in biological shield of the TRIGA MARK II reactor

International Nuclear Information System (INIS)

Bozic, M.; Zagar, T.; Ravnik, M.

2003-01-01

Neutron fluxes in different spatial locations in biological shield are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Libraries used with TORT code were BUGLE-96 library (coupled library with 47 neutron groups and 20 gamma groups) and VITAMIN-B6 library (coupled library with 199 neutron groups and 42 gamma groups). BUGLE-96 library is derived from VITAMIN-B6 library. 2-D and 3-D models for homogeneous type of problem (without inserted beam port 4) and problem with asymmetry (non-homogeneous problem; inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. The main purpose is to verify the possibility for using 2-D approximation model instead of large 3-D model in some calculations. Another purpose of this paper was to compare neutron spectral constants obtained from neutron fluxes (3-D model) determined with smaller BUGLE-96 library with new constants obtained from fluxes calculated with bigger VITAMIN-B6 library. These neutron spectral constants are used in isotopic calculation with SCALE code package (ORIGEN-S). In past only neutron spectral constants determined by neutron fluxes from BUGLE-96 library were used. Experimental results used for isotopic composition comparison are available from irradiation experiment with selected type of concrete and other materials in beam port 4 (irradiation channel 4) in TRIGA Mark II reactor. These experimental results were used as a benchmark in this paper. (author)

A comparison of two efficient nonlinear heat conduction methodologies using a two-dimensional time-dependent benchmark problem

International Nuclear Information System (INIS)

Wilson, G.L.; Rydin, R.A.; Orivuori, S.

1988-01-01

Two highly efficient nonlinear time-dependent heat conduction methodologies, the nonlinear time-dependent nodal integral technique (NTDNT) and IVOHEAT are compared using one- and two-dimensional time-dependent benchmark problems. The NTDNT is completely based on newly developed time-dependent nodal integral methods, whereas IVOHEAT is based on finite elements in space and Crank-Nicholson finite differences in time. IVOHEAT contains the geometric flexibility of the finite element approach, whereas the nodal integral method is constrained at present to Cartesian geometry. For test problems where both methods are equally applicable, the nodal integral method is approximately six times more efficient per dimension than IVOHEAT when a comparable overall accuracy is chosen. This translates to a factor of 200 for a three-dimensional problem having relatively homogeneous regions, and to a smaller advantage as the degree of heterogeneity increases

Aircraft Engine Gas Path Diagnostic Methods: Public Benchmarking Results

Science.gov (United States)

Simon, Donald L.; Borguet, Sebastien; Leonard, Olivier; Zhang, Xiaodong (Frank)

2013-01-01

Recent technology reviews have identified the need for objective assessments of aircraft engine health management (EHM) technologies. To help address this issue, a gas path diagnostic benchmark problem has been created and made publicly available. This software tool, referred to as the Propulsion Diagnostic Method Evaluation Strategy (ProDiMES), has been constructed based on feedback provided by the aircraft EHM community. It provides a standard benchmark problem enabling users to develop, evaluate and compare diagnostic methods. This paper will present an overview of ProDiMES along with a description of four gas path diagnostic methods developed and applied to the problem. These methods, which include analytical and empirical diagnostic techniques, will be described and associated blind-test-case metric results will be presented and compared. Lessons learned along with recommendations for improving the public benchmarking processes will also be presented and discussed.

Intercomparison of the finite difference and nodal discrete ordinates and surface flux transport methods for a LWR pool-reactor benchmark problem in X-Y geometry

International Nuclear Information System (INIS)

O'Dell, R.D.; Stepanek, J.; Wagner, M.R.

1983-01-01

The aim of the present work is to compare and discuss the three of the most advanced two dimensional transport methods, the finite difference and nodal discrete ordinates and surface flux method, incorporated into the transport codes TWODANT, TWOTRAN-NODAL, MULTIMEDIUM and SURCU. For intercomparison the eigenvalue and the neutron flux distribution are calculated using these codes in the LWR pool reactor benchmark problem. Additionally the results are compared with some results obtained by French collision probability transport codes MARSYAS and TRIDENT. Because the transport solution of this benchmark problem is close to its diffusion solution some results obtained by the finite element diffusion code FINELM and the finite difference diffusion code DIFF-2D are included

A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

Energy Technology Data Exchange (ETDEWEB)

Hoff, Gabriela [Pontificia Univ. Catolica do Rio Grande do Sul (PUCRS), Porto Alegre, RS (Brazil). Grupo de Experimentacao e Simulacao Computacional em Fisica Medica; Costa, Paulo Roberto, E-mail: pcosta@if.usp.br [Universidade de Sao Paulo (IF/USP), SP (Brazil). Dept. de Fisica Nuclear. Lab. de Dosimetria das Radiacoes e Fisica Medica

2013-03-15

The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and compared to The American Association of Physicists in Medicine (AAPM) data for lead. Their influence on the transmission curves as well the study of the influence of the shielding material composition and beam geometry on the outgoing spectra were performed. Characteristics of transmitted spectra, such as shape, average energy and Half-Value Layer (HVL), were also evaluated. The Geant4 toolkit benchmark for the energy resulting from the positron annihilation phenomena and its application in transmission curves description shown good agreement between data published by American Association on Physicists in Medicine task group 108 and experimental data published by Brazil. The transmission properties for different material compositions were also studied and have shown low dependency with the considered thicknesses. The broad and narrow beams configuration presented significant differences on the result. The fitting parameter for determining the transmission curves equations, according to Archer model is presented for different material. As conclusion were defined that beam geometry has significant influence and the composition has low influence on transmission curves for shielding design for the range of energy applied to PET. (author)

A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

International Nuclear Information System (INIS)

Hoff, Gabriela; Costa, Paulo Roberto

2013-01-01

The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and compared to The American Association of Physicists in Medicine (AAPM) data for lead. Their influence on the transmission curves as well the study of the influence of the shielding material composition and beam geometry on the outgoing spectra were performed. Characteristics of transmitted spectra, such as shape, average energy and Half-Value Layer (HVL), were also evaluated. The Geant4 toolkit benchmark for the energy resulting from the positron annihilation phenomena and its application in transmission curves description shown good agreement between data published by American Association on Physicists in Medicine task group 108 and experimental data published by Brazil. The transmission properties for different material compositions were also studied and have shown low dependency with the considered thicknesses. The broad and narrow beams configuration presented significant differences on the result. The fitting parameter for determining the transmission curves equations, according to Archer model is presented for different material. As conclusion were defined that beam geometry has significant influence and the composition has low influence on transmission curves for shielding design for the range of energy applied to PET. (author)

An Eigenstructure Assignment Approach to FDI for the Industrial Actuator Benchmark Test

DEFF Research Database (Denmark)

Jørgensen, R.B.; Patton, R.J.; Chen, J.

1995-01-01

This paper examines the robustness in modelling uncertainties of an observer-based fault detection and isolation scheme applied to the industrial actuator benchmark problem.......This paper examines the robustness in modelling uncertainties of an observer-based fault detection and isolation scheme applied to the industrial actuator benchmark problem....

Radiation shielding analysis

International Nuclear Information System (INIS)

Moon, S.H.; Ha, C.W.; Kwon, S.K.; Lee, J.K.; Choi, H.S.

1982-01-01

The theoretical bases of radiation streaming analysis in power reactors, such as ducts or reactor cavity, have been investigated. Discrete ordinates-Monte Carlo or Monte Carlo-Monte Carlo coupling techniques are suggested for the streaming analysis of ducts or reactor cavity. Single albedo scattering approximation code (SINALB) has been developed for simple and quick estimation of gamma-ray ceiling scattering, where the ceiling is assumed to be semi-infinite medium. This code has been employed to calculate the gamma-ray ceiling scattering effects in the laboratory containing a Co-60 source. The SINALB is applicable to gamma-ray scattering, only where the ceiling is thicker than Σsup(-1) and the height is at least twice higher than the shield wall. This code can be used for the purpose of preliminary radiation shield design. The MORSE code has been improved to analyze the gamma-ray scattering problem with on approximation method in respect to the random walk and estimation processes. This improved MORSE code has been employed to the gamma-ray ceiling scattering problem. The results of the improved MORSE calculation are in good agreement with the SINALB and standard MORSE. (Author)

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

Utilizing of computational tools on the modelling of a simplified problem of neutron shielding

Energy Technology Data Exchange (ETDEWEB)

Lessa, Fabio da Silva Rangel; Platt, Gustavo Mendes; Alves Filho, Hermes [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Inst. Politecnico]. E-mails: fsrlessa@gmail.com; gmplatt@iprj.uerj.br; halves@iprj.uerj.br

2007-07-01

In the current technology level, the investigation of several problems is studied through computational simulations whose results are in general satisfactory and much less expensive than the conventional forms of investigation (e.g., destructive tests, laboratory measures, etc.). Almost all of the modern scientific studies are executed using computational tools, as computers of superior capacity and their systems applications to make complex calculations, algorithmic iterations, etc. Besides the considerable economy in time and in space that the Computational Modelling provides, there is a financial economy to the scientists. The Computational Modelling is a modern methodology of investigation that asks for the theoretical study of the identified phenomena in the problem, a coherent mathematical representation of such phenomena, the generation of a numeric algorithmic system comprehensible for the computer, and finally the analysis of the acquired solution, or still getting use of pre-existent systems that facilitate the visualization of these results (editors of Cartesian graphs, for instance). In this work, was being intended to use many computational tools, implementation of numeric methods and a deterministic model in the study and analysis of a well known and simplified problem of nuclear engineering (the neutron transport), simulating a theoretical problem of neutron shielding with physical-material hypothetical parameters, of neutron flow in each space junction, programmed with Scilab version 4.0. (author)

Utilizing of computational tools on the modelling of a simplified problem of neutron shielding

International Nuclear Information System (INIS)

Lessa, Fabio da Silva Rangel; Platt, Gustavo Mendes; Alves Filho, Hermes

2007-01-01

In the current technology level, the investigation of several problems is studied through computational simulations whose results are in general satisfactory and much less expensive than the conventional forms of investigation (e.g., destructive tests, laboratory measures, etc.). Almost all of the modern scientific studies are executed using computational tools, as computers of superior capacity and their systems applications to make complex calculations, algorithmic iterations, etc. Besides the considerable economy in time and in space that the Computational Modelling provides, there is a financial economy to the scientists. The Computational Modelling is a modern methodology of investigation that asks for the theoretical study of the identified phenomena in the problem, a coherent mathematical representation of such phenomena, the generation of a numeric algorithmic system comprehensible for the computer, and finally the analysis of the acquired solution, or still getting use of pre-existent systems that facilitate the visualization of these results (editors of Cartesian graphs, for instance). In this work, was being intended to use many computational tools, implementation of numeric methods and a deterministic model in the study and analysis of a well known and simplified problem of nuclear engineering (the neutron transport), simulating a theoretical problem of neutron shielding with physical-material hypothetical parameters, of neutron flow in each space junction, programmed with Scilab version 4.0. (author)

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Shielding problems set by the use of a natural uranium target with a linear electron accelerator. Shielding and safety systems necessary

International Nuclear Information System (INIS)

Vialettes, Henry; Rocchesani, Jean; Lemure, Pierre

1971-06-01

The use of a natural uranium target for neutron production with a linear electron accelerator set special shielding problems due to the fact that, to standard photonuclear reactions, are added photoneutron induced photofission reactions giving rise to fission products of which the untimely liberation could cause very serious contamination problems. On the occasion of a recent accident on the target used with the Saclay 60 MeV linear accelerator, activity measurements were carried out on a certain number of samples taken. This revealed the presence of some twenty radionuclides of hall-lives between 30 minutes and 30 years and of activities such that the combustion of 1 g of target would release about 30 mCi of fission products of medium and short half-life (over 1 hour), This figure shows the magnitude of a contamination accident on a unit of this type, which is why the present report describes the systems to be employed in order on the one hand to detect the appearance of contamination as quickly as possible, and on the other hand to channel and retain this contamination so as to avoid a personnel contamination accident and/or the spread of contamination towards the outside [fr

Analytic flux formulas and tables of shielding functions

International Nuclear Information System (INIS)

Wallace, O.J.

1981-06-01

Hand calculations of radiation flux and dose rates are often useful in evaluating radiation shielding and in determining the scope of a problem. The flux formulas appropriate to such calculations are almost always based on the point kernel and allow for at most the consideration of laminar slab shields. These formulas often require access to tables of values of integral functions for effective use. Flux formulas and function tables appropriate to calculations involving homogeneous source regions with the shapes of lines, disks, slabs, truncated cones, cylinders, and spheres are presented. Slab shields may be included in most of these calculations, and the effect of a cylindrical shield surrounding a cylindrical source may be estimated. Detector points may be located axially, laterally, or interior to a cylindrical source. Line sources may be tilted with respect to a slab shield. All function tables are given for a wide range of arguments

International handbook of evaluated criticality safety benchmark experiments

International Nuclear Information System (INIS)

2010-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span over 55,000 pages and contain 516 evaluations with benchmark specifications for 4,405 critical, near critical, or subcritical configurations, 24 criticality alarm placement / shielding configurations with multiple dose points for each, and 200 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these evaluations; however, benchmark specifications are not derived for such experiments (in some cases models are provided in an appendix). Approximately 770 experimental configurations are categorized as unacceptable for use as criticality safety benchmark experiments. Additional evaluations are in progress and will be

Shielding plugs

International Nuclear Information System (INIS)

Makishima, Kenji.

1986-01-01

Purpose: In shielding plugs of an LMFBR type reactor, to restrain natural convection of heat in an annular space between a thermal shield layer and a shield shell, to prevent the lowering of heat-insulation performance, and to alleviate a thermal stress in a reactor container and the shield shell. Constitution: A ring-like leaf spring split in the direction of height is disposed in an annular space between a thermal shield layer and a shield shell. In consequence, the space is partitioned in the direction of height and, therefore, if axial temperature conditions and space width are the same and the space is low, the natural convection is hard to occur. Thus the rise of upper surface temperature of the shielding plugs can prevent the lowering of the heat insulation performance which will result in the increment of shielding plug cooling capacity, thereby improving reliability. In the meantime, since there is mounted an earthquake-resisting support, the thermal shield layer will move for a slight gap in case of an earthquake, being supported by the earthquake-resisting support, and the movement of the thermal shield layer is restricted, thereby maintaining integrity without increasing the stroke of the ring-like spring. (Kawakami, Y.)

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

Experimental shielding evaluation of the radiation protection provided by the structurally significant components of residential structures.

Science.gov (United States)

Dickson, E D; Hamby, D M

2014-03-01

The human health and environmental effects following a postulated accidental release of radioactive material to the environment have been a public and regulatory concern since the early development of nuclear technology. These postulated releases have been researched extensively to better understand the potential risks for accident mitigation and emergency planning purposes. The objective of this investigation is to provide an updated technical basis for contemporary building shielding factors for the US housing stock. Building shielding factors quantify the protection from ionising radiation provided by a certain building type. Much of the current data used to determine the quality of shielding around nuclear facilities and urban environments is based on simplistic point-kernel calculations for 1950s era suburbia and is no longer applicable to the densely populated urban environments realised today. To analyse a building's radiation shielding properties, the ideal approach would be to subject a variety of building types to various radioactive sources and measure the radiation levels in and around the building. While this is not entirely practicable, this research analyses the shielding effectiveness of ten structurally significant US housing-stock models (walls and roofs) important for shielding against ionising radiation. The experimental data are used to benchmark computational models to calculate the shielding effectiveness of various building configurations under investigation from two types of realistic environmental source terms. Various combinations of these ten shielding models can be used to develop full-scale computational housing-unit models for building shielding factor calculations representing 69.6 million housing units (61.3%) in the United States. Results produced in this investigation provide a comparison between theory and experiment behind building shielding factor methodology.

Experimental shielding evaluation of the radiation protection provided by the structurally significant components of residential structures

International Nuclear Information System (INIS)

Dickson, E D; Hamby, D M

2014-01-01

The human health and environmental effects following a postulated accidental release of radioactive material to the environment have been a public and regulatory concern since the early development of nuclear technology. These postulated releases have been researched extensively to better understand the potential risks for accident mitigation and emergency planning purposes. The objective of this investigation is to provide an updated technical basis for contemporary building shielding factors for the US housing stock. Building shielding factors quantify the protection from ionising radiation provided by a certain building type. Much of the current data used to determine the quality of shielding around nuclear facilities and urban environments is based on simplistic point-kernel calculations for 1950s era suburbia and is no longer applicable to the densely populated urban environments realised today. To analyse a building’s radiation shielding properties, the ideal approach would be to subject a variety of building types to various radioactive sources and measure the radiation levels in and around the building. While this is not entirely practicable, this research analyses the shielding effectiveness of ten structurally significant US housing-stock models (walls and roofs) important for shielding against ionising radiation. The experimental data are used to benchmark computational models to calculate the shielding effectiveness of various building configurations under investigation from two types of realistic environmental source terms. Various combinations of these ten shielding models can be used to develop full-scale computational housing-unit models for building shielding factor calculations representing 69.6 million housing units (61.3%) in the United States. Results produced in this investigation provide a comparison between theory and experiment behind building shielding factor methodology. (paper)

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

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

International Criticality Safety Benchmark Evaluation Project (ICSBEP) - ICSBEP 2015 Handbook

International Nuclear Information System (INIS)

Bess, John D.

2015-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy (DOE). The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirements and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross-section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span approximately 69000 pages and contain 567 evaluations with benchmark specifications for 4874 critical, near-critical or subcritical configurations, 31 criticality alarm placement/shielding configurations with multiple dose points for each, and 207 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the handbook are benchmark specifications for neutron activation foil and thermoluminescent dosimeter measurements performed at the SILENE critical assembly in Valduc, France as part of a joint venture in 2010 between the US DOE and the French Alternative Energies and Atomic Energy Commission (CEA). A photograph of this experiment is shown on the front cover. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these

Analysis of the NEACRP PWR rod ejection benchmark problems with DIF3D-K

International Nuclear Information System (INIS)

Kim, M.H.

1994-01-01

Analyses of the NEACRP PWR rod ejection transient benchmark problems with the DIF3D-K nodal kinetics code are presented. The DIF3D-K results are shown to be in generally good agreement with results obtained using other codes, in particular reference results previously generated with the PANTHER code. The sensitivity of the transient results to the DIF3D-K input parameters (such as time step size, radial and axial node sizes, and the mesh structure employed for fuel pin heat conduction calculation) are evaluated and discussed. In addition, the potential in reducing computational effort by application of the improved quasistatic scheme (IQS) to these rod ejection transients, which involve very significant flux shape changes and thermal-hydraulic feedback is evaluated

A comparative study of upwind and MacCormack schemes for CAA benchmark problems

Science.gov (United States)

Viswanathan, K.; Sankar, L. N.

1995-01-01

In this study, upwind schemes and MacCormack schemes are evaluated as to their suitability for aeroacoustic applications. The governing equations are cast in a curvilinear coordinate system and discretized using finite volume concepts. A flux splitting procedure is used for the upwind schemes, where the signals crossing the cell faces are grouped into two categories: signals that bring information from outside into the cell, and signals that leave the cell. These signals may be computed in several ways, with the desired spatial and temporal accuracy achieved by choosing appropriate interpolating polynomials. The classical MacCormack schemes employed here are fourth order accurate in time and space. Results for categories 1, 4, and 6 of the workshop's benchmark problems are presented. Comparisons are also made with the exact solutions, where available. The main conclusions of this study are finally presented.

SHREDI, Neutron Flux and Neutron Activation in 2-D Shields by Removal Diffusion

International Nuclear Information System (INIS)

Daneri, A.; Toselli, G.

1976-01-01

1 - Nature of physical problem solved: SHREDI is a removal - diffusion neutron shielding code. The program computes neutron fluxes and activations in bidimensional sections (x,y or r,z) of the shield. It is also possible to consider shielding points with the same y or z coordinate (mono-dimensional problems). 2 - Method of solution: The integrals which define the removal fluxes are computed in some shield points by means of a particular algorithm based on the Simpson's and trapezoidal rules. For the diffusion calculation the finite difference method is used. The removal sources are interpolated in all diffusion points by Chebyshev polynomials. 3 - Restrictions on the complexity of the problem: Maxima: number of removal energy groups NGR = 40; number of diffusion energy groups NGD = 40; number of the reactor core and shield materials NCMP = 50; number of core mesh points in r (or x) direction for integral calculation = 75; number of core mesh points in z (or y) direction for integral calculation = 75; number of core mesh points in theta (or z) direction for integral calculation = 75; number of shield mesh points for the neutron flux calculation in r (or x) direction NPX = 200; number of shield mesh points for the neutron flux calculation in z (or y) direction NPY = 200; n.b. (NPX * NPY) le 12000

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)

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)

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

International Nuclear Information System (INIS)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

Results of the first benchmark physics tests of a metallic-fueled, demonstration-size, liquid metal reactor are reported. A simple, two-zone, cylindrical conventional assembly was built with three distinctly different compositions to represent the stages of the Integral Fast Reactor fuel cycle. Experiments included criticality, control, power distribution, reaction rate ratios, reactivity coefficients, shielding, kinetics and spectrum. Analysis was done with 3-D nodal diffusion calculations and ENDFIB-V.2 cross sections. Predictions of the ZPPR-15 reactor physics parameters agreed sufficiently well with the measured values to justify confidence in design analyses for metallic-fueled LMRs

Case mix classification and a benchmark set for surgery scheduling

NARCIS (Netherlands)

Leeftink, Gréanne; Hans, Erwin W.

Numerous benchmark sets exist for combinatorial optimization problems. However, in healthcare scheduling, only a few benchmark sets are known, mainly focused on nurse rostering. One of the most studied topics in the healthcare scheduling literature is surgery scheduling, for which there is no widely

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.

SHREDI a removal diffusion shielding code for x-y and r-z geometries

International Nuclear Information System (INIS)

Daneri, A.; Toselli, G.

1974-01-01

The SHREDI, a removal diffusion neutron shielding code written in FORTRAN for IBM 370/165, is presented. The code computes neutron fluxes or adjoint fluxes and activations in bidimensional sections of the shield. It is also possible to consider shielding points with the same coordinate (y or z) (monodimensional problems)

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

MCNP: Photon benchmark problems

International Nuclear Information System (INIS)

Whalen, D.J.; Hollowell, D.E.; Hendricks, J.S.

1991-09-01

The recent widespread, markedly increased use of radiation transport codes has produced greater user and institutional demand for assurance that such codes give correct results. Responding to these pressing requirements for code validation, the general purpose Monte Carlo transport code MCNP has been tested on six different photon problem families. MCNP was used to simulate these six sets numerically. Results for each were compared to the set's analytical or experimental data. MCNP successfully predicted the analytical or experimental results of all six families within the statistical uncertainty inherent in the Monte Carlo method. From this we conclude that MCNP can accurately model a broad spectrum of photon transport problems. 8 refs., 30 figs., 5 tabs

Future directions in shielding methods and analysis

International Nuclear Information System (INIS)

Goldstein, H.

1987-01-01

Over the nearly half century history of shielding against reactor radiation, there has been a see-saw battle between theory and measurement. During that period the capability and accuracy of calculational methods have been enormously improved. The microscopic cross sections needed as input to the theoretical computations are now also known to adequate accuracy (with certain exceptions). Nonetheless, there remain substantial classes of shielding problems not yet accessible to satisfactory computational methods, particularly where three-dimensional geometries are involved. This paper discusses promising avenues to approach such problems, especially in the light of recent and expected advances in supercomputers. In particular, it seems that Monte Carlo methods should be much more advantageous in the new computer environment than they have been in the past

ICSBEP-2007, International Criticality Safety Benchmark Experiment Handbook

International Nuclear Information System (INIS)

Blair Briggs, J.

2007-01-01

1 - Description: The Critically Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United Sates Department of Energy. The project quickly became an international effort as scientist from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization of Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA). This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material. The example calculations presented do not constitute a validation of the codes or cross section data. The work of the ICSBEP is documented as an International Handbook of Evaluated Criticality Safety Benchmark Experiments. Currently, the handbook spans over 42,000 pages and contains 464 evaluations representing 4,092 critical, near-critical, or subcritical configurations and 21 criticality alarm placement/shielding configurations with multiple dose points for each and 46 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. The handbook is intended for use by criticality safety analysts to perform necessary validations of their calculational techniques and is expected to be a valuable tool for decades to come. The ICSBEP Handbook is available on DVD. You may request a DVD by completing the DVD Request Form on the internet. Access to the Handbook on the Internet requires a password. You may request a password by completing the Password Request Form. The Web address is: http://icsbep.inel.gov/handbook.shtml 2 - Method of solution: Experiments that are found

Demonstration study on shielding safety analysis code. 7

Energy Technology Data Exchange (ETDEWEB)

Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

2000-03-01

Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) To improve the detection sensitivity of pulse neutron measurement, two neutron detectors and some electronic circuits are added to the system constructed last year. (2) To estimate the neutron dose at the distant point from the facility instead of the commercialized rem-counter, a {sup 3}He detector with paraffin moderator is equipped to the system. (3) Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility was measured in the distance up to 300 m. The results show that the time structure of pulsed neutrons almost disappears at the further points than 150 m. (4) In the distance from 90 m to 300 m ordinal total counting method without gate pulse are applied to detect the neutrons. (5) The experimental results of space dependency up to 300 m is fitted fairly well by the Gui's response function. (author)

Phase 2 testing of ENDF/B-VI shielding data

International Nuclear Information System (INIS)

Ingersoll, D.T.; Wright, R.Q.; Slater, C.O.

1992-01-01

Version 6 of the US Evaluated Nuclear Data File (ENDF/B-VI) was released in early 1990 and is currently undergoing phase 2 testing. In Phase 2 testing, the evaluated data are approximately processed and used in an integral manner to predict the solution of previously specified benchmark experiments. Results are presented for the initial testing of several light elements and structural materials which are important for shielding applications. These initial tests indicate that the relatively subtle changes made to the iron data and the major modernization of the boron-11 data in Version 6 both represent significant and positive advancements in the quality of the evaluated data files

Present and future problems of radiation shielding for maritime transport of nuclear spent fuels

International Nuclear Information System (INIS)

Ueki, K.; Nariyama, N.; Ohashi, A.

2000-01-01

The transport of spent fuels with casks began in September 1999 by the exclusive spent fuel transport vessel the 'Rokuei Maru'. The casks have been transported to the reprocessing plant at Rokkasho-village in Aomori Prefecture. The 'Rokuei Maru' is approximately 100 m-length, 16.5 m-width and 3,000 gross-tons. The 20 NFT casks can be loaded into 5 holds. At the present time, the NFT casks can carry spent fuels of up to 44,000 MWD/MTU. Serpentine concrete is employed as a neutron shields in the hatch covers, the bulkheads, and the house front of the accommodations except the wheelhouse. Polyethylene covers the side walls in each hold. The neutron shielding ability of serpentine concrete and polyethylene was investigated by a shielding experiment using a 252 Cf-neutron source. The shielding experiment was analyzed with the Monte Carlo code MCNP 4B. In the near future, on-board experiment will be carried out to measure the dose-equivalent rate distributions in the 'Rokuei Maru' and the measured data and the Monte Carlo analysis of it will establish the radiation safety of the ship. (author)

Uranium-lead shielding for nuclear material transportation systems

International Nuclear Information System (INIS)

Lusk, E.C.; Miller, N.E.; Basham, S.J. Jr.

1978-01-01

The basis for the selection of shielding materials for spent fuel shipping containers is described with comments concerning the favorable and unfavorable aspects of steel, lead, and depleted uranium. A concept for a new type of material made of depleted uranium and lead is described which capitalizes on the best cask shielding characteristics of both materials. This cask shielding is made by filling the shielding cavity with pieces of depleted uranium and then backfilling the interstitial voids with lead. The lead would be bonded to the uranium and also to the cask shells if desired. Shielding density approaching 80 percent of that of solid uranium could be achieved, while a density of 65 percent is readily obtainable. This material should overcome the problems of the effect of lead melting in the fire accident, high thermal gradients at uranium-stainless steel interfaces and at a major reduction in cost over that of a solid uranium shielded cask. A development program is described to obtain information on the properties of the composite material to aid in design analysis and licensing and to define the fabrication techniques

Detector Background Reduction by Passive and Active Shielding

International Nuclear Information System (INIS)

Bikit, I.; Bikit, K.; Forkapic, S.; Mrda, D.; Nikolov, J.; Slivka, J.; Todorovic, N.

2013-01-01

The operational problems of the gamma ray spectrometer shielded passively with 12 cm of lead and actively by five 0.5 m × 0.5 m × 0.05 m plastic veto shields are described. The active shielding effect from both environmental gamma ray, cosmic muons and neutrons was investigated. For anticoincidence gating wide range of scintillator pulses, corresponding to the energy range of 150 keV-75 MeV, were used. With the optimal set up the integral background, for the energy region of 50 - 3000 keV, of 0.31 c/s was achieved. The detector mass related background was 0.345 c/(kg s). The 511 keV annihilation line was reduced by the factor of 7 by the anticoincidence gate. It is shown that the plastic shields increase the neutron capture gamma line intensities due to neutron termalization.(author)

Calculation of the BREN house shielding experiments

International Nuclear Information System (INIS)

Woolson, William A.; Gritzner, Michael L.

1987-01-01

The BREN house transmission experiments provide an excellent set of measurements to validate the calculational procedures that will be used to derive house shielding estimates for the revised dosimetry of the survivors of the Hiroshima and Nagasaki A-bombs. The BREN experiments were performed in realistic full scale models of Japanese residences. Although the radiation spectra and relative intensities of neutrons and gamma rays incident on the houses from the HPRR and the 60 Co source are not appropriate for direct application to the A-bomb survivors, they cover the full energy range of importance. The codes and calculations required to compare with BREN experiments are the same as those needed for the A-bomb dosimetry. They consist of a two-dimensional discrete-ordinates calculation of the free field coupled to an adjoint Monte Carlo calculation in detailed house geometry. The agreement obtained between calculations and the experiments is excellent for neutrons and 60 Co gamma rays. Every house transmission calculation spanning simple to complex configurations and detector locations for the 60 Co and HPRR was within an acceptable margin of error. The gamma-ray TF calculations for the reactor source did not agree well with the experiments. Analysis of this discrepancy, however, strongly indicates that the problem probably does not reside in the calculational procedure but in the measurements themselves. In conclusion, it is believed that the excellent agreement of our calculations with the BREN experiments validates the calculational procedure which is planed to be applied o estimating the house shielding for survivors of the Hiroshima and Nagasaki A-bombs. Certainly, the calculations for Hiroshima and Nagasaki will involve modifications to the code used for the computations reported here, but to the extent that these modifications involve increased calculational complexity to treat more realistic materials and configurations, the benchmark established by these

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)

NEACRP comparison of codes for the radiation protection assessment of transportation packages. Solutions to problems 1 - 4

International Nuclear Information System (INIS)

Avery, A.F.; Locke, H.F.

1992-03-01

In 1985 the Reactor Physics Committee of the Nuclear Energy Agency initiated an intercomparison of codes for the calculation of the performance of shielding for the transportation of spent reactor fuel. The results of the application of a range of codes to the prediction of the dose-rates in the four theoretical benchmarks set to examine the attenuation of radiation through a variety of cask geometries are presented in this report. The contributions from neutrons, fission product gamma-rays and secondary gamma-rays are tabulated separately, and grouped according to the type of method of calculation employed. A brief discussion is included for each set of results, and overall comparisons of the methods, codes, and nuclear data are made. A number of conclusions are drawn on the advantages and disadvantages of the various methods of calculation, based upon the results of their application to these four benchmark problems

Concepts for benchmark problem development for fracture mechanics application in safety evaluation of nuclear piping in subcreep service

International Nuclear Information System (INIS)

Reich, M.; Esztergar, E.P.; Erdogan, F.; Gray, T.G.F.; Spence, J.

1979-01-01

This report provides basic concepts and a review of the problem areas associated with the development of analytical and experimental programs for a systematic evaluation and comparison of the currently available fracture mechanics theories. The basis for such an evaluation is conceived as a series of benchmark problems which are accurately specified examples of geometry, loading, and environmental conditions, characteristic of large diameter thin wall piping systems in nuclear service. Starting from the simplest test coupons for cracked plate specimens, the program is to be designed in such a way that the range of validity and relative merit of the competing assessment methods can be evaluated and the results applied to increasingly more complex test configurations and ultimately to real piping systems. (Auth.)

Results of LWR core transient benchmarks

International Nuclear Information System (INIS)

Finnemann, H.; Bauer, H.; Galati, A.; Martinelli, R.

1993-10-01

LWR core transient (LWRCT) benchmarks, based on well defined problems with a complete set of input data, are used to assess the discrepancies between three-dimensional space-time kinetics codes in transient calculations. The PWR problem chosen is the ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal - hydraulic models of the codes. The 63 sets of submitted solutions are analyzed by comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations. The BWR problems considered are reactivity excursions caused by cold water injection and pressurization events. In the present paper, only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution the evaluation of the 8 sets of BWR contributions relies on a synthetic comparative discussion. The results of this first phase of LWRCT benchmark calculations are quite satisfactory, though there remain some unresolved issues. It is therefore concluded that even more challenging problems can be successfully tackled in a suggested second test phase. (authors). 46 figs., 21 tabs., 3 refs

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.

Application of the decoupling scheme on complex neutron-gamma shielding problems

Energy Technology Data Exchange (ETDEWEB)

Feher, S. [Institute of Nuclear Technology, Technical University of Budapest, Budapest (Hungary); Leege, P.F.A. de; Hoogenboom, J.E.; Kloosterman, J.L. [Interfaculty Reactor Institute, Delft University of Technology, Delft (Netherlands)

2000-03-01

Coupled neutron-gamma shielding calculations using S{sub n} transport theory can be time consuming, especially for two- and three-dimensional geometries. In general, the CPU time of these calculations increases stronger than linear with increasing number of neutron and gamma energy groups, and depends on the order of Legendre expansion and number of S{sub n} directions used. This fact induced the idea of the decoupling method, which seems applicable to accelerate coupled neutron-gamma shielding calculations. The data included in a combined neutron-gamma library can be readily separated into a library containing neutron data only and another library containing gamma data only. Separate calculations for neutrons and gammas are performed on complex geometries using a different Legendre order expansion for neutrons and gammas. CPU savings of 60 to 85% can be achieved for the two-dimensional DORT and three-dimensional TORT calculations respectively. (author)

Two-fluid dusty shocks: simple benchmarking problems and applications to protoplanetary discs

Science.gov (United States)

Lehmann, Andrew; Wardle, Mark

2018-05-01

The key role that dust plays in the interstellar medium has motivated the development of numerical codes designed to study the coupled evolution of dust and gas in systems such as turbulent molecular clouds and protoplanetary discs. Drift between dust and gas has proven to be important as well as numerically challenging. We provide simple benchmarking problems for dusty gas codes by numerically solving the two-fluid dust-gas equations for steady, plane-parallel shock waves. The two distinct shock solutions to these equations allow a numerical code to test different forms of drag between the two fluids, the strength of that drag and the dust to gas ratio. We also provide an astrophysical application of J-type dust-gas shocks to studying the structure of accretion shocks on to protoplanetary discs. We find that two-fluid effects are most important for grains larger than 1 μm, and that the peak dust temperature within an accretion shock provides a signature of the dust-to-gas ratio of the infalling material.

A history of radiation shielding of x-ray therapy rooms

International Nuclear Information System (INIS)

McGinley, P.H.; Miner, M.S.

1996-01-01

In this report the history of shielding for radiation treatment rooms is traced from the time of the discovery of x rays to the present. During the early part of the twentieth century the hazards from ionizing radiation were recognized and the use of lead and other materials became common place for shielding against x rays. Techniques for the calculation of the shield thickness needed for x ray protection were developed in the 1920's, and shielding materials were characterized in terms of the half value layer or simple exponential factors. At the same time, better knowledge of the interaction between radiation and matter was acquired. With the development of high energy medical accelerators after 1940, new and more complex shielding problems had to be addressed. Recently, shielding requirements have become more stringent as standards for exposure of personnel and the general public have been reduced. The art of shielding of radiation treatment facilities is still being developed, and the need for a revision of the reports on shielding of medical accelerators from the National Council on Radiation Protection and Measurements is emphasized in this article. (author). 61 Refs., 3 Tabs

SCORE-4, 2-D Removal Diffusion in X-Y or R-Z Geometry for Rectangular Shields

International Nuclear Information System (INIS)

Richardson, B.L.

1974-01-01

1 - Nature of physical problem solved: The neutron flux is calculated for a shield made up of rectangular regions. The geometry is either x-y or r-z. 2 - Method of solution: Removal fluxes and sources throughout the shield regions are calculated from a given reactor core power distribution using a point kernel method. The diffusion neutron fluxes are obtained from the removal source distribution using an iterative Method of solution. 3 - Restrictions on the complexity of the problem: The amount of fast core required for the program depends on the size of shield being calculated. For example, a 100 by 100 mesh shielding calculation would require approximately 300 k bytes. Larger problems could be solved by increasing the fast storage requirements

Benchmarking of the computer code and the thirty foot side drop analysis for the Shippingport (RPV/NST package)

International Nuclear Information System (INIS)

Bumpus, S.E.; Gerhard, M.A.; Hovingh, J.; Trummer, D.J.; Witte, M.C.

1989-01-01

This paper presents the benchmarking of a finite element computer code and the subsequent results from the code simulating the 30 foot side drop impact of the RPV/NST transport package from the decommissioned Shippingport Nuclear Power Station. The activated reactor pressure vessel (RPV), thermal shield, and other reactor external components were encased in concrete contained by the neutron shield tank (NST) and a lifting skirt. The Shippingport RPV/NST package, a Type B Category II package, weighs approximately 900 tons and has 17.5 ft diameter and 40.7 ft. length. For transport of the activated components from Shippingport to the burial site, the Safety Analysis Report for Packaging (SARP) demonstrated that the package can withstand the hypothetical accidents of DOE Order 5480.3 including 10 CFR 71. Mathematical simulations of these accidents can substitute for actual tests if the simulated results satisfy the acceptance criteria. Any such mathematical simulation, including the modeling of the materials, must be benchmarked to experiments that duplicate the loading conditions of the tests. Additional confidence in the simulations is justified if the test specimens are configured similar to the package

Deep-penetration calculations in concrete and iron for shielding of proton therapy accelerators

International Nuclear Information System (INIS)

Sheu, Rong-Jiun; Chen, Yen-Fu; Lin, Uei-Tyng; Jiang, Shiang-Huei

2012-01-01

Proton accelerators in the energy range of approximately 200 MeV have become increasingly popular for cancer treatment in recent years. These proton therapy facilities usually involve bulky concrete or iron in their shielding design or accelerator structure. Simple shielding data, such as source terms or attenuation lengths for various proton energies and materials are useful in designing accelerator shielding. Understanding the appropriateness or uncertainties associated with these data, which are largely generated from Monte Carlo simulations, is critical to the quality of a shielding design. This study demonstrated and investigated the problems of deep-penetration calculations on the estimation of shielding parameters through an extensive comparison between the FLUKA and MCNPX calculations for shielding against a 200-MeV proton beam hitting an iron target. Simulations of double-differential neutron production from proton bombardment were validated by comparison with experimental data. For the concrete shielding, the FLUKA calculated depth–dose distributions were consistent with the MCNPX results, except for some discrepancies in backward directions. However, for the iron shielding, if FLUKA is used inappropriately then overestimation of neutron attenuation can be expected as shown by this work because of the multigroup treatment for low-energy neutrons in FLUKA. Two neutron energy group structures, three degrees of self-shielding correction, and two iron compositions were considered in this study. Significant variation of the resulting attenuation lengths indicated the importance of problem-dependent multigroup cross sections and proper modeling of iron composition in deep-penetration calculations.

Electromagnetic shielding

International Nuclear Information System (INIS)

Tzeng, Wen-Shian V.

1991-01-01

Electromagnetic interference (EMI) shielding materials are well known in the art in forms such as gaskets, caulking compounds, adhesives, coatings and the like for a variety of EMI shielding purposes. In the past, where high shielding performance is necessary, EMI shielding has tended to use silver particles or silver coated copper particles dispersed in a resin binder. More recently, aluminum core silver coated particles have been used to reduce costs while maintaining good electrical and physical properties. (author). 8 figs

Benchmarking with high-order nodal diffusion methods

International Nuclear Information System (INIS)

Tomasevic, D.; Larsen, E.W.

1993-01-01

Significant progress in the solution of multidimensional neutron diffusion problems was made in the late 1970s with the introduction of nodal methods. Modern nodal reactor analysis codes provide significant improvements in both accuracy and computing speed over earlier codes based on fine-mesh finite difference methods. In the past, the performance of advanced nodal methods was determined by comparisons with fine-mesh finite difference codes. More recently, the excellent spatial convergence of nodal methods has permitted their use in establishing reference solutions for some important bench-mark problems. The recent development of the self-consistent high-order nodal diffusion method and its subsequent variational formulation has permitted the calculation of reference solutions with one node per assembly mesh size. In this paper, we compare results for four selected benchmark problems to those obtained by high-order response matrix methods and by two well-known state-of-the-art nodal methods (the open-quotes analyticalclose quotes and open-quotes nodal expansionclose quotes methods)

Comparative study between NCRP-49 and NCRP-147 methodologies for shielding calculus to fluoroscopy rooms

International Nuclear Information System (INIS)

Ferreira, Christiano Eduardo Martins

2011-01-01

The walls of a fluoroscopy room must be shielded to prevent unnecessary exposures to technicians and public individuals. Thus this dissertation aims to describe the methodologies contained in two documents which are references for the calculation of shielding those rooms. They are the National Council on Radiation Protection and Measurements Report No. 49 (NCRP Report No. 49) and No. 147 (NCRP Report No. 147), the latter being more recent publication. And based on such description was made a comparative study between the two methodologies, using for this, as a benchmark, spreadsheets computer program developed by Wolfram Mathematica 6. With that we could reach the final thickness of the barriers to a Standard Plan for a fluoroscopy room (provided by Siemens) and noted that the NCRP-49 presents a methodology with results more conservative. (author)

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

International Nuclear Information System (INIS)

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

1997-03-01

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

Neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) MCNP ''Benchmark CAD Model'' with the ATTILA discrete ordinance code

International Nuclear Information System (INIS)

Youssef, M.Z.; Feder, R.; Davis, I.

2007-01-01

The ITER IT has adopted the newly developed FEM, 3-D, and CAD-based Discrete Ordinates code, ATTILA for the neutronics studies contingent on its success in predicting key neutronics parameters and nuclear field according to the stringent QA requirements set forth by the Management and Quality Program (MQP). ATTILA has the advantage of providing a full flux and response functions mapping everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. The ITER neutronics community had agreed to use a standard CAD model of ITER (40 degree sector, denoted ''Benchmark CAD Model'') to compare results for several responses selected for calculation benchmarking purposes to test the efficiency and accuracy of the CAD-MCNP approach developed by each party. Since ATTILA seems to lend itself as a powerful design tool with minimal turnaround time, it was decided to benchmark this model with ATTILA as well and compare the results to those obtained with the CAD MCNP calculations. In this paper we report such comparison for five responses, namely: (1) Neutron wall load on the surface of the 18 shield blanket module (SBM), (2) Neutron flux and nuclear heating rate in the divertor cassette, (3) nuclear heating rate in the winding pack of the inner leg of the TF coil, (4) Radial flux profile across dummy port plug and shield plug placed in the equatorial port, and (5) Flux at seven point locations situated behind the equatorial port plug. (orig.)

About the Scythian Shields

Directory of Open Access Journals (Sweden)

About the Scythian Shields

2017-10-01

Full Text Available Shields played major role in the armament system of the Scythians. Made from organic materials, they are poorly traced on the materials of archaeological excavations. Besides, scaly surface of shields was often perceived in practice as the remnants of the scaly armor. E. V. Chernenko was able to discern the difference between shields’ scaly plates and armor scales. The top edge of the scales was bent inwards, and shield plates had a wire fixation. These observations let significantly increase the number of shields, found in the burial complexes of the Scythians. The comparison of archaeological materials and the images of Scythian warriors allow distinguishing the main forms of Scythian shields. All shields are divided into fencing shields and cover shields. The fencing shields include round wooden shields, reinforced with bronze sheet, and round moon-shaped shields with a notch at the top, with a metal scaly surface. They came to the Scythians under the Greek influence and are known in the monuments of the 4th century BC. Oval shields with scaly surface (back cover shields were used by the Scythian cavalry. They protected the rider in case of frontal attack, and moved back in case of maneuver or closein fighting. Scythian battle tactics were based on rapid approaching the enemy and throwing spears and further rapid withdrawal. Spears stuck in the shields of enemies, forcing them to drop the shields, uncover, and in this stage of the battle the archers attacked the disorganized ranks of the enemy. That was followed by the stage of close fight. Oval form of a wooden shield with leather covering was used by the Scythian infantry and spearmen. Rectangular shields, including wooden shields and the shields pleached from rods, represented a special category. The top of such shield was made of wood, and a pleached pad on leather basis was attached to it. This shield could be a reliable protection from arrows, but it could not protect against javelins

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.

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

Shielding methods development in the United States

International Nuclear Information System (INIS)

Mynatt, F.R.

1977-01-01

A generalized shielding methodology has been developed in the U.S.A. that is adaptable to the shielding analyses of all reactor types. Thus far used primarily for liquid-metal fast breeder reactors, the methodology includes several component activities: (1) developing methods for calculating radiation transport through reactor-shield systems; (2) processing cross-section libraries; (3) performing design calculations for specific systems; (4) performing and analyzing pertinent integral experiments; (5) performing sensitivity studies on both the design calculations and the experimental analyses; and, finally, (6) calculating shield design parameters and their uncertainties. The criteria for the methodology are a 5 to 10 percent accuracy for responses at locations near the core and a factor of 2 accuracy for responses at distant locations. The methodology has been successfully adapted to most in-vessel and ex-vessel problems encountered in the shield analyses of the Fast Flux Test Facility and the Fast Flux Test Facility and the Clinch River Breeder Reactor; however, improved techniques are needed for calculating regions in which radiation streaming is dominant. Areas of the methodology in which significant progress has recently been made are those involving the development of cross-section libraries, sensitivity analysis methods, and transport codes

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

Benchmarking the Netherlands. Benchmarking for growth

International Nuclear Information System (INIS)

2003-01-01

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

Benchmarking the Netherlands. Benchmarking for growth

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-01-01

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

Accelerator shielding experts meet at CERN

CERN Multimedia

CERN Bulletin

2010-01-01

Fifteen years after its first CERN edition, the Shielding Aspects of Accelerator, Targets and Irradiation Facility (SATIF) conference was held again here from 2-4 June. Now at its 10th edition, SATIF10 brought together experts from all over the world to discuss issues related to the shielding techniques. They set out the scene for an improved collaboration and discussed novel shielding solutions.   This was the most attended meeting of the series with more than 65 participants from 34 institutions and 14 countries. “We welcomed experts from many different laboratories around the world. We come from different contexts but we face similar problems. In this year’s session, among other things, we discussed ways for improving the effectiveness of calculations versus real data, as well as experimental solutions to investigate the damage that radiation produces on various materials and the electronics”, says Marco Silari, Chair of the conference and member of the DGS/RP gro...

Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

International Nuclear Information System (INIS)

Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

2015-01-01

As the system characteristics, the target in the spectrometer emits approximately 1012 neutrons/s. To efficiently shield the neutron, the shielding door designs are proposed for the LSDS system through a comparison of the direct shield and maze designs. Hence, to guarantee the radiation safety for the facility, the door design is a compulsory course of the development of the LSDS system. To improve the shielding rates, 250x250 covering structure was added as a subsidiary around the spectrometer. In this study, the evaluations of the suggested shielding designs were conducted using MCNP code. The suggested door design and covering structures can shield the neutron efficiently, thus all evaluations of all conditions are satisfied within the public dose limits. From the Monte Carlo code simulation, Resin(Indoor type) and Tungsten(Outdoor type) were selected as the shielding door materials. From a comparative evaluation of the door thickness, In and Out door thickness was selected 50 cm

Protocol of source shielding maintenance in a level measurement systems

International Nuclear Information System (INIS)

Gonzales, E.; Figueroa, J.

1996-01-01

Maintenance labor of the source shielding and locking system is not performed in many Venezuelan enterprises that employ radioactive level gauge in large container. The lack of maintenance and the ambient long lasting action have produced impairment of many devices and their given parts rise to economical and radiological protection problems. In order to help to solve the mentioned problems, principally to reduce the unjustified dose to workers, the IVIC Health Physics Service worked out a protocol to perform, in a safety way, the maintenance of source shielding and its locking system. This protocol is presented in this paper. (authors)

Shieldings for X-ray radiotherapy facilities calculated by computer

International Nuclear Information System (INIS)

Pedrosa, Paulo S.; Farias, Marcos S.; Gavazza, Sergio

2005-01-01

This work presents a methodology for calculation of X-ray shielding in facilities of radiotherapy with help of computer. Even today, in Brazil, the calculation of shielding for X-ray radiotherapy is done based on NCRP-49 recommendation establishing a methodology for calculating required to the elaboration of a project of shielding. With regard to high energies, where is necessary the construction of a labyrinth, the NCRP-49 is not very clear, so that in this field, studies were made resulting in an article that proposes a solution to the problem. It was developed a friendly program in Delphi programming language that, through the manual data entry of a basic design of architecture and some parameters, interprets the geometry and calculates the shields of the walls, ceiling and floor of on X-ray radiation therapy facility. As the final product, this program provides a graphical screen on the computer with all the input data and the calculation of shieldings and the calculation memory. The program can be applied in practical implementation of shielding projects for radiotherapy facilities and can be used in a didactic way compared to NCRP-49.

Optimization of the National Ignition Facility primary shield design

International Nuclear Information System (INIS)

Annese, C.E.; Watkins, E.F.; Greenspan, E.; Miller, W.F.

1993-10-01

Minimum cost design concepts of the primary shield for the National Ignition laser fusion experimental Facility (NIF) are searched with the help of the optimization code SWAN. The computational method developed for this search involves incorporating the time dependence of the delayed photon field within effective delayed photon production cross sections. This method enables one to address the time-dependent problem using relatively simple, time-independent transport calculations, thus significantly simplifying the design process. A novel approach was used for the identification of the optimal combination of constituents that will minimize the shield cost; it involves the generation, with SWAN, of effectiveness functions for replacing materials on an equal cost basis. The minimum cost shield design concept was found to consist of a mixture of polyethylene and low cost, low activation materials such as SiC, with boron added near the shield boundaries

Hydrogen Induced Cracking of Drip Shield

Energy Technology Data Exchange (ETDEWEB)

G. De

2003-02-24

One potential failure mechanism for titanium and its alloys under repository conditions is via the absorption of atomic hydrogen in the metal crystal lattice. The resulting decreased ductility and fracture toughness may lead to brittle mechanical fracture called hydrogen-induced cracking (HIC) or hydrogen embrittlement. For the current design of the engineered barrier without backfill, HIC may be a problem since the titanium drip shield can be galvanically coupled to rock bolts (or wire mesh), which may fall onto the drip shield, thereby creating conditions for hydrogen production by electrochemical reaction. The purpose of this scientific analysis and modeling activity is to evaluate whether the drip shield will fail by HIC or not under repository conditions within 10,000 years of emplacement. This Analysis and Model Report (AMR) addresses features, events, and processes related to hydrogen induced cracking of the drip shield. REV 00 of this AMR served as a feed to ''Waste Package Degradation Process Model Report'' and was developed in accordance with the activity section ''Hydrogen Induced Cracking of Drip Shield'' of the development plan entitled ''Analysis and Model Reports to Support Waste Package PMR'' (CRWMS M&O 1999a). This AMR, prepared according to ''Technical Work Plan for: Waste Package Materials Data Analyses and Modeling'' (BSC 2002), is to feed the License Application.

Development of shielding design analysis system

International Nuclear Information System (INIS)

Tada, Keiko; Shiraki, Takako

2001-03-01

The aim of this work is to develop insufficient auxiliary routines which manage input and output data and interface the main codes and to establish a shielding design analysis system on work stations (SUN, DEC). In shielding design analyses, one- and two- dimensional (1-D and 2-D) transport Sn codes are used mainly with some auxiliary codes which generate input data of Sn calculation and edit Sn calculation outputs. The main transport calculation codes can be obtained from the Code Center of RIST (Research Organization for Information Science and Technology). In this work, peripheral codes are developed to generate cross sections, produce Sn quadrature sets, edit calculation outputs or draw contour figures. In shielding calculations around a reactor, the boot-strapping technique is often employed to treat a large area extending from the core to the biological shield to improve the calculation accuracy. When a three-dimensional (3-D) calculation for a complex geometry with shielding defects, 2-D and 3-D coupling calculation is employed frequently. To use this coupling method conversion cods are prepared which read flux file from DORT and prepare an external boundary source file for the 2-D or the 3-D calculation codes. For further conveniences well used data such as the Sn quadrature sets, the dose rate conversion factors, the reaction cross section sets are stored as a data base and code manuals including sample inputs of typical problems are prepared which are comprehensible to beginners. (author)

A New Mechanism for Network Monitoring and Shielding in Wireless LAN

Directory of Open Access Journals (Sweden)

Jiujun Cheng

2014-01-01

Full Text Available Wireless LAN (WLAN technology is developing rapidly with the help of wireless communication technology and social demand. During the development of WLAN, the security is more and more important, and wireless monitoring and shielding are of prime importance for network security. In this paper, we have explored various security issues of IEEE 802.11 based wireless network and analyzed numerous problems in implementing the wireless monitoring and shielding system. We identify the challenges which monitoring and shielding system needs to be aware of, and then provide a feasible mechanism to avoid those challenges. We implemented an actual wireless LAN monitoring and shielding system on Maemo operating system to monitor wireless network data stream efficiently and solve the security problems of mobile users. More importantly, the system analyzes wireless network protocols efficiently and flexibly, reveals rich information of the IEEE 802.11 protocol such as traffic distribution and different IP connections, and graphically displays later. Moreover, the system running results show that the system has the capability to work stably, and accurately and analyze the wireless protocols efficiently.

Shielding assessment for the proposed HRIBF upgrade to the National ISOL Facility

International Nuclear Information System (INIS)

Slater, C.O.; Olsen, D.K.; Johnson, J.O.; Lillie, R.A.; Gabriel, T.A.

1997-04-01

An upgrade of the existing ORNL Holifield Radioactive Ion Beam Facility (HRIBF) to the National Radioactive Ion Beam Isotope Separator On Line (RIB ISOL) Facility is being proposed. Part of the upgrade involves increasing the source proton energy and current, resulting in more intense, higher energy radiation. Shielding requirements for the proposed upgrade to the HRIBF have been assessed with respect to weight, space, and dose-rate constraints. Shielding assessments were made for operating, shutdown, and accident conditions. The results indicate reasonable shielding solutions for the target room except for the marginal dose rate on the roof. Shielding requirements in the target room were greatly reduced by decisions to move the target to a more interior room and to direct the proton beam downward into the target. A slightly more difficult shielding problem arises for proton beam extraction losses from the cyclotron. Here, the assumed isotropic beam losses (hence, neutron emissions) mean higher roof dose rates than those over the target room unless substantial localized shielding is placed over the cyclotron. Shutdown dose rates were found to present no problems. While dose rates through the sides of the facility during accident conditions will probably satisfy the accident dose-rate constraints, dose rates above the roof will be well above the constraints unless a solution is devised to shield the locations where beam losses are likely to occur. Ground activation analysis was postponed for this study

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.

Shielding study of a fusion machine. Elaboration of a global shielding calculation scheme for the Tokamak tore Supra

International Nuclear Information System (INIS)

Diop, C.M'B.

1984-01-01

This thesis presents a global shielding calculation scheme for neutron and gamma rays arising from the Tokamak TORE SUPRA fusion device, in which a deuterium plasma is used. To study the shield parameters we have elabored a important chaining of neutron and gamma transport codes, TRIPOLI, ANISN, MERCURE 4, allowing to evaluate the radial and skyshine components of the dose rate behind the concrete shield. The study of thermonuclear neutron activation is fundamental to define a tokamak exploitation strategy. For this, two formalisme have been developed. They are based on a modelization of the activation reaction rates according to TRIPOLI, ANISN, and MERCURE 4 codes capabilities. The first one calculates, in one dimensional geometry, the desactivation gamma dose rate inside the vacuum chamber. The second one is a tridimensional model which determines the spatial variation of the gamma dose rate in the machine room. The problem of the existence of runaway electrons and associated secondaries radiations, bremsstrahlung gamma rays particularly, is approched. The results which are presented have contributed to define the parameters of the concrete shield and a strategy for TORE SUPRA Tokamak exploitation [fr

Comparison of MCNP5 and experimental results on neutron shielding effects for materials

Energy Technology Data Exchange (ETDEWEB)

Torres, D. A. (Daniel A.); Mosteller, R. D. (Russell D.); Sweezy, J. E. (Jeremy E.)

2004-01-01

The MCNP Radiation-Shielding Validation Suite was created to assess the impact on dose rates and attenuation factors of future improvements in the MCNP Monte Carlo code or its nuclear data libraries. However, it does not currently contain any deep-penetration cases. For this reason, a set of deep-penetration benchmarks has been investigated for possible inclusion in the Suite. Overall, the MCNP5 results match the measured values quite well. Furthermore, with the exception of Resin-F, there is no systematic trend in the ratio of calculated to measured results.

Benchmarking the Multidimensional Stellar Implicit Code MUSIC

Science.gov (United States)

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

2017-04-01

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

FLUKA shielding calculations for the FAIR project

International Nuclear Information System (INIS)

Fehrenbacher, Georg; Kozlova, Ekaterina; Radon, Torsten; Sokolov, Alexey

2015-01-01

FAIR is an international accelerator project being in construction at GSI Helmholtz center for heavy ion research in Darmstadt. The Monte Carlo program FLUKA is used to study radiation protection problems. The contribution deals with general application possibilities of FLUKA and for FAIR with respect the radiation protection planning. The necessity to simulate the radiation transport through shielding of several meters thickness and to determine the equivalent doses outside the shielding with sufficient accuracy is demonstrated using two examples under consideration of the variance reduction. Results of simulation calculations for activation estimation in accelerator facilities are presented.

Shield design and streaming calculations for the sodium cooled PEC reactor

International Nuclear Information System (INIS)

Prosperi, M.; Tavoni, R.; Travaglini, N.

1977-01-01

This paper summarises the shielding calculations carried out for the PEC reactor. A brief description of calculation methods and of the work carried out to set them up is given; the most representative calculations with the relative isoflux curves are also referred. A general outline is then given for the main shielding problems of the PEC reactor

Electromagnetic shielding formulae

International Nuclear Information System (INIS)

Dahlberg, E.

1979-02-01

This addendum to an earlier collection of electromagnetic shielding formulae (TRITA-EPP-75-27) contains simple transfer matrices suitable for calculating the quasistatic shielding efficiency for multiple transverse-field and axial-field cylindrical and spherical shields, as well as for estimating leakage fields from long coaxial cables and the normal-incidence transmission of a plane wave through a multiple plane shield. The differences and similarities between these cases are illustrated by means of equivalent circuits and transmission line analogies. The addendum also includes a discussion of a possible heuristic improvement of some shielding formulae. (author)

Status of the ITER tokamak nuclear shielding and radiological protection design

Energy Technology Data Exchange (ETDEWEB)

Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Chaffard, P.Y.; Izquierdo, J. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Juarez, R. [UNED, Juan del Rosal 12, Madrid 28040 (Spain); Pampin, R.; Portone, A. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain)

2016-11-01

Highlights: • Comprehensive review of design status of the ITER tokamak regarding nuclear shielding. • Investigation of shield design options and streaming mitigation measures. • Review of state-of-the-art shutdown dose rate analyses for selected port systems. - Abstract: Nuclear shielding of the ITER tokamak encompasses several systems and interfaces in a complex radiation environment. Therefore any shielding design has to involve a series of structures, systems and components in an integrated approach. This is evident for the complex ex-vessel radiation environment with streaming and leakage of plasma neutrons and subsequent activation of ex-vessel structures which give raise to excessive shutdown dose rates in accessible areas of the cryostat. The paper reviews recent nuclear analyses related to the performance of primary shields and highlights challenges toward an integrated nuclear shielding design. The general need of propagation of shielding requirements is highlighted in the context of radiation cross talk due to penetrations. Radiation streaming through gaps and penetrations is a key problem in any efficient shield design. The impact on the evolving radiation environment due to several design options along streaming paths such as port gaps, as well as their modeling for nuclear analysis, is presented. Implications regarding design integration and compliance with integrated shielding requirements and ALARA dose are finally given.

Smallest-Small-World Cellular Harmony Search for Optimization of Unconstrained Benchmark Problems

Directory of Open Access Journals (Sweden)

Sung Soo Im

2013-01-01

Full Text Available We presented a new hybrid method that combines cellular harmony search algorithms with the Smallest-Small-World theory. A harmony search (HS algorithm is based on musical performance processes that occur when a musician searches for a better state of harmony. Harmony search has successfully been applied to a wide variety of practical optimization problems. Most of the previous researches have sought to improve the performance of the HS algorithm by changing the pitch adjusting rate and harmony memory considering rate. However, there has been a lack of studies to improve the performance of the algorithm by the formation of population structures. Therefore, we proposed an improved HS algorithm that uses the cellular automata formation and the topological structure of Smallest-Small-World network. The improved HS algorithm has a high clustering coefficient and a short characteristic path length, having good exploration and exploitation efficiencies. Nine benchmark functions were applied to evaluate the performance of the proposed algorithm. Unlike the existing improved HS algorithm, the proposed algorithm is expected to have improved algorithmic efficiency from the formation of the population structure.

Some experience of shielding calculations by combinatorial method

International Nuclear Information System (INIS)

Korobejnikov, V.V.; Oussanov, V.I.

1996-01-01

Some aspects of the compound systems shielding calculations by a combinatorial approach are discussed. The effectiveness of such an approach is based on the fundamental characteristic of a compound system: if some element of the system have in itself mathematical or physical properties favorable for calculation, these properties may be used in a combinatorial approach and are lost when the system is being calculated in the whole by a direct approach. The combinatorial technique applied is well known. A compound system are being splitting for two or more auxiliary subsystems (so that calculation each of them is a more simple problem than calculation of the original problem (or at last is a soluble problem if original one is not). Calculation of every subsystem are carried out by suitable method and code, the coupling being made through boundary conditions or boundary source. The special consideration in the paper is given to a fast reactor shielding combinatorial analysis and to the testing of the results received. (author)

The use of gonad shielding in paediatric hip and pelvis radiographs.

Science.gov (United States)

Fawcett, S L; Barter, S J

2009-05-01

The problem of inaccurate placement of gonad shields in children has been highlighted by several publications nationally and internationally over the past decade and more. Here, we review the literature and present the results of a regional audit designed to assess the use and accuracy of placement of gonad shields for hip and pelvis radiographs in children. 100 consecutive anteroposterior hip or pelvis radiographs in patients under the age of 16 years were reviewed in each of 9 centres. We also included the most recent and all previously available relevant radiographs. A total of 2405 radiographs were reviewed with regard to the presence of a shield and to the accuracy of any shield placement with respect to gonad protection and visualization of orthopaedic landmarks. It is recommended that gonad shields are used in all follow-up paediatric pelvis radiographs. Our results show they were only used in 70% of such cases. When placed, only 38% of all shields were considered to be positioned accurately. For cases where shielding was indicated, an accurately placed shield was present in just 26% of radiographs. Formal written departmental guidelines for shield use were only available in two centres. We conclude that clear guidelines need to be formulated which, together with shield redesign, improved training and audit, should increase effective gonad protection for children.

The implementation of benchmarking process in marketing education services by Ukrainian universities

OpenAIRE

G.V. Okhrimenko

2016-01-01

The aim of the article. The consideration of theoretical and practical aspects of benchmarking at universities is the main task of this research. At first, the researcher identified the essence of benchmarking. It involves comparing the characteristics of college or university leading competitors in the industry and copying of proven designs. Benchmarking tries to eliminate the fundamental problem of comparison – the impossibility of being better than the one from whom they borrow solution. B...

Light-refractory radiation shielding materials using diatomites and zeolites

International Nuclear Information System (INIS)

Murakami, Hideki

2005-01-01

It has been recently shown that diatomites and zeolites have some useful characteristics for radiation shielding materials. In this study, the availability of these materials for unexpected accidents in the nuclear sites is examined. The diatomites and zeolites, compared to existing shielding materials, have superior characteristics; low density and light weight, low in radiation-induced problem, high-heat resistance, remain unaltered by the addition of an acid except hydrofluoric acid, porous and large specific surface area, and also excellent water-absorbing property. These porous materials could also expand the shielding energy range applied and be used for fast- and thermal-neutrons, and γ ray. In addition, these materials are easy to store for long periods of time against emergency because of their natural rocks. From the examinations, it is cleared that diatomites and zeolites have excellent properties as radiation shielding materials for emergency use. (author)

Faraday-Shielded dc Stark-Shift-Free Optical Lattice Clock

Science.gov (United States)

Beloy, K.; Zhang, X.; McGrew, W. F.; Hinkley, N.; Yoon, T. H.; Nicolodi, D.; Fasano, R. J.; Schäffer, S. A.; Brown, R. C.; Ludlow, A. D.

2018-05-01

We demonstrate the absence of a dc Stark shift in an ytterbium optical lattice clock. Stray electric fields are suppressed through the introduction of an in-vacuum Faraday shield. Still, the effectiveness of the shielding must be experimentally assessed. Such diagnostics are accomplished by applying high voltage to six electrodes, which are grounded in normal operation to form part of the Faraday shield. Our measurements place a constraint on the dc Stark shift at the 10-20 level, in units of the clock frequency. Moreover, we discuss a potential source of error in strategies to precisely measure or cancel nonzero dc Stark shifts, attributed to field gradients coupled with the finite spatial extent of the lattice-trapped atoms. With this consideration, we find that Faraday shielding, complemented with experimental validation, provides both a practically appealing and effective solution to the problem of dc Stark shifts in optical lattice clocks.

Progress Toward Electrostatic Radiation Shielding of Interplanetary Spacecraft: Strategies, Concepts and Technical Challenges of Human Exploration Beyond Low Earth Orbit

Science.gov (United States)

Metzger, Philip T.; Lane, John E.; Youngquist, Robert C.

2004-01-01

The radiation problem is a serious obstacle to solar system exploration. Electrostatic shielding was previously dismissed as unworkable. This was based on the false assumption that radial symmetry is needed to provide isotropic protection. KSC recently demonstrated the feasibility of asymmetric, multipole electrostatic shielding. Combined with passive shielding it might solve the radiation problem

Monte Carlo code criticality benchmark comparisons for waste packaging

International Nuclear Information System (INIS)

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

1992-07-01

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

NEACRP thermal fission product benchmark

International Nuclear Information System (INIS)

Halsall, M.J.; Taubman, C.J.

1989-09-01

The objective of the thermal fission product benchmark was to compare the range of fission product data in use at the present time. A simple homogeneous problem was set with 200 atoms H/1 atom U235, to be burnt up to 1000 days and then decay for 1000 days. The problem was repeated with 200 atoms H/1 atom Pu239, 20 atoms H/1 atom U235 and 20 atoms H/1 atom Pu239. There were ten participants and the submissions received are detailed in this report. (author)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Shielding member for thermonuclear device

Energy Technology Data Exchange (ETDEWEB)

Onozuka, Masanori

1997-06-30

In a thermonuclear device for shielding fast neutrons by shielding members disposed in a shielding vessel (vacuum vessel and structures such as a blanket disposed in the vacuum vessel), the shielding member comprises a large number of shielding wires formed fine and short so as to have elasticity. The shielding wires are sealed in a shielding vessel together with water, and when the width of the shielding vessel is changed, the shielding wires follow after the change of the width while elastically deforming in the shielding vessel, so that great stress and deformation are not formed thereby enabling to improve reliability. In addition, the length, the diameter and the shape of each of the shielding wires can be selected in accordance with the shielding space of the shielding vessel. Even if the shape of the shielding vessel is complicated, the shielding wires can be inserted easily. Accordingly, the filling rate of the shielding members can be changed easily. It can be produced more easily compared with a conventional spherical pebbles. It can be produced more easily than existent spherical shielding pebbles thereby enabling to reduce the production cost. (N.H.)

Nuclear data needs for fast breeder reactor shielding

International Nuclear Information System (INIS)

Oblow, E.M.; Perey, F.G.

1978-11-01

A review of neutron and gamma-ray cross section data needs for fast reactor shielding is presented in light of the recent advances made in assessing these needs through sensitivity studies. Total and partial cross sections and energy and angular distribution data for neutrons are surveyed as well as gamma-ray production cross sections. The strengths and deficiencies of currently available benchmark-quality integral experiments are also discussed with respect to their use in creating adjusted cross section libraries for design work. The availability of first round covariance data in ENDF/B-IV and plans for ENDF/B-V are also reviewed. This latter information makes it possible to quantitatively assess the quality of current cross section data libraries and also puts adjustment and data assessment procedures on a firmer basis

Concept of spatial channel theory applied to reactor shielding analysis

International Nuclear Information System (INIS)

Williams, M.L.; Engle, W.W. Jr.

1977-01-01

The concept of channel theory is used to locate spatial regions that are important in contributing to a shielding response. The method is analogous to the channel-theory method developed for ascertaining important energy channels in cross-section analysis. The mathematical basis for the theory is shown to be the generalized reciprocity relation, and sample problems are given to exhibit and verify properties predicted by the mathematical equations. A practical example is cited from the shielding analysis of the Fast Flux Test Facility performed at Oak Ridge National Laboratory, in which a perspective plot of channel-theory results was found useful in locating streaming paths around the reactor cavity shield

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)

Compilation report of VHTRC temperature coefficient benchmark calculations

Energy Technology Data Exchange (ETDEWEB)

Yasuda, Hideshi; Yamane, Tsuyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1995-11-01

A calculational benchmark problem has been proposed by JAERI to an IAEA Coordinated Research Program, `Verification of Safety Related Neutronic Calculation for Low-enriched Gas-cooled Reactors` to investigate the accuracy of calculation results obtained by using codes of the participating countries. This benchmark is made on the basis of assembly heating experiments at a pin-in block type critical assembly, VHTRC. Requested calculation items are the cell parameters, effective multiplication factor, temperature coefficient of reactivity, reaction rates, fission rate distribution, etc. Seven institutions from five countries have joined the benchmark works. Calculation results are summarized in this report with some remarks by the authors. Each institute analyzed the problem by applying the calculation code system which was prepared for the HTGR development of individual country. The values of the most important parameter, k{sub eff}, by all institutes showed good agreement with each other and with the experimental ones within 1%. The temperature coefficient agreed within 13%. The values of several cell parameters calculated by several institutes did not agree with the other`s ones. It will be necessary to check the calculation conditions again for getting better agreement. (J.P.N.).

A contribution to shielding effectiveness analysis of shielded tents

Directory of Open Access Journals (Sweden)

Vranić Zoran M.

2004-01-01

Full Text Available An analysis of shielding effectiveness (SE of the shielded tents made of the metallised fabrics is given. First, two electromagnetic characteristic fundamental for coupling through electrically thin shield, the skin depth break frequency and the surface resistance or transfer impedance, is defined and analyzed. Then, the transfer function and the SE are analyzed regarding to the frequency range of interest to the Electromagnetic Compatibility (EMC Community.

Cross-section fluctuations and self-shielding effects in the unresolved resonance region - International Evaluation Co-operation volume 15

International Nuclear Information System (INIS)

Froehner, F.H.; Larson, Duane C.; Tagesen, Siegfried; Petrizzi, Luigi; Hasegawa, Akira; Nakagawa, Tsuneo; Hogenbirk, Alfred; Weigmann, H.

1995-01-01

A Working Party on International Evaluation Co-operation was established under the sponsorship of the OECD/NEA Nuclear Science Committee (NSC) to promote the exchange of information on nuclear data evaluations, validation, and related topics. Its aim is also to provide a framework for co-operative activities between members of the major nuclear data evaluation projects. This includes the possible exchange of scientists in order to encourage co-operation. Requirements for experimental data resulting from this activity are compiled. The Working Party determines common criteria for evaluated nuclear data files with a view to assessing and improving the quality and completeness of evaluated data. The Parties to the project are: ENDF (United States), JEFF/EFF (NEA Data Bank Member countries), and JENDL (Japan). Co-operation with evaluation projects of non-OECD countries are organised through the Nuclear Data Section of the International Atomic Energy Agency (IAEA). NEA/NSC Subgroup 15 has had the task to assess self-shielding effects in the unresolved resonance range of structural materials, in particular their importance at various energies, and possible ways to deal with them in shielding and activation work. The principal results achieved are summarised briefly, in particular: - New data base consisting of high-resolution transmission data measured at Oak Ridge and Geel; - Improved theoretical understanding of cross-section fluctuations, including their prediction, that has been derived from the Hauser-Feshbach theory; - Benchmark results on the importance of self-shielding in iron at various energies; - Consequences for information storage in evaluated nuclear data files; - Practical utilisation of self-shielding information from evaluated files. Benchmark results as well as the Hauser-Feshbach theory show that self-shielding effects are important up to a 4-or 5-MeV neutron energy. Fluctuation factors extracted from high-resolution total cross-section data can be

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

Benchmarking LES with wall-functions and RANS for fatigue problems in thermal–hydraulics systems

Energy Technology Data Exchange (ETDEWEB)

Tunstall, R., E-mail: ryan.tunstall@manchester.ac.uk [School of MACE, The University of Manchester, Manchester M13 9PL (United Kingdom); Laurence, D.; Prosser, R. [School of MACE, The University of Manchester, Manchester M13 9PL (United Kingdom); Skillen, A. [Scientific Computing Department, STFC Daresbury Laboratory, Warrington WA4 4AD (United Kingdom)

2016-11-15

Highlights: • We benchmark LES with blended wall-functions and low-Re RANS for a pipe bend and T-Junction. • Blended wall-laws allow the first cell from the wall to be placed anywhere in the boundary layer. • In both cases LES predictions improve as the first cell wall spacing is reduced. • Near-wall temperature fluctuations in the T-Junction are overpredicted by wall-modelled LES. • The EBRSM outperforms other RANS models for the pipe bend. - Abstract: In assessing whether nuclear plant components such as T-Junctions are likely to suffer thermal fatigue problems in service, CFD techniques need to provide accurate predictions for wall temperature fluctuations. Though it has been established that this is within the capabilities of wall-resolved LES, its high computational cost has prevented widespread usage in industry. In the present paper the suitability of LES with blended wall-functions, that allow the first cell to be placed in any part of the boundary layer, is assessed. Numerical results for the flows through a 90° pipe bend and a T-Junction are compared against experimental data. Both test cases contain areas where equilibrium laws are violated in practice. It is shown that reducing the first cell wall spacing improves agreement with experimental data by limiting the extent from the wall in which the solution is constrained to an equilibrium law. The LES with wall-function approach consistently overpredicts the near-wall temperature fluctuations in the T-Junction, suggesting that it can be considered as a conservative approach. We also benchmark a range of low-Re RANS models. EBRSM predictions for the 90° pipe bend are in significantly better agreement with experimental data than those from the other models. There are discrepancies from all RANS models in the case of the T-Junction.

Transparent fast neutron shielding material and shielding method

International Nuclear Information System (INIS)

Nashimoto, Tetsuji; Katase, Haruhisa.

1993-01-01

Polyisobutylene having a viscosity average molecular weight of 20,000 to 80,000 and a hydrogen atom density of greater than 7.0 x 10 22 /cm 3 is used as a fast neutron shielding material. The shielding material is excellent in the shielding performance against fast neutrons, and there is no worry of leakage even when holes should be formed to a vessel. Further, it is excellent in fabricability, relatively safe even upon occurrence of fire and, in addition, it is transparent to enable to observe contents easily. (T.M.)

Electromagnetically shielded building

Energy Technology Data Exchange (ETDEWEB)

Takahashi, T; Nakamura, M; Yabana, Y; Ishikawa, T; Nagata, K

1992-04-21

This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs.

Electromagnetically shielded building

International Nuclear Information System (INIS)

Takahashi, T.; Nakamura, M.; Yabana, Y.; Ishikawa, T.; Nagata, K.

1992-01-01

This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs

Radiation shielding concrete

International Nuclear Information System (INIS)

Kunishima, Shigeru.

1990-01-01

The radiation shielding concretes comprise water, cement, fine aggregates consisting of serpentines and blown mist slags, coarse aggregates consisting of serpentines and kneading materials. Since serpentines containing a relatively great amount of water of crystallization in rocks as coarse aggregates and fine aggregates, the hydrogen content in the radiation shielding concretes is increased and the neutron shielding effect is improved. In addition, since serpentines are added as the fine aggregates and blown mists slags of a great specific gravity are used, the specific gravity of the shielding concretes is increased to improve the γ-ray shielding effect. Further, by the use of the kneading material having a water reducing effect and fluidizing effect, and by the bearing effect of the spherical blown mist slags used as the fine aggregates, concrete fluidity can be increased. Accordingly, workability of the radiation shielding concretes can be improved. (T.M.)

Shielding augmentation of roll-on shield from NAPS to Kaiga-2

International Nuclear Information System (INIS)

Pradhan, A.S.; Kumar, A.N.

2000-01-01

Extensive radiation field surveys were conducted in NAPS and KAPS reactor buildings as a part of commissioning checks on radiation shielding. During such surveys, dose rate higher than the expected values were noticed in fuelling machine service areas. A movable shield, separating high field area fuelling machine vault and low field area fuelling machine service area, known as roll-on shield was identified as one of the causes of high field in fuelling machine service area along with weaker end-shield. This paper discusses systematic approach adopted in bringing down the dose rates in fuelling machine service area by augmentation of roll-on shield. (author)

Grounding and shielding circuits and interference

CERN Document Server

Morrison, Ralph

2016-01-01

Applies basic field behavior in circuit design and demonstrates how it relates to grounding and shielding requirements and techniques in circuit design This book connects the fundamentals of electromagnetic theory to the problems of interference in all types of electronic design. The text covers power distribution in facilities, mixing of analog and digital circuitry, circuit board layout at high clock rates, and meeting radiation and susceptibility standards. The author examines the grounding and shielding requirements and techniques in circuit design and applies basic physics to circuit behavior. The sixth edition of this book has been updated with new material added throughout the chapters where appropriate. The presentation of the book has also been rearranged in order to reflect the current trends in the field.

Benchmarking optimization solvers for structural topology optimization

DEFF Research Database (Denmark)

Rojas Labanda, Susana; Stolpe, Mathias

2015-01-01

solvers in IPOPT and FMINCON, and the sequential quadratic programming method in SNOPT, are benchmarked on the library using performance profiles. Whenever possible the methods are applied to both the nested and the Simultaneous Analysis and Design (SAND) formulations of the problem. The performance...

EFFECTS OF INTERFACES ON GAMMA SHIELDING

Energy Technology Data Exchange (ETDEWEB)

Clifford, C. E.

1963-06-15

A survey is presented of studies of interface effects in gamma shielding problems. These studies are grouped into three types of approaches, viz.: sources at the interface; radiation backscattered from the interface; and radiation transmitted through the interface. A bibliography of 54 references is included. Limitations on the applicability of the results are discussed. (T.F.H.)

Wake Shield Target Protection

International Nuclear Information System (INIS)

Valmianski, Emanuil I.; Petzoldt, Ronald W.; Alexander, Neil B.

2003-01-01

The heat flux from both gas convection and chamber radiation on a direct drive target must be limited to avoid target damage from excessive D-T temperature increase. One of the possibilities of protecting the target is a wake shield flying in front of the target. A shield will also reduce drag force on the target, thereby facilitating target tracking and position prediction. A Direct Simulation Monte Carlo (DSMC) code was used to calculate convection heat loads as boundary conditions input into ANSYS thermal calculations. These were used for studying the quality of target protection depending on various shapes of shields, target-shield distance, and protective properties of the shield moving relative to the target. The results show that the shield can reduce the convective heat flux by a factor of 2 to 5 depending on pressure, temperature, and velocity. The protective effect of a shield moving relative to the target is greater than the protective properties of a fixed shield. However, the protective effect of a shield moving under the drag force is not sufficient for bringing the heat load on the target down to the necessary limit. Some other ways of diminishing heat flux using a protective shield are discussed

Status of multigroup cross-section data for shielding applications

International Nuclear Information System (INIS)

Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.

1983-01-01

Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V

GARLIC, a shielding program for GAmma Radiation from Line- and Cylinder-sources

Energy Technology Data Exchange (ETDEWEB)

Roos, Matts

1959-07-15

GARLIC is a program for computing the gamma ray flux or dose rate at a shielded idotropic point detector, due to a line source or the line equivalent of a cylindrical source. The source strength distribution along the line must be either uniform or an arbitrary part of the positive half-cycle of a cosine function. The line source can be oriented arbitrarily with respect to the main shield and the detector, except that the detector must not be located on the line source or on its extension. The main source is a homogeneous plane slab in which scattered radiation is accounted for by multiplying each point element of the line source by a point source build-up factor inside the integral over the point elements. Between, the main shield and the line source additional shields can be introduced, which are either plane slabs, parallel to the main shield, or cylindrical rings, coaxial with the line source. Scattered radiation in the additional shields can only be accounted for by constant build-up factors outside the integral. GARLIC-xyz is an extended version particularly suited for the frequently met problem of shielding a room containing a large number of line sources in different positions. The program computes the angles and linear dimensions of a problem for GARLIC when the positions of the detector point and the end points of the line source are given as points in an arbitrary rectangular coordinate system. As an example the isodose curves in water are presented for a monoenergetic cosine-distributed line source at several source energies and for an operating fuel element of the Swedish reactor R3.

Recycling radioactive scrap metal by producing concrete shielding with steel granules

International Nuclear Information System (INIS)

Sappok, M.

1996-01-01

Siempelkamp foundry at Krefeld, Germany, developed a method for recycling radioactively contaminated steel from nuclear installations. The material is melted and used for producing shielding plates, containers, etc., on a cast-iron basis. Because the percentage of stainless steel has recently increased significantly, problems in the production of high-quality cast iron components have also grown. The metallurgy, the contents of nickel and chromium especially, does not allow for the recycling of stainless steel in a percentage to make this process economical. In Germany, the state of the art is to use shielded concrete containers for the transport of low active waste; this concrete is produced by using hematite as an additive for increasing shielding efficiency. The plan was to produce steel granules from radioactive scrap metal as a substitute for hematite in shielding concrete

Optimal beta-ray shielding thicknesses for different therapeutic radionuclides and shielding materials

International Nuclear Information System (INIS)

Cho, Yong In; Kim, Ja Mee; Kim, Jung Hoon

2017-01-01

To better understand the distribution of deposited energy of beta and gamma rays according to changes in shielding materials and thicknesses when radionuclides are used for therapeutic nuclear medicine, a simulation was conducted. The results showed that due to the physical characteristics of each therapeutic radionuclide, the thicknesses of shielding materials at which beta-ray shielding takes place varied. Additional analysis of the shielding of gamma ray was conducted for radionuclides that emit both beta and gamma rays simultaneously with results showing shielding effects proportional to the atomic number and density of the shielding materials. Also, analysis of bremsstrahlung emission after beta-ray interactions in the simulation revealed that the occurrence of bremsstrahlung was relatively lower than theoretically calculated and varied depending on different radionuclides. (authors)

A large-scale benchmark of gene prioritization methods.

Science.gov (United States)

Guala, Dimitri; Sonnhammer, Erik L L

2017-04-21

In order to maximize the use of results from high-throughput experimental studies, e.g. GWAS, for identification and diagnostics of new disease-associated genes, it is important to have properly analyzed and benchmarked gene prioritization tools. While prospective benchmarks are underpowered to provide statistically significant results in their attempt to differentiate the performance of gene prioritization tools, a strategy for retrospective benchmarking has been missing, and new tools usually only provide internal validations. The Gene Ontology(GO) contains genes clustered around annotation terms. This intrinsic property of GO can be utilized in construction of robust benchmarks, objective to the problem domain. We demonstrate how this can be achieved for network-based gene prioritization tools, utilizing the FunCoup network. We use cross-validation and a set of appropriate performance measures to compare state-of-the-art gene prioritization algorithms: three based on network diffusion, NetRank and two implementations of Random Walk with Restart, and MaxLink that utilizes network neighborhood. Our benchmark suite provides a systematic and objective way to compare the multitude of available and future gene prioritization tools, enabling researchers to select the best gene prioritization tool for the task at hand, and helping to guide the development of more accurate methods.

Developing a benchmark for emotional analysis of music.

Science.gov (United States)

Aljanaki, Anna; Yang, Yi-Hsuan; Soleymani, Mohammad

2017-01-01

Music emotion recognition (MER) field rapidly expanded in the last decade. Many new methods and new audio features are developed to improve the performance of MER algorithms. However, it is very difficult to compare the performance of the new methods because of the data representation diversity and scarcity of publicly available data. In this paper, we address these problems by creating a data set and a benchmark for MER. The data set that we release, a MediaEval Database for Emotional Analysis in Music (DEAM), is the largest available data set of dynamic annotations (valence and arousal annotations for 1,802 songs and song excerpts licensed under Creative Commons with 2Hz time resolution). Using DEAM, we organized the 'Emotion in Music' task at MediaEval Multimedia Evaluation Campaign from 2013 to 2015. The benchmark attracted, in total, 21 active teams to participate in the challenge. We analyze the results of the benchmark: the winning algorithms and feature-sets. We also describe the design of the benchmark, the evaluation procedures and the data cleaning and transformations that we suggest. The results from the benchmark suggest that the recurrent neural network based approaches combined with large feature-sets work best for dynamic MER.

Benchmarking in University Toolbox

Directory of Open Access Journals (Sweden)

Katarzyna Kuźmicz

2015-06-01

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

TORE-SUPRA: design of thermal radiation shield at 80 K

International Nuclear Information System (INIS)

Aymar, R.; Cordier, J.J.; Deschamps, P.; Gauthier, A.; Perin, J.P.

1982-09-01

The TORE-SUPRA superconducting toroidal magnet operating at liquid helium temperature, must be protected against thermal radiation from the vessels. For this purpose, stainless steel heat shields, cooled at 80 K, are positioned between coil casings at 4.5 K and the vessels, and constitute a double stiff toroid which completely surrounds the magnet. Mockups have been manufactured to study their design and operating problems. Calculations have also been made to analyse the mechanical behaviour of these shields

The shielding calculation for the CN guide shielding assembly in HANARO

International Nuclear Information System (INIS)

Kim, H. S.; Lee, B. C.; Lee, K. H.; Kim, H.

2006-01-01

The cold neutron research facility in HANARO is under construction. The area including neutron guides and rotary shutter in the reactor hall should be shielded by the guide shielding assembly which is constructed of heavy concrete blocks and structure. The guide shielding assembly is divided into 2 parts, A and B. Part A is about 6.4 meters apart from the reactor biological shield and it is constructed of heavy concrete blocks whose density is above 4.0g/cm 3 . And part B is a fixed heavy concrete structure whose density is above 3.5g/cm 3 . The rotary shutter is also made with heavy concrete whose density is above 4.0g/cm 3 and includes 5 neutron guides inside. It can block the neutron beam by rotating when CNS is not operating. The dose criterion outside the guide shielding assembly is established as 12.5 μSv/hr which is also applied to reactor shielding in HANARO

Shielding synchrotron light sources: Advantages of circular shield walls tunnels

Energy Technology Data Exchange (ETDEWEB)

Kramer, S.L. [Design and Accelerator Operations Consulting, 568 Wintergreen Ct Ridge, NY 11961 (United States); Ghosh, V.J.; Breitfeller, M. [NSLS-II, Brookhaven National Laboratory, Upton, NY 11973 (United States)

2016-08-11

Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produce significantly higher neutron component dose to the experimental floor than lower energy injection and ramped operations. High energy neutrons produced in the forward direction from thin target beam losses are a major component of the dose rate outside the shield walls of the tunnel. The convention has been to provide thicker 90° ratchet walls to reduce this dose to the beam line users. We present an alternate circular shield wall design, which naturally and cost effectively increases the path length for this forward radiation in the shield wall and thereby substantially decreasing the dose rate for these beam losses. This shield wall design will greatly reduce the dose rate to the users working near the front end optical components but will challenge the beam line designers to effectively utilize the longer length of beam line penetration in the shield wall. Additional advantages of the circular shield wall tunnel are that it's simpler to construct, allows greater access to the insertion devices and the upstream in tunnel beam line components, as well as reducing the volume of concrete and therefore the cost of the shield wall.

Three-dimensional coupled Monte Carlo-discrete ordinates computational scheme for shielding calculations of large and complex nuclear facilities

International Nuclear Information System (INIS)

Chen, Y.; Fischer, U.

2005-01-01

Shielding calculations of advanced nuclear facilities such as accelerator based neutron sources or fusion devices of the tokamak type are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. While the complexity of the geometry in the shielding calculation can be hardly handled by the discrete ordinates method, the deep penetration of radiation through bulk shields is a severe challenge for the Monte Carlo particle transport technique. This work proposes a dedicated computational scheme for coupled Monte Carlo-Discrete Ordinates transport calculations to handle this kind of shielding problems. The Monte Carlo technique is used to simulate the particle generation and transport in the target region with both complex geometry and reaction physics, and the discrete ordinates method is used to treat the deep penetration problem in the bulk shield. The coupling scheme has been implemented in a program system by loosely integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a newly developed coupling interface program for mapping process. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities. (authors)

Neutron and gamma ray transport calculations in shielding system

Energy Technology Data Exchange (ETDEWEB)

Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

Library Benchmarking

Directory of Open Access Journals (Sweden)

Wiji Suwarno

2017-02-01

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

The PAC-MAN model: Benchmark case for linear acoustics in computational physics

Science.gov (United States)

Ziegelwanger, Harald; Reiter, Paul

2017-10-01

Benchmark cases in the field of computational physics, on the one hand, have to contain a certain complexity to test numerical edge cases and, on the other hand, require the existence of an analytical solution, because an analytical solution allows the exact quantification of the accuracy of a numerical simulation method. This dilemma causes a need for analytical sound field formulations of complex acoustic problems. A well known example for such a benchmark case for harmonic linear acoustics is the ;Cat's Eye model;, which describes the three-dimensional sound field radiated from a sphere with a missing octant analytically. In this paper, a benchmark case for two-dimensional (2D) harmonic linear acoustic problems, viz., the ;PAC-MAN model;, is proposed. The PAC-MAN model describes the radiated and scattered sound field around an infinitely long cylinder with a cut out sector of variable angular width. While the analytical calculation of the 2D sound field allows different angular cut-out widths and arbitrarily positioned line sources, the computational cost associated with the solution of this problem is similar to a 1D problem because of a modal formulation of the sound field in the PAC-MAN model.

Compilation of MCNP data library based on JENDL-3T and test through analysis of benchmark experiment

International Nuclear Information System (INIS)

Sakurai, K.; Sasamoto, N.; Kosako, K.; Ishikawa, T.; Sato, O.; Oyama, Y.; Narita, H.; Maekawa, H.; Ueki, K.

1989-01-01

Based on an evaluated nuclear data library JENDL-3T, a temporary version of JENDL-3, a pointwise neutron cross section library for MCNP code is compiled which involves 39 nuclides from H-1 to Am-241 which are important for shielding calculations. Compilation is performed with the code system which consists of the nuclear data processing code NJOY-83 and library compilation code MACROS. Validity of the code system and reliability of the library are certified by analysing benchmark experiments. (author)

INTOR radiation shielding for personnel access

International Nuclear Information System (INIS)

Gohar, Y.; Abdou, M.

1981-01-01

The INTOR reactor shield system consists of the blanket, bulk shield, penetration shield, component shield, and biological shield. The bulk shield consists of two parts: (a) the inboard shield; and (b) the outboard shield. The distinction between the different components of the shield system is essential to satisfy the different design constraints and achieve various objectives

Neutron shieldings

International Nuclear Information System (INIS)

Tarutani, Kohei

1979-01-01

Purpose: To decrease the stresses resulted by the core bendings to the base of an entrance nozzle. Constitution: Three types of round shielding rods of different diameter are arranged in a hexagonal tube. The hexagonal tube is provided with several spacer pads receiving the loads from the core constrain mechanism at its outer circumference, a handling head for a fuel exchanger at its top and an entrance nozzle for self-holding the neutron shieldings and flowing heat-removing coolants at its bottom. The diameters for R 1 , R 2 and R 3 for the round shielding rods are designed as: 0.1 R 1 2 1 and 0.2 R 1 2 1 . Since a plurality of shielding rods of small diameter are provided, soft structure are obtained and a plurality of coolant paths are formed. (Furukawa, Y.)

Performance of Multi-chaotic PSO on a shifted benchmark functions set

Energy Technology Data Exchange (ETDEWEB)

Pluhacek, Michal; Senkerik, Roman; Zelinka, Ivan [Tomas Bata University in Zlín, Faculty of Applied Informatics Department of Informatics and Artificial Intelligence nám. T.G. Masaryka 5555, 760 01 Zlín (Czech Republic)

2015-03-10

In this paper the performance of Multi-chaotic PSO algorithm is investigated using two shifted benchmark functions. The purpose of shifted benchmark functions is to simulate the time-variant real-world problems. The results of chaotic PSO are compared with canonical version of the algorithm. It is concluded that using the multi-chaotic approach can lead to better results in optimization of shifted functions.

Performance of Multi-chaotic PSO on a shifted benchmark functions set

International Nuclear Information System (INIS)

Pluhacek, Michal; Senkerik, Roman; Zelinka, Ivan

2015-01-01

In this paper the performance of Multi-chaotic PSO algorithm is investigated using two shifted benchmark functions. The purpose of shifted benchmark functions is to simulate the time-variant real-world problems. The results of chaotic PSO are compared with canonical version of the algorithm. It is concluded that using the multi-chaotic approach can lead to better results in optimization of shifted functions

Test One to Test Many: A Unified Approach to Quantum Benchmarks

Science.gov (United States)

Bai, Ge; Chiribella, Giulio

2018-04-01

Quantum benchmarks are routinely used to validate the experimental demonstration of quantum information protocols. Many relevant protocols, however, involve an infinite set of input states, of which only a finite subset can be used to test the quality of the implementation. This is a problem, because the benchmark for the finitely many states used in the test can be higher than the original benchmark calculated for infinitely many states. This situation arises in the teleportation and storage of coherent states, for which the benchmark of 50% fidelity is commonly used in experiments, although finite sets of coherent states normally lead to higher benchmarks. Here, we show that the average fidelity over all coherent states can be indirectly probed with a single setup, requiring only two-mode squeezing, a 50-50 beam splitter, and homodyne detection. Our setup enables a rigorous experimental validation of quantum teleportation, storage, amplification, attenuation, and purification of noisy coherent states. More generally, we prove that every quantum benchmark can be tested by preparing a single entangled state and measuring a single observable.

Radiation Shielding and Hydrogen Storage with Multifunctional Carbon, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — This project addresses two vital problems for long-term space travel activities: radiation shielding and hydrogen storage for power and propulsion. While both...

Benchmarking and Performance Measurement.

Science.gov (United States)

Town, J. Stephen

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

CSI-EPT in Presence of RF-Shield for MR-Coils.

Science.gov (United States)

Arduino, Alessandro; Zilberti, Luca; Chiampi, Mario; Bottauscio, Oriano

2017-07-01

Contrast source inversion electric properties tomography (CSI-EPT) is a recently developed technique for the electric properties tomography that recovers the electric properties distribution starting from measurements performed by magnetic resonance imaging scanners. This method is an optimal control approach based on the contrast source inversion technique, which distinguishes itself from other electric properties tomography techniques for its capability to recover also the local specific absorption rate distribution, essential for online dosimetry. Up to now, CSI-EPT has only been described in terms of integral equations, limiting its applicability to homogeneous unbounded background. In order to extend the method to the presence of a shield in the domain-as in the recurring case of shielded radio frequency coils-a more general formulation of CSI-EPT, based on a functional viewpoint, is introduced here. Two different implementations of CSI-EPT are proposed for a 2-D transverse magnetic model problem, one dealing with an unbounded domain and one considering the presence of a perfectly conductive shield. The two implementations are applied on the same virtual measurements obtained by numerically simulating a shielded radio frequency coil. The results are compared in terms of both electric properties recovery and local specific absorption rate estimate, in order to investigate the requirement of an accurate modeling of the underlying physical problem.

Building Bridges Between Geoscience and Data Science through Benchmark Data Sets

Science.gov (United States)

Thompson, D. R.; Ebert-Uphoff, I.; Demir, I.; Gel, Y.; Hill, M. C.; Karpatne, A.; Güereque, M.; Kumar, V.; Cabral, E.; Smyth, P.

2017-12-01

The changing nature of observational field data demands richer and more meaningful collaboration between data scientists and geoscientists. Thus, among other efforts, the Working Group on Case Studies of the NSF-funded RCN on Intelligent Systems Research To Support Geosciences (IS-GEO) is developing a framework to strengthen such collaborations through the creation of benchmark datasets. Benchmark datasets provide an interface between disciplines without requiring extensive background knowledge. The goals are to create (1) a means for two-way communication between geoscience and data science researchers; (2) new collaborations, which may lead to new approaches for data analysis in the geosciences; and (3) a public, permanent repository of complex data sets, representative of geoscience problems, useful to coordinate efforts in research and education. The group identified 10 key elements and characteristics for ideal benchmarks. High impact: A problem with high potential impact. Active research area: A group of geoscientists should be eager to continue working on the topic. Challenge: The problem should be challenging for data scientists. Data science generality and versatility: It should stimulate development of new general and versatile data science methods. Rich information content: Ideally the data set provides stimulus for analysis at many different levels. Hierarchical problem statement: A hierarchy of suggested analysis tasks, from relatively straightforward to open-ended tasks. Means for evaluating success: Data scientists and geoscientists need means to evaluate whether the algorithms are successful and achieve intended purpose. Quick start guide: Introduction for data scientists on how to easily read the data to enable rapid initial data exploration. Geoscience context: Summary for data scientists of the specific data collection process, instruments used, any pre-processing and the science questions to be answered. Citability: A suitable identifier to

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 testing and independent verification of the VS2DT computer code

International Nuclear Information System (INIS)

McCord, J.T.

1994-11-01

The finite difference flow and transport simulator VS2DT was benchmark tested against several other codes which solve the same equations (Richards equation for flow and the Advection-Dispersion equation for transport). The benchmark problems investigated transient two-dimensional flow in a heterogeneous soil profile with a localized water source at the ground surface. The VS2DT code performed as well as or better than all other codes when considering mass balance characteristics and computational speed. It was also rated highly relative to the other codes with regard to ease-of-use. Following the benchmark study, the code was verified against two analytical solutions, one for two-dimensional flow and one for two-dimensional transport. These independent verifications show reasonable agreement with the analytical solutions, and complement the one-dimensional verification problems published in the code's original documentation

Demonstration study on shielding safety analysis code (8)

Energy Technology Data Exchange (ETDEWEB)

Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan)

2001-03-01

Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated. (1) A {sup 3}He detector and some instruments are added to the former detection system to increase the detection sensitivity in pulsed neutron measurements. Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility are measured in the distance up to 350 m. (2) To estimate the spectrum of leakage neutron from the facility, {sup 3}He detector with moderators is constructed and the response functions of the detector are calculated using the MCNP simulation code. The leakage spectrum in the facility are measured and unfolded using the SAND-II code. (3) Using the EGS code and/or MCNP code, neutron yields by the photo-nuclear reaction in the lead target are calculated. Then, the neutron fluence at some points including the duct (from which neutrons leaks and is considered to be a skyshine source) is simulated by MCNP MONTE CARLO code. (4) In the distance up to 350 m from the facility, neutron fluence due to the skyshine process are calculated and compared with the experimental results. The comparison gives a fairly good agreement. (author)

Benchmarking in the Netherlands

International Nuclear Information System (INIS)

1999-01-01

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

Radiation shielding plate

International Nuclear Information System (INIS)

Kobayashi, Torakichi; Sugawara, Takeo.

1983-01-01

Purpose: To reduce the weight and stabilize the configuration of a radiation shielding plate which is used in close contact with an object to be irradiated with radiation rays. Constitution: The radiation shielding plate comprises a substrate made of lead glass and a metallic lead coating on the surface of the substrate by means of plating, vapor deposition or the like. Apertures for permeating radiation rays are formed to the radiation shielding plate. Since the shielding plate is based on a lead glass plate, a sufficient mechanical strength can be obtained with a thinner structure as compared with the conventional plate made of metallic lead. Accordingly, if the shielding plate is disposed on a soft object to be irradiated with radiation rays, the object and the plate itself less deform to obtain a radiation irradiation pattern with distinct edges. (Moriyama, K.)

Infinite slab-shield dose calculations

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

I calculated neutron and gamma-ray equivalent doses leaking through a variety of infinite (laminate) slab-shields. In the shield computations, I used, as the incident neutron spectrum, the leakage spectrum (<20 MeV) calculated for the LANSCE tungsten production target at 90 degree to the target axis. The shield thickness was fixed at 60 cm. The results of the shield calculations show a minimum in the total leakage equivalent dose if the shield is 40-45 cm of iron followed by 20-15 cm of borated (5% B) polyethylene. High-performance shields can be attained by using multiple laminations. The calculated dose at the shield surface is very dependent on shield material. 4 refs., 4 figs., 1 tab

Designing shields for KeV photons with genetic algorithms

International Nuclear Information System (INIS)

Asbury, Stephen; Holloway, James P.

2011-01-01

Shielding of x-ray sources and low energy gamma rays is often accomplished with lead aprons, comprising a thin layer (0.5 mm to 1 mm) of lead or similar high-Z material. In previous work the authors used Genetic Algorithms to explore the design of a shadow shield for space applications. Now those techniques have been applied to the problem of shielding humans from low energy gamma radiation. This paper uses a simple geometry to explore layering various materials as a method to reduce mass and dose for thin gamma shields. The genetic algorithms discover layers of materials with various Z is in fact more effective than an equivalent mass of Pb alone for lower energy gammas, but as the incident radiation energy increases the efficacy of such layering diminishes. The utility of varying Z for lower energy gammas is in part due to their complementary K-edges, where one material compensates for the transmission that would occur just below the K-edge in another material. (author)

Electromagnetic shield

International Nuclear Information System (INIS)

Miller, J.S.

1987-01-01

An electromagnetic shield is described comprising: closed, electrically-conductive rings, each having an open center; and binder means for arranging the rings in a predetermined, fixed relationship relative to each other, the so-arranged rings and binder means defining an outer surface; wherein electromagnetic energy received by the shield from a source adjacent its outer surface induces an electrical current to flow in a predetermined direction adjacent and parallel to the outer surface, through the rings; and wherein each ring is configured to cause source-induced alternating current flowing through the portion of the ring closest to the outer surface to electromagnetically induce an oppositely-directed current in the portion of the ring furthest from the surface, such oppositely-directed current bucking any source-induced current in the latter ring portion and thus reducing the magnitude of current flowing through it, whereby the electromagnetic shielding effected by the shield is enhanced

Magnetic Materials Characterization and Modeling for the Enhanced Design of Magnetic Shielding of Cryomodules in Particle Accelerators

Energy Technology Data Exchange (ETDEWEB)

Sah, Sanjay [Virginia Commonwealth Univ., Richmond, VA (United States)

2016-05-31

Particle accelerators produce beams of high-energy particles, which are used for both fundamental and applied scientific research and are critical to the development of accelerator driven sub-critical reactor systems. An effective magnetic shield is very important to achieve higher quality factor (Qo) of the cryomodule of a particle accelerator. The allowed value of field inside the cavity due to all external fields (particularly the Earth’s magnetic field) is ~15 mG or less. The goal of this PhD dissertation is to comprehensively study the magnetic properties of commonly used magnetic shielding materials at both cryogenic and room temperatures. This knowledge can be used for the enhanced design of magnetic shields of cryomodes (CM) in particle accelerators. To this end, we first studied the temperature dependent magnetization behavior (M-H curves) of Amumetal and A4K under different annealing and deformation conditions. This characterized the effect of stress or deformation induced during the manufacturing processes and subsequent restoration of high permeability with appropriate heat treatment. Next, an energy based stochastic model for temperature dependent anhysteretic magnetization behavior of ferromagnetic materials was proposed and benchmarked against experimental data. We show that this model is able to simulate and explain the magnetic behavior of as rolled, deformed and annealed amumetal and A4K over a large range of temperatures. The experimental results for permeability are then used in a finite element model (FEM) in COMSOL to evaluate the shielding effectiveness of multiple shield designs at room temperature as well as cryogenic temperature. This work could serve as a guideline for future design, development and fabrication of magnetic shields of CMs.

The SWAN/NPSOL code system for multivariable multiconstraint shield optimization

International Nuclear Information System (INIS)

Watkins, E.F.; Greenspan, E.

1995-01-01

SWAN is a useful code for optimization of source-driven systems, i.e., systems for which the neutron and photon distribution is the solution of the inhomogeneous transport equation. Over the years, SWAN has been applied to the optimization of a variety of nuclear systems, such as minimizing the thickness of fusion reactor blankets and shields, the weight of space reactor shields, the cost for an ICF target chamber shield, and the background radiation for explosive detection systems and maximizing the beam quality for boron neutron capture therapy applications. However, SWAN's optimization module can handle up to a single constraint and was inefficient in handling problems with many variables. The purpose of this work is to upgrade SWAN's optimization capability

WLUP benchmarks

International Nuclear Information System (INIS)

Leszczynski, Francisco

2002-01-01

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

NRC-BNL Benchmark Program on Evaluation of Methods for Seismic Analysis of Coupled Systems

International Nuclear Information System (INIS)

Chokshi, N.; DeGrassi, G.; Xu, J.

1999-01-01

A NRC-BNL benchmark program for evaluation of state-of-the-art analysis methods and computer programs for seismic analysis of coupled structures with non-classical damping is described. The program includes a series of benchmarking problems designed to investigate various aspects of complexities, applications and limitations associated with methods for analysis of non-classically damped structures. Discussions are provided on the benchmarking process, benchmark structural models, and the evaluation approach, as well as benchmarking ground rules. It is expected that the findings and insights, as well as recommendations from this program will be useful in developing new acceptance criteria and providing guidance for future regulatory activities involving licensing applications of these alternate methods to coupled systems

Shielding benchmark experiments and sensitivity studies in progress at some European laboratories

International Nuclear Information System (INIS)