Benchmark density functional theory calculations for nanoscale conductance

DEFF Research Database (Denmark)

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

2008-01-01

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

Calculation of the fifth atomic energy research dynamic benchmark with APROS

International Nuclear Information System (INIS)

Puska Eija Karita; Kontio Harii

1998-01-01

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

The fifth AER dynamic benchmark calculation with hextran-smabre

International Nuclear Information System (INIS)

Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Benchmark neutron porosity log calculations

International Nuclear Information System (INIS)

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

1989-01-01

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

Reactor calculation benchmark PCA blind test results

International Nuclear Information System (INIS)

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

1980-01-01

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

Reactor calculation benchmark PCA blind test results

Energy Technology Data Exchange (ETDEWEB)

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

1980-01-01

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

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

International Nuclear Information System (INIS)

Ishikawa, Makoto

2003-01-01

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

HEU benchmark calculations and LEU preliminary calculations for IRR-1

International Nuclear Information System (INIS)

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

2004-01-01

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

Benchmark calculation of subchannel analysis codes

International Nuclear Information System (INIS)

1996-02-01

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

Benchmark calculation of nuclear design code for HCLWR

International Nuclear Information System (INIS)

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

1986-01-01

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

Monte Carlo benchmark calculations for 400MWTH PBMR core

International Nuclear Information System (INIS)

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

2007-01-01

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

Benchmark testing calculations for 232Th

International Nuclear Information System (INIS)

Liu Ping

2003-01-01

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

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

3-D extension C5G7 MOX benchmark calculation using threedant code

International Nuclear Information System (INIS)

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

2005-01-01

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

Reactor fuel depletion benchmark of TINDER

International Nuclear Information System (INIS)

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

2014-01-01

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

MOx Depletion Calculation Benchmark

International Nuclear Information System (INIS)

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

2016-01-01

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

EA-MC Neutronic Calculations on IAEA ADS Benchmark 3.2

Energy Technology Data Exchange (ETDEWEB)

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

2006-01-15

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

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 calculation programme concerning typical LMFBR structures

International Nuclear Information System (INIS)

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

1982-01-01

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

FENDL-2 and associated benchmark calculations

International Nuclear Information System (INIS)

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

1992-03-01

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

Method of characteristics - Based sensitivity calculations for international PWR benchmark

International Nuclear Information System (INIS)

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

2013-01-01

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

Benchmark Calculations of Noncovalent Interactions of Halogenated Molecules

Czech Academy of Sciences Publication Activity Database

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

2012-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Burn-up TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

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

1998-01-01

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

COVE 2A Benchmarking calculations using NORIA

International Nuclear Information System (INIS)

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

1991-10-01

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

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

International Nuclear Information System (INIS)

Haenaelaeinen, Anitta

1998-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Benchmark Calculations on Halden IFA-650 LOCA Test Results

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

Van de Wetering, T.F.H.

1993-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

BENCHMARKING ORTEC ISOTOPIC MEASUREMENTS AND CALCULATIONS

Energy Technology Data Exchange (ETDEWEB)

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

2008-09-29

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

Highly Accurate Calculations of the Phase Diagram of Cold Lithium

Science.gov (United States)

Shulenburger, Luke; Baczewski, Andrew

The phase diagram of lithium is particularly complicated, exhibiting many different solid phases under the modest application of pressure. Experimental efforts to identify these phases using diamond anvil cells have been complemented by ab initio theory, primarily using density functional theory (DFT). Due to the multiplicity of crystal structures whose enthalpy is nearly degenerate and the uncertainty introduced by density functional approximations, we apply the highly accurate many-body diffusion Monte Carlo (DMC) method to the study of the solid phases at low temperature. These calculations span many different phases, including several with low symmetry, demonstrating the viability of DMC as a method for calculating phase diagrams for complex solids. Our results can be used as a benchmark to test the accuracy of various density functionals. This can strengthen confidence in DFT based predictions of more complex phenomena such as the anomalous melting behavior predicted for lithium at high pressures. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. DOE's National Nuclear Security Administration under contract DE-AC04-94AL85000.

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

Directory of Open Access Journals (Sweden)

Stanisz Przemysław

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Dourougie, C

1998-07-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Ellis, RJ

2001-02-02

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

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

Science.gov (United States)

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

2015-04-01

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

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

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

CFD-calculations to a core catcher benchmark

International Nuclear Information System (INIS)

Willschuetz, H.G.

1999-04-01

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

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

International Nuclear Information System (INIS)

Dorning, J.

1997-09-01

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

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

International Nuclear Information System (INIS)

Ishikawa, M.

2002-01-01

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

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

Science.gov (United States)

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

1991-01-01

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

Accurate structures and energetics of neutral-framework zeotypes from dispersion-corrected DFT calculations

Science.gov (United States)

Fischer, Michael; Angel, Ross J.

2017-05-01

Density-functional theory (DFT) calculations incorporating a pairwise dispersion correction were employed to optimize the structures of various neutral-framework compounds with zeolite topologies. The calculations used the PBE functional for solids (PBEsol) in combination with two different dispersion correction schemes, the D2 correction devised by Grimme and the TS correction of Tkatchenko and Scheffler. In the first part of the study, a benchmarking of the DFT-optimized structures against experimental crystal structure data was carried out, considering a total of 14 structures (8 all-silica zeolites, 4 aluminophosphate zeotypes, and 2 dense phases). Both PBEsol-D2 and PBEsol-TS showed an excellent performance, improving significantly over the best-performing approach identified in a previous study (PBE-TS). The temperature dependence of lattice parameters and bond lengths was assessed for those zeotypes where the available experimental data permitted such an analysis. In most instances, the agreement between DFT and experiment improved when the experimental data were corrected for the effects of thermal motion and when low-temperature structure data rather than room-temperature structure data were used as a reference. In the second part, a benchmarking against experimental enthalpies of transition (with respect to α-quartz) was carried out for 16 all-silica zeolites. Excellent agreement was obtained with the PBEsol-D2 functional, with the overall error being in the same range as the experimental uncertainty. Altogether, PBEsol-D2 can be recommended as a computationally efficient DFT approach that simultaneously delivers accurate structures and energetics of neutral-framework zeotypes.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

One dimensional benchmark calculations using diffusion theory

International Nuclear Information System (INIS)

Ustun, G.; Turgut, M.H.

1986-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

DeHart, M.D.

1993-01-01

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

JENDL-4.0 benchmarking for fission reactor applications

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

Arigane, Kenji

1987-04-01

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

The ORSphere Benchmark Evaluation and Its Potential Impact on Nuclear Criticality Safety

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; Margaret A. Marshall; J. Blair Briggs

2013-10-01

In the early 1970’s, critical experiments using an unreflected metal sphere of highly enriched uranium (HEU) were performed with the focus to provide a “very accurate description…as an ideal benchmark for calculational methods and cross-section data files.” Two near-critical configurations of the Oak Ridge Sphere (ORSphere) were evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results from those benchmark experiments were then compared with additional unmoderated and unreflected HEU metal benchmark experiment configurations currently found in the ICSBEP Handbook. For basic geometries (spheres, cylinders, and slabs) the eigenvalues calculated using MCNP5 and ENDF/B-VII.0 were within 3 of their respective benchmark values. There appears to be generally a good agreement between calculated and benchmark values for spherical and slab geometry systems. Cylindrical geometry configurations tended to calculate low, including more complex bare HEU metal systems containing cylinders. The ORSphere experiments do not calculate within their 1s uncertainty and there is a possibility that the effect of the measured uncertainties for the GODIVA I benchmark may need reevaluated. There is significant scatter in the calculations for the highly-correlated ORCEF cylinder experiments, which are constructed from close-fitting HEU discs and annuli. Selection of a nuclear data library can have a larger impact on calculated eigenvalue results than the variation found within calculations of a given experimental series, such as the ORCEF cylinders, using a single nuclear data set.

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-07-01

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

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

International Nuclear Information System (INIS)

Elina Syrjaelahti; Anitta Haemaelaeinen

2005-01-01

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

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

International Nuclear Information System (INIS)

Corsi, F.

1985-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

DeHart, M.D.

1993-01-01

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

Benchmark calculations with simple phantom for neutron dosimetry (2)

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

Akie, Hiroshi; Ishiguro, Yukio; Takano, Hideki

1988-10-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

Okuno, Hiroshi

2003-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

CERN Document Server

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

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

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

CERN Document Server

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

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

MCNP neutron benchmarks

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-15

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

LCEs for Naval Reactor Benchmark Calculations

International Nuclear Information System (INIS)

W.J. Anderson

1999-01-01

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

Benchmark calculations by KENO-Va using the JEF 2.2 library

Energy Technology Data Exchange (ETDEWEB)

Markova, L.

1994-12-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

HELIOS calculations for UO2 lattice benchmarks

International Nuclear Information System (INIS)

Mosteller, R.D.

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Tensor-decomposed vibrational coupled-cluster theory: Enabling large-scale, highly accurate vibrational-structure calculations

Science.gov (United States)

Madsen, Niels Kristian; Godtliebsen, Ian H.; Losilla, Sergio A.; Christiansen, Ove

2018-01-01

A new implementation of vibrational coupled-cluster (VCC) theory is presented, where all amplitude tensors are represented in the canonical polyadic (CP) format. The CP-VCC algorithm solves the non-linear VCC equations without ever constructing the amplitudes or error vectors in full dimension but still formally includes the full parameter space of the VCC[n] model in question resulting in the same vibrational energies as the conventional method. In a previous publication, we have described the non-linear-equation solver for CP-VCC calculations. In this work, we discuss the general algorithm for evaluating VCC error vectors in CP format including the rank-reduction methods used during the summation of the many terms in the VCC amplitude equations. Benchmark calculations for studying the computational scaling and memory usage of the CP-VCC algorithm are performed on a set of molecules including thiadiazole and an array of polycyclic aromatic hydrocarbons. The results show that the reduced scaling and memory requirements of the CP-VCC algorithm allows for performing high-order VCC calculations on systems with up to 66 vibrational modes (anthracene), which indeed are not possible using the conventional VCC method. This paves the way for obtaining highly accurate vibrational spectra and properties of larger molecules.

An accurate and linear-scaling method for calculating charge-transfer excitation energies and diabatic couplings

Energy Technology Data Exchange (ETDEWEB)

Pavanello, Michele [Department of Chemistry, Rutgers University, Newark, New Jersey 07102-1811 (United States); Van Voorhis, Troy [Department of Chemistry, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139-4307 (United States); Visscher, Lucas [Amsterdam Center for Multiscale Modeling, VU University, De Boelelaan 1083, 1081 HV Amsterdam (Netherlands); Neugebauer, Johannes [Theoretische Organische Chemie, Organisch-Chemisches Institut der Westfaelischen Wilhelms-Universitaet Muenster, Corrensstrasse 40, 48149 Muenster (Germany)

2013-02-07

Quantum-mechanical methods that are both computationally fast and accurate are not yet available for electronic excitations having charge transfer character. In this work, we present a significant step forward towards this goal for those charge transfer excitations that take place between non-covalently bound molecules. In particular, we present a method that scales linearly with the number of non-covalently bound molecules in the system and is based on a two-pronged approach: The molecular electronic structure of broken-symmetry charge-localized states is obtained with the frozen density embedding formulation of subsystem density-functional theory; subsequently, in a post-SCF calculation, the full-electron Hamiltonian and overlap matrix elements among the charge-localized states are evaluated with an algorithm which takes full advantage of the subsystem DFT density partitioning technique. The method is benchmarked against coupled-cluster calculations and achieves chemical accuracy for the systems considered for intermolecular separations ranging from hydrogen-bond distances to tens of Angstroms. Numerical examples are provided for molecular clusters comprised of up to 56 non-covalently bound molecules.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-14

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

ROLAIDS-CPM: A code for accurate resonance absorption calculations

International Nuclear Information System (INIS)

Kruijf, W.J.M. de.

1993-08-01

ROLAIDS is used to calculate group-averaged cross sections for specific zones in a one-dimensional geometry. This report describes ROLAIDS-CPM which is an extended version of ROLAIDS. The main extension in ROLAIDS-CPM is the possibility to use the collision probability method for a slab- or cylinder-geometry instead of the less accurate interface-currents method. In this way accurate resonance absorption calculations can be performed with ROLAIDS-CPM. ROLAIDS-CPM has been developed at ECN. (orig.)

Criticality benchmark comparisons leading to cross-section upgrades

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Freudenreich, W.E.; Gruppelaar, H

1998-12-01

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

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

International Nuclear Information System (INIS)

Freudenreich, W.E.; Gruppelaar, H.

1998-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

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

International Nuclear Information System (INIS)

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

1996-06-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Static benchmarking of the NESTLE advanced nodal code

International Nuclear Information System (INIS)

Mosteller, R.D.

1997-01-01

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

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

International Nuclear Information System (INIS)

Miyoshi, Yoshinori; Yamamoto, Toshihiro; Nakamura, Takemi

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-15

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

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

International Nuclear Information System (INIS)

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

2006-07-01

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

Accurate calculation of the geometric measure of entanglement for multipartite quantum states

Science.gov (United States)

Teng, Peiyuan

2017-07-01

This article proposes an efficient way of calculating the geometric measure of entanglement using tensor decomposition methods. The connection between these two concepts is explored using the tensor representation of the wavefunction. Numerical examples are benchmarked and compared. Furthermore, we search for highly entangled qubit states to show the applicability of this method.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Ellis, RJ

2001-06-01

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

Development of a method to accurately calculate the Dpb and quickly predict the strength of a chemical bond

International Nuclear Information System (INIS)

Du, Xia; Zhao, Dong-Xia; Yang, Zhong-Zhi

2013-01-01

Highlights: ► A method from new respect to characterize and measure the bond strength is proposed. ► We calculate the D pb of a series of various bonds to justify our approach. ► A quite good linear relationship of the D pb with the bond lengths for series of various bonds is shown. ► Take the prediction of strengths of C–H and N–H bonds for base pairs in DNA as a practical application of our method. - Abstract: A new approach to characterize and measure bond strength has been developed. First, we propose a method to accurately calculate the potential acting on an electron in a molecule (PAEM) at the saddle point along a chemical bond in situ, denoted by D pb . Then, a direct method to quickly evaluate bond strength is established. We choose some familiar molecules as models for benchmarking this method. As a practical application, the D pb of base pairs in DNA along C–H and N–H bonds are obtained for the first time. All results show that C 7 –H of A–T and C 8 –H of G–C are the relatively weak bonds that are the injured positions in DNA damage. The significance of this work is twofold: (i) A method is developed to calculate D pb of various sizable molecules in situ quickly and accurately; (ii) This work demonstrates the feasibility to quickly predict the bond strength in macromolecules

Analysis on First Criticality Benchmark Calculation of HTR-10 Core

International Nuclear Information System (INIS)

Zuhair; Ferhat-Aziz; As-Natio-Lasman

2000-01-01

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

How Accurately can we Calculate Thermal Systems?

International Nuclear Information System (INIS)

Cullen, D; Blomquist, R N; Dean, C; Heinrichs, D; Kalugin, M A; Lee, M; Lee, Y; MacFarlan, R; Nagaya, Y; Trkov, A

2004-01-01

I would like to determine how accurately a variety of neutron transport code packages (code and cross section libraries) can calculate simple integral parameters, such as K eff , for systems that are sensitive to thermal neutron scattering. Since we will only consider theoretical systems, we cannot really determine absolute accuracy compared to any real system. Therefore rather than accuracy, it would be more precise to say that I would like to determine the spread in answers that we obtain from a variety of code packages. This spread should serve as an excellent indicator of how accurately we can really model and calculate such systems today. Hopefully, eventually this will lead to improvements in both our codes and the thermal scattering models that they use in the future. In order to accomplish this I propose a number of extremely simple systems that involve thermal neutron scattering that can be easily modeled and calculated by a variety of neutron transport codes. These are theoretical systems designed to emphasize the effects of thermal scattering, since that is what we are interested in studying. I have attempted to keep these systems very simple, and yet at the same time they include most, if not all, of the important thermal scattering effects encountered in a large, water-moderated, uranium fueled thermal system, i.e., our typical thermal reactors

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-07-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

a Proposed Benchmark Problem for Scatter Calculations in Radiographic Modelling

Science.gov (United States)

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

2009-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Krepper, E

1998-10-01

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

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

Analysis of the MZA/MZB benchmarks with modern nuclear data sets

International Nuclear Information System (INIS)

Rooijen, W.F.G. van

2013-01-01

Highlights: • ERANOS libraries are produced based on four modern nuclear data sets. • The MOZART MZA/MZB benchmarks are analyzed with these li- braries. • Results are generally acceptable in an academic context, but for highly accurate applications data adjustment is required. • Some discrepancies between the calculations and the benchmark results remain and cannot be readily explained. • Successful generation of ECCO libraries and covariance data for ERA- NOS. - Abstract: For fast reactor design and analysis, our laboratory uses, amongst others, the ERANOS code system. Unfortunately, the publicly available version of ERANOS does not have the most recent nuclear data. Therefore, it was decided to implement an integrated processing system to generate cross sections libraries for the ECCO cell code, as well as covariance data. Cross sections are generated from the original ENDF files. For our purposes, it is important to ascertain that the ECCO cross section libraries are of adequate quality to allow design and analysis of advanced fast reactors in an academic context. In this paper, we present an analysis of the MZA/MZB benchmarks with nuclear data from JENDL-4.0, JEFF-3.1.2 and ENDF/B-VII.1. Results are that reactivity is generally well predicted, with an uncertainty of about 1% due to covariances of the nuclear data. Reaction rate ratios are satisfactorily calculated, as well as the flux spectrum and reaction rate traverses. Some problems remain: the magnitude of the void effect is not satisfactorily calculated, and reaction rate traverses are not always satisfactorily calculated. On the whole, the ECCO libraries are sufficient for design and analysis tasks in an academic context. For high-precision calculations, such as required for licensing tasks and detailed design calculations, data adjustment is still necessary as the “native” covariance data in the ENDF files is not accurate enough

A benchmark test of computer codes for calculating average resonance parameters

International Nuclear Information System (INIS)

Ribon, P.; Thompson, A.

1983-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-09-01

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

Application of dynamic pseudo fission products and actinides for accurate burnup calculations

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L.

1996-09-01

The introduction of pseudo fission products for accurate fine-group spectrum calculations during burnup is discussed. The calculation of the density of the pseudo nuclides is done before each spectrum calculation from the actual densities and their cross sections of all nuclides to be lumped into a pseudo fission product. As there are also many actinides formed in the fuel during its life cycle, a pseudo actinide with fission cross section is also introduced. From a realistic burnup calculation it is demonstrated that only a few fission products and actinides need to be included explicitly in a spectrum calculation. All other fission products and actinides can be accurately represented in the pseudo nuclides. (author)

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Calculations of different transmutation concepts. An international benchmark exercise

International Nuclear Information System (INIS)

2000-01-01

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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

International Nuclear Information System (INIS)

Oh, I.; Rothe, R.E.

1978-01-01

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

Experimental Benchmarking of Fire Modeling Simulations. Final Report

International Nuclear Information System (INIS)

Greiner, Miles; Lopez, Carlos

2003-01-01

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

Fast and accurate calculation of dilute quantum gas using Uehling–Uhlenbeck model equation

Energy Technology Data Exchange (ETDEWEB)

Yano, Ryosuke, E-mail: ryosuke.yano@tokiorisk.co.jp

2017-02-01

The Uehling–Uhlenbeck (U–U) model equation is studied for the fast and accurate calculation of a dilute quantum gas. In particular, the direct simulation Monte Carlo (DSMC) method is used to solve the U–U model equation. DSMC analysis based on the U–U model equation is expected to enable the thermalization to be accurately obtained using a small number of sample particles and the dilute quantum gas dynamics to be calculated in a practical time. Finally, the applicability of DSMC analysis based on the U–U model equation to the fast and accurate calculation of a dilute quantum gas is confirmed by calculating the viscosity coefficient of a Bose gas on the basis of the Green–Kubo expression and the shock layer of a dilute Bose gas around a cylinder.

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

International Nuclear Information System (INIS)

Kim, Y.I.

2002-01-01

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

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

Science.gov (United States)

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

2014-05-01

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

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

Directory of Open Access Journals (Sweden)

Tanaka Ken-ichi

2016-01-01

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

Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

International Nuclear Information System (INIS)

Bowman, S.M.

1991-04-01

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

Detailed resonance absorption calculations with the Monte Carlo code MCNP and collision probability version of the slowing down code ROLAIDS

International Nuclear Information System (INIS)

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

1994-01-01

Very accurate Mote Carlo calculations with Monte Carlo Code have been performed to serve as reference for benchmark calculations on resonance absorption by U 238 in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (authors). 9 refs., 5 figs., 2 tabs

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-12-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-11-01

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

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

International Nuclear Information System (INIS)

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

1993-11-01

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

Interactions of model biomolecules. Benchmark CC calculations within MOLCAS

Energy Technology Data Exchange (ETDEWEB)

Urban, Miroslav [Slovak University of Technology in Bratislava, Faculty of Materials Science and Technology in Trnava, Institute of Materials Science, Bottova 25, SK-917 24 Trnava, Slovakia and Department of Physical and Theoretical Chemistry, Faculty of Natural Scie (Slovakia); Pitoňák, Michal; Neogrády, Pavel; Dedíková, Pavlína [Department of Physical and Theoretical Chemistry, Faculty of Natural Sciences, Comenius University, Mlynská dolina, SK-842 15 Bratislava (Slovakia); Hobza, Pavel [Institute of Organic Chemistry and Biochemistry and Center for Complex Molecular Systems and biomolecules, Academy of Sciences of the Czech Republic, Prague (Czech Republic)

2015-01-22

We present results using the OVOS approach (Optimized Virtual Orbitals Space) aimed at enhancing the effectiveness of the Coupled Cluster calculations. This approach allows to reduce the total computer time required for large-scale CCSD(T) calculations about ten times when the original full virtual space is reduced to about 50% of its original size without affecting the accuracy. The method is implemented in the MOLCAS computer program. When combined with the Cholesky decomposition of the two-electron integrals and suitable parallelization it allows calculations which were formerly prohibitively too demanding. We focused ourselves to accurate calculations of the hydrogen bonded and the stacking interactions of the model biomolecules. Interaction energies of the formaldehyde, formamide, benzene, and uracil dimers and the three-body contributions in the cytosine – guanine tetramer are presented. Other applications, as the electron affinity of the uracil affected by solvation are also shortly mentioned.

An Accurate Technique for Calculation of Radiation From Printed Reflectarrays

DEFF Research Database (Denmark)

Zhou, Min; Sorensen, Stig B.; Jorgensen, Erik

2011-01-01

The accuracy of various techniques for calculating the radiation from printed reflectarrays is examined, and an improved technique based on the equivalent currents approach is proposed. The equivalent currents are found from a continuous plane wave spectrum calculated by use of the spectral dyadic...... Green's function. This ensures a correct relation between the equivalent electric and magnetic currents and thus allows an accurate calculation of the radiation over the entire far-field sphere. A comparison to DTU-ESA Facility measurements of a reference offset reflectarray designed and manufactured...

Accurate quantum calculations of the reaction rates for H/D+ CH4

NARCIS (Netherlands)

Harrevelt, R. van; Nyman, G.; Manthe, U.

2007-01-01

In previous work [T. Wu, H. J. Werner, and U. Manthe, Science 306, 2227 (2004)], accurate quantum reaction rate calculations of the rate constant for the H+CH4 -> CH3+H-2 reaction have been presented. Both the electronic structure calculations and the nuclear dynamics calculations are converged with

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 Benchmark Computational Analysis

International Nuclear Information System (INIS)

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

2000-01-01

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

WWER-1000 Burnup Credit Benchmark (CB5)

International Nuclear Information System (INIS)

Manolova, M.A.

2002-01-01

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

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

International Nuclear Information System (INIS)

Svarny, J.; Mikolas, P.

1999-01-01

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

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

International Nuclear Information System (INIS)

Hordosy, G.; Maraczy, Cs.

2000-01-01

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

Benchmarks for GADRAS performance validation

International Nuclear Information System (INIS)

Mattingly, John K.; Mitchell, Dean James; Rhykerd, Charles L. Jr.

2009-01-01

The performance of the Gamma Detector Response and Analysis Software (GADRAS) was validated by comparing GADRAS model results to experimental measurements for a series of benchmark sources. Sources for the benchmark include a plutonium metal sphere, bare and shielded in polyethylene, plutonium oxide in cans, a highly enriched uranium sphere, bare and shielded in polyethylene, a depleted uranium shell and spheres, and a natural uranium sphere. The benchmark experimental data were previously acquired and consist of careful collection of background and calibration source spectra along with the source spectra. The calibration data were fit with GADRAS to determine response functions for the detector in each experiment. A one-dimensional model (pie chart) was constructed for each source based on the dimensions of the benchmark source. The GADRAS code made a forward calculation from each model to predict the radiation spectrum for the detector used in the benchmark experiment. The comparisons between the GADRAS calculation and the experimental measurements are excellent, validating that GADRAS can correctly predict the radiation spectra for these well-defined benchmark sources.

Analysis of an OECD/NEA high-temperature reactor benchmark

International Nuclear Information System (INIS)

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

2006-01-01

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

Core Benchmarks Descriptions

International Nuclear Information System (INIS)

Pavlovichev, A.M.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

A Heterogeneous Medium Analytical Benchmark

International Nuclear Information System (INIS)

Ganapol, B.D.

1999-01-01

A benchmark, called benchmark BLUE, has been developed for one-group neutral particle (neutron or photon) transport in a one-dimensional sub-critical heterogeneous plane parallel medium with surface illumination. General anisotropic scattering is accommodated through the Green's Function Method (GFM). Numerical Fourier transform inversion is used to generate the required Green's functions which are kernels to coupled integral equations that give the exiting angular fluxes. The interior scalar flux is then obtained through quadrature. A compound iterative procedure for quadrature order and slab surface source convergence provides highly accurate benchmark qualities (4- to 5- places of accuracy) results

Impact of the 235U Covariance Data in Benchmark Calculations

International Nuclear Information System (INIS)

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

2008-01-01

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

Impact of the 235U covariance data in benchmark calculations

International Nuclear Information System (INIS)

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

2008-01-01

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

Synthesis of the OECD/NEA-PSI CFD benchmark exercise

Energy Technology Data Exchange (ETDEWEB)

Andreani, Michele, E-mail: Michele.andreani@psi.ch; Badillo, Arnoldo; Kapulla, Ralf

2016-04-01

Highlights: • A benchmark exercise on stratification erosion in containment was conducted using a test in the PANDA facility. • Blind calculations were provided by nineteen participants. • Results were compared with experimental data. • A ranking was made. • A large spread of results was observed, with very few simulations providing accurate results for the most important variables, though not for velocities. - Abstract: The third International Benchmark Exercise (IBE-3) conducted under the auspices of OECD/NEA is based on the comparison of blind CFD simulations with experimental data addressing the erosion of a stratified layer by an off-axis buoyant jet in a large vessel. The numerical benchmark exercise is based on a dedicated experiment in the PANDA facility conducted at the Paul Scherrer Institut (PSI) in Switzerland, using only one vessel. The use of non-prototypical fluids (i.e. helium as simulant for hydrogen, and air as simulant for steam), and the consequent absence of the complex physical effects produced by steam condensation enhanced the suitability of the data for CFD validation purposes. The test started with a helium–air layer at the top of the vessel and air in the lower part. The helium-rich layer was gradually eroded by a low-momentum air/helium jet emerging at a lower elevation. Blind calculation results were submitted by nineteen participants, and the calculation results have been compared with the PANDA data. This report, adopting the format of the reports for the two previous exercises, includes a ranking of the contributions, where the largest weight is given to the time progression of the erosion of the helium-rich layer. In accordance with the limited scope of the benchmark exercise, this report is more a collection of comparisons between calculated results and data than a synthesis. Therefore, the few conclusions are based on the mere observation of the agreement of the various submissions with the test result, and do not

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

International Nuclear Information System (INIS)

Mosteller, R.D.

1997-01-01

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

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

International Nuclear Information System (INIS)

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

1995-09-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Accurate alpha sticking fractions from improved calculations relevant for muon catalyzed fusion

International Nuclear Information System (INIS)

Szalewicz, K.

1990-05-01

Recent experiments have shown that under proper conditions a single muon may catalyze almost two hundred fusions in its lifetime. This process proceeds through formation of muonic molecular ions. Properties of these ions are central to the understanding of the phenomenon. Our work included the most accurate calculations of the energy levels and Coulombic sticking fractions for tdμ and other muonic molecular ions, calculations of Auger transition rates, calculations of corrections to the energy levels due to interactions with the most molecule, and calculation of the reactivation of muons from α particles. The majority of our effort has been devoted to the theory and computation of the influence of the strong nuclear forces on fusion rates and sticking fractions. We have calculated fusion rates for tdμ including the effects of nuclear forces on the molecular wave functions. We have also shown that these results can be reproduced to almost four digit accuracy by using a very simple quasifactorizable expression which does not require modifications of the molecular wave functions. Our sticking fractions are more accurate than any other theoretical values. We have used a more sophisticated theory than any other work and our numerical calculations have converged to at least three significant digits

Criticality safety benchmarking of PASC-3 and ECNJEF1.1

International Nuclear Information System (INIS)

Li, J.

1992-09-01

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

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.

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.

Application of the random vibration approach in the seismic analysis of LMFBR structures - Benchmark calculations

International Nuclear Information System (INIS)

Preumont, A.; Shilab, S.; Cornaggia, L.; Reale, M.; Labbe, P.; Noe, H.

1992-01-01

This benchmark exercise is the continuation of the state-of-the-art review (EUR 11369 EN) which concluded that the random vibration approach could be an effective tool in seismic analysis of nuclear power plants, with potential advantages on time history and response spectrum techniques. As compared to the latter, the random vibration method provides an accurate treatment of multisupport excitations, non classical damping as well as the combination of high-frequency modal components. With respect to the former, the random vibration method offers direct information on statistical variability (probability distribution) and cheaper computations. The disadvantages of the random vibration method are that it is based on stationary results, and requires a power spectral density input instead of a response spectrum. A benchmark exercise to compare the three methods from the various aspects mentioned above, on one or several simple structures has been made. The following aspects have been covered with the simplest possible models: (i) statistical variability, (ii) multisupport excitation, (iii) non-classical damping. The random vibration method is therefore concluded to be a reliable method of analysis. Its use is recommended, particularly for preliminary design, owing to its computational advantage on multiple time history analysis

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

DEFF Research Database (Denmark)

Grandjean, Philippe; Budtz-Joergensen, Esben

2013-01-01

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

[Do you mean benchmarking?].

Science.gov (United States)

Bonnet, F; Solignac, S; Marty, J

2008-03-01

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

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.

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

International Nuclear Information System (INIS)

Caldeira, A.D.

1987-05-01

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

Dynamic benchmarking of simulation codes

International Nuclear Information System (INIS)

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

1996-01-01

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

3-D neutron transport benchmarks

International Nuclear Information System (INIS)

Takeda, T.; Ikeda, H.

1991-03-01

A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of K eff , control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes

An accurate determination of the flux within a slab

International Nuclear Information System (INIS)

Ganapol, B.D.; Lapenta, G.

1993-01-01

During the past decade, several articles have been written concerning accurate solutions to the monoenergetic neutron transport equation in infinite and semi-infinite geometries. The numerical formulations found in these articles were based primarily on the extensive theoretical investigations performed by the open-quotes transport greatsclose quotes such as Chandrasekhar, Busbridge, Sobolev, and Ivanov, to name a few. The development of numerical solutions in infinite and semi-infinite geometries represents an example of how mathematical transport theory can be utilized to provide highly accurate and efficient numerical transport solutions. These solutions, or analytical benchmarks, are useful as open-quotes industry standards,close quotes which provide guidance to code developers and promote learning in the classroom. The high accuracy of these benchmarks is directly attributable to the rapid advancement of the state of computing and computational methods. Transport calculations that were beyond the capability of the open-quotes supercomputersclose quotes of just a few years ago are now possible at one's desk. In this paper, we again build upon the past to tackle the slab problem, which is of the next level of difficulty in comparison to infinite media problems. The formulation is based on the monoenergetic Green's function, which is the most fundamental transport solution. This method of solution requires a fast and accurate evaluation of the Green's function, which, with today's computational power, is now readily available

Reducing dose calculation time for accurate iterative IMRT planning

International Nuclear Information System (INIS)

Siebers, Jeffrey V.; Lauterbach, Marc; Tong, Shidong; Wu Qiuwen; Mohan, Radhe

2002-01-01

A time-consuming component of IMRT optimization is the dose computation required in each iteration for the evaluation of the objective function. Accurate superposition/convolution (SC) and Monte Carlo (MC) dose calculations are currently considered too time-consuming for iterative IMRT dose calculation. Thus, fast, but less accurate algorithms such as pencil beam (PB) algorithms are typically used in most current IMRT systems. This paper describes two hybrid methods that utilize the speed of fast PB algorithms yet achieve the accuracy of optimizing based upon SC algorithms via the application of dose correction matrices. In one method, the ratio method, an infrequently computed voxel-by-voxel dose ratio matrix (R=D SC /D PB ) is applied for each beam to the dose distributions calculated with the PB method during the optimization. That is, D PB xR is used for the dose calculation during the optimization. The optimization proceeds until both the IMRT beam intensities and the dose correction ratio matrix converge. In the second method, the correction method, a periodically computed voxel-by-voxel correction matrix for each beam, defined to be the difference between the SC and PB dose computations, is used to correct PB dose distributions. To validate the methods, IMRT treatment plans developed with the hybrid methods are compared with those obtained when the SC algorithm is used for all optimization iterations and with those obtained when PB-based optimization is followed by SC-based optimization. In the 12 patient cases studied, no clinically significant differences exist in the final treatment plans developed with each of the dose computation methodologies. However, the number of time-consuming SC iterations is reduced from 6-32 for pure SC optimization to four or less for the ratio matrix method and five or less for the correction method. Because the PB algorithm is faster at computing dose, this reduces the inverse planning optimization time for our implementation

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-15

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

Power reactor pressure vessel benchmarks

International Nuclear Information System (INIS)

Rahn, F.J.

1978-01-01

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

Accurate density-functional calculations on large systems: Fullerenes and magnetic clusters

International Nuclear Information System (INIS)

Dunlap, B.I.

1996-01-01

Efforts to accurately compute all-electron density-functional energies for large molecules and clusters using Gaussian basis sets will be reviewed. The foundation of this effort, variational fitting, will be described and followed by three applications of the method. The first application concerns fullerenes. When first discovered, C 60 is quite unstable relative to the higher fullerenes. In addition, to raising questions about the relative abundance of the various fullerenes, this work conflicted with the then state-of-the art density-funcitonal calculations on crystalline graphite. Now high accuracy molecular and band structure calculations are in fairly good agreement. Second, we have used these methods to design transition metal clusters having the highest magnetic moment by maximizing the symmetry-required degeneracy of the one-electron orbitals. Most recently, we have developed accurate, variational generalized-gradient approximation (GGA) forces for use in geometry optimization of clusters and in molecular-dynamics simulations of friction. The GGA optimized geometries of a number of large clusters will be given

Vver-1000 Mox core computational benchmark

International Nuclear Information System (INIS)

2006-01-01

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

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

Vibrational multiconfiguration self-consistent field theory: implementation and test calculations.

Science.gov (United States)

Heislbetz, Sandra; Rauhut, Guntram

2010-03-28

A state-specific vibrational multiconfiguration self-consistent field (VMCSCF) approach based on a multimode expansion of the potential energy surface is presented for the accurate calculation of anharmonic vibrational spectra. As a special case of this general approach vibrational complete active space self-consistent field calculations will be discussed. The latter method shows better convergence than the general VMCSCF approach and must be considered the preferred choice within the multiconfigurational framework. Benchmark calculations are provided for a small set of test molecules.

Pool critical assembly pressure vessel facility benchmark

International Nuclear Information System (INIS)

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

1997-07-01

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

Verification and validation benchmarks.

Energy Technology Data Exchange (ETDEWEB)

Oberkampf, William Louis; Trucano, Timothy Guy

2007-02-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Benchmark 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)

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Accelerator shielding benchmark problems

International Nuclear Information System (INIS)

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

1993-01-01

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

Shielding benchmark problems

International Nuclear Information System (INIS)

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

1978-09-01

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

Quantum Chemical Benchmarking, Validation, and Prediction of Acidity Constants for Substituted Pyridinium Ions and Pyridinyl Radicals.

Science.gov (United States)

Keith, John A; Carter, Emily A

2012-09-11

Sensibly modeling (photo)electrocatalytic reactions involving proton and electron transfer with computational quantum chemistry requires accurate descriptions of protonated, deprotonated, and radical species in solution. Procedures to do this are generally nontrivial, especially in cases that involve radical anions that are unstable in the gas phase. Recently, pyridinium and the corresponding reduced neutral radical have been postulated as key catalysts in the reduction of CO2 to methanol. To assess practical methodologies to describe the acid/base chemistry of these species, we employed density functional theory (DFT) in tandem with implicit solvation models to calculate acidity constants for 22 substituted pyridinium cations and their corresponding pyridinyl radicals in water solvent. We first benchmarked our calculations against experimental pyridinium deprotonation energies in both gas and aqueous phases. DFT with hybrid exchange-correlation functionals provide chemical accuracy for gas-phase data and allow absolute prediction of experimental pKas with unsigned errors under 1 pKa unit. The accuracy of this economical pKa calculation approach was further verified by benchmarking against highly accurate (but very expensive) CCSD(T)-F12 calculations. We compare the relative importance and sensitivity of these energies to selection of solvation model, solvation energy definitions, implicit solvation cavity definition, basis sets, electron densities, model geometries, and mixed implicit/explicit models. After determining the most accurate model to reproduce experimentally-known pKas from first principles, we apply the same approach to predict pKas for radical pyridinyl species that have been proposed relevant under electrochemical conditions. This work provides considerable insight into the pitfalls using continuum solvation models, particularly when used for radical species.

Accurate quasiparticle calculation of x-ray photoelectron spectra of solids.

Science.gov (United States)

Aoki, Tsubasa; Ohno, Kaoru

2018-05-31

It has been highly desired to provide an accurate and reliable method to calculate core electron binding energies (CEBEs) of crystals and to understand the final state screening effect on a core hole in high resolution x-ray photoelectron spectroscopy (XPS), because the ΔSCF method cannot be simply used for bulk systems. We propose to use the quasiparticle calculation based on many-body perturbation theory for this problem. In this study, CEBEs of band-gapped crystals, silicon, diamond, β-SiC, BN, and AlP, are investigated by means of the GW approximation (GWA) using the full ω integration and compared with the preexisting XPS data. The screening effect on a deep core hole is also investigated in detail by evaluating the relaxation energy (RE) from the core and valence contributions separately. Calculated results show that not only the valence electrons but also the core electrons have an important contribution to the RE, and the GWA have a tendency to underestimate CEBEs due to the excess RE. This underestimation can be improved by introducing the self-screening correction to the GWA. The resulting C1s, B1s, N1s, Si2p, and Al2p CEBEs are in excellent agreement with the experiments within 1 eV absolute error range. The present self-screening corrected GW approach has the capability to achieve the highly accurate prediction of CEBEs without any empirical parameter for band-gapped crystals, and provide a more reliable theoretical approach than the conventional ΔSCF-DFT method.

Accurate quasiparticle calculation of x-ray photoelectron spectra of solids

Science.gov (United States)

Aoki, Tsubasa; Ohno, Kaoru

2018-05-01

It has been highly desired to provide an accurate and reliable method to calculate core electron binding energies (CEBEs) of crystals and to understand the final state screening effect on a core hole in high resolution x-ray photoelectron spectroscopy (XPS), because the ΔSCF method cannot be simply used for bulk systems. We propose to use the quasiparticle calculation based on many-body perturbation theory for this problem. In this study, CEBEs of band-gapped crystals, silicon, diamond, β-SiC, BN, and AlP, are investigated by means of the GW approximation (GWA) using the full ω integration and compared with the preexisting XPS data. The screening effect on a deep core hole is also investigated in detail by evaluating the relaxation energy (RE) from the core and valence contributions separately. Calculated results show that not only the valence electrons but also the core electrons have an important contribution to the RE, and the GWA have a tendency to underestimate CEBEs due to the excess RE. This underestimation can be improved by introducing the self-screening correction to the GWA. The resulting C1s, B1s, N1s, Si2p, and Al2p CEBEs are in excellent agreement with the experiments within 1 eV absolute error range. The present self-screening corrected GW approach has the capability to achieve the highly accurate prediction of CEBEs without any empirical parameter for band-gapped crystals, and provide a more reliable theoretical approach than the conventional ΔSCF-DFT method.

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-01

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

Accurate calculations of bound rovibrational states for argon trimer

Energy Technology Data Exchange (ETDEWEB)

Brandon, Drew; Poirier, Bill [Department of Chemistry and Biochemistry, and Department of Physics, Texas Tech University, Box 41061, Lubbock, Texas 79409-1061 (United States)

2014-07-21

This work presents a comprehensive quantum dynamics calculation of the bound rovibrational eigenstates of argon trimer (Ar{sub 3}), using the ScalIT suite of parallel codes. The Ar{sub 3} rovibrational energy levels are computed to a very high level of accuracy (10{sup −3} cm{sup −1} or better), and up to the highest rotational and vibrational excitations for which bound states exist. For many of these rovibrational states, wavefunctions are also computed. Rare gas clusters such as Ar{sub 3} are interesting because the interatomic interactions manifest through long-range van der Waals forces, rather than through covalent chemical bonding. As a consequence, they exhibit strong Coriolis coupling between the rotational and vibrational degrees of freedom, as well as highly delocalized states, all of which renders accurate quantum dynamical calculation difficult. Moreover, with its (comparatively) deep potential well and heavy masses, Ar{sub 3} is an especially challenging rare gas trimer case. There are a great many rovibrational eigenstates to compute, and a very high density of states. Consequently, very few previous rovibrational state calculations for Ar{sub 3} may be found in the current literature—and only for the lowest-lying rotational excitations.

Shielding benchmark problems, (2)

International Nuclear Information System (INIS)

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

1980-02-01

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

RISKIND verification and benchmark comparisons

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

RISKIND verification and benchmark comparisons

International Nuclear Information System (INIS)

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

1997-08-01

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

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Science.gov (United States)

Veis, Libor; Pittner, Jiří

2010-11-21

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

Verification and validation benchmarks

International Nuclear Information System (INIS)

Oberkampf, William Louis; Trucano, Timothy Guy

2007-01-01

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

Verification and validation benchmarks

International Nuclear Information System (INIS)

Oberkampf, William L.; Trucano, Timothy G.

2008-01-01

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

An efficient and accurate method for calculating nonlinear diffraction beam fields

Energy Technology Data Exchange (ETDEWEB)

Jeong, Hyun Jo; Cho, Sung Jong; Nam, Ki Woong; Lee, Jang Hyun [Division of Mechanical and Automotive Engineering, Wonkwang University, Iksan (Korea, Republic of)

2016-04-15

This study develops an efficient and accurate method for calculating nonlinear diffraction beam fields propagating in fluids or solids. The Westervelt equation and quasilinear theory, from which the integral solutions for the fundamental and second harmonics can be obtained, are first considered. A computationally efficient method is then developed using a multi-Gaussian beam (MGB) model that easily separates the diffraction effects from the plane wave solution. The MGB models provide accurate beam fields when compared with the integral solutions for a number of transmitter-receiver geometries. These models can also serve as fast, powerful modeling tools for many nonlinear acoustics applications, especially in making diffraction corrections for the nonlinearity parameter determination, because of their computational efficiency and accuracy.

Monte Carlo benchmarking: Validation and progress

International Nuclear Information System (INIS)

Sala, P.

2010-01-01

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

PHEBUS-FPTO Benchmark calculations

International Nuclear Information System (INIS)

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

1991-01-01

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

Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

International Nuclear Information System (INIS)

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

2014-01-01

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

Five- and six-electron harmonium atoms: Highly accurate electronic properties and their application to benchmarking of approximate 1-matrix functionals

Science.gov (United States)

Cioslowski, Jerzy; Strasburger, Krzysztof

2018-04-01

Electronic properties of several states of the five- and six-electron harmonium atoms are obtained from large-scale calculations employing explicitly correlated basis functions. The high accuracy of the computed energies (including their components), natural spinorbitals, and their occupation numbers makes them suitable for testing, calibration, and benchmarking of approximate formalisms of quantum chemistry and solid state physics. In the case of the five-electron species, the availability of the new data for a wide range of the confinement strengths ω allows for confirmation and generalization of the previously reached conclusions concerning the performance of the presently known approximations for the electron-electron repulsion energy in terms of the 1-matrix that are at heart of the density matrix functional theory (DMFT). On the other hand, the properties of the three low-lying states of the six-electron harmonium atom, computed at ω = 500 and ω = 1000, uncover deficiencies of the 1-matrix functionals not revealed by previous studies. In general, the previously published assessment of the present implementations of DMFT being of poor accuracy is found to hold. Extending the present work to harmonically confined systems with even more electrons is most likely counterproductive as the steep increase in computational cost required to maintain sufficient accuracy of the calculated properties is not expected to be matched by the benefits of additional information gathered from the resulting benchmarks.

Application of an accurate thermal hydraulics solver in VTT's reactor dynamics codes

International Nuclear Information System (INIS)

Rajamaeki, M.; Raety, H.; Kyrki-Rajamaeki, R.; Eskola, M.

1998-01-01

VTT's reactor dynamics codes are developed further and new more detailed models are created for tasks related to increased safety requirements. For thermal hydraulics calculations an accurate general flow model based on a new solution method PLIM has been developed. It has been applied in VTT's one-dimensional TRAB and three-dimensional HEXTRAN codes. Results of a demanding international boron dilution benchmark defined by VTT are given and compared against results of other codes with original or improved boron tracking. The new PLIM method not only allows the accurate modelling of a propagating boron dilution front, but also the tracking of a temperature front, which is missed by the special boron tracking models. (orig.)

Accurate Calculation of Fringe Fields in the LHC Main Dipoles

CERN Document Server

Kurz, S; Siegel, N

2000-01-01

The ROXIE program developed at CERN for the design and optimization of the superconducting LHC magnets has been recently extended in a collaboration with the University of Stuttgart, Germany, with a field computation method based on the coupling between the boundary element (BEM) and the finite element (FEM) technique. This avoids the meshing of the coils and the air regions, and avoids the artificial far field boundary conditions. The method is therefore specially suited for the accurate calculation of fields in the superconducting magnets in which the field is dominated by the coil. We will present the fringe field calculations in both 2d and 3d geometries to evaluate the effect of connections and the cryostat on the field quality and the flux density to which auxiliary bus-bars are exposed.

Revaluering benchmarking - A topical theme for the construction industry

DEFF Research Database (Denmark)

Rasmussen, Grane Mikael Gregaard

2011-01-01

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

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

International Nuclear Information System (INIS)

Yoshizawa, Nobuaki; Meigo, Shin-ichiro

2001-01-01

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

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

International Nuclear Information System (INIS)

Svarny, J.; Mikolas, P.

1999-01-01

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

KAERI results for BN600 full MOX benchmark (Phase 4)

International Nuclear Information System (INIS)

Lee, Kibog Lee

2003-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-06-01

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

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

International Nuclear Information System (INIS)

Li, M; Chetty, I; Zhong, H

2014-01-01

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

MCNP simulation of the TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

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

1996-01-01

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

Improved Patient Size Estimates for Accurate Dose Calculations in Abdomen Computed Tomography

Energy Technology Data Exchange (ETDEWEB)

Lee, Chang-Lae [Yonsei University, Wonju (Korea, Republic of)

2017-07-15

The radiation dose of CT (computed tomography) is generally represented by the CTDI (CT dose index). CTDI, however, does not accurately predict the actual patient doses for different human body sizes because it relies on a cylinder-shaped head (diameter : 16 cm) and body (diameter : 32 cm) phantom. The purpose of this study was to eliminate the drawbacks of the conventional CTDI and to provide more accurate radiation dose information. Projection radiographs were obtained from water cylinder phantoms of various sizes, and the sizes of the water cylinder phantoms were calculated and verified using attenuation profiles. The effective diameter was also calculated using the attenuation of the abdominal projection radiographs of 10 patients. When the results of the attenuation-based method and the geometry-based method shown were compared with the results of the reconstructed-axial-CT-image-based method, the effective diameter of the attenuation-based method was found to be similar to the effective diameter of the reconstructed-axial-CT-image-based method, with a difference of less than 3.8%, but the geometry-based method showed a difference of less than 11.4%. This paper proposes a new method of accurately computing the radiation dose of CT based on the patient sizes. This method computes and provides the exact patient dose before the CT scan, and can therefore be effectively used for imaging and dose control.

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...

Criticality Benchmark Results Using Various MCNP Data Libraries

International Nuclear Information System (INIS)

Frankle, Stephanie C.

1999-01-01

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

Reactor group constants and benchmark test

Energy Technology Data Exchange (ETDEWEB)

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

2001-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Primm III, RT

2002-05-29

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

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

International Nuclear Information System (INIS)

Primm III, RT

2002-01-01

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

An Effective Method to Accurately Calculate the Phase Space Factors for β"-β"- Decay

International Nuclear Information System (INIS)

Horoi, Mihai; Neacsu, Andrei

2016-01-01

Accurate calculations of the electron phase space factors are necessary for reliable predictions of double-beta decay rates and for the analysis of the associated electron angular and energy distributions. We present an effective method to calculate these phase space factors that takes into account the distorted Coulomb field of the daughter nucleus, yet it allows one to easily calculate the phase space factors with good accuracy relative to the most exact methods available in the recent literature.

VENUS-2 Benchmark Problem Analysis with HELIOS-1.9

International Nuclear Information System (INIS)

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

2014-01-01

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

Analysis of a molten salt reactor benchmark

International Nuclear Information System (INIS)

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

2013-01-01

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

Fast and accurate calculation of the properties of water and steam for simulation

International Nuclear Information System (INIS)

Szegi, Zs.; Gacs, A.

1990-01-01

A basic principle simulator was developed at the CRIP, Budapest, for real time simulation of the transients of WWER-440 type nuclear power plants. Its integral part is the fast and accurate calculation of the thermodynamic properties of water and steam. To eliminate successive approximations, the model system of the secondary coolant circuit requires binary forms which are known as inverse functions, countinuous when crossing the saturation line, accurate and coherent for all argument combinations. A solution which reduces the computer memory and execution time demand is reported. (author) 36 refs.; 5 figs.; 3 tabs

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)

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

International Nuclear Information System (INIS)

2000-01-01

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

Direct Calculation of Permeability by High-Accurate Finite Difference and Numerical Integration Methods

KAUST Repository

Wang, Yi

2016-07-21

Velocity of fluid flow in underground porous media is 6~12 orders of magnitudes lower than that in pipelines. If numerical errors are not carefully controlled in this kind of simulations, high distortion of the final results may occur [1-4]. To fit the high accuracy demands of fluid flow simulations in porous media, traditional finite difference methods and numerical integration methods are discussed and corresponding high-accurate methods are developed. When applied to the direct calculation of full-tensor permeability for underground flow, the high-accurate finite difference method is confirmed to have numerical error as low as 10-5% while the high-accurate numerical integration method has numerical error around 0%. Thus, the approach combining the high-accurate finite difference and numerical integration methods is a reliable way to efficiently determine the characteristics of general full-tensor permeability such as maximum and minimum permeability components, principal direction and anisotropic ratio. Copyright © Global-Science Press 2016.

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

International Nuclear Information System (INIS)

Kloosterman, Jan Leen

1999-01-01

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

Procedures of grasp92 code to calculate accurate Dirac-Coulomb energy for the ground sate of helium atom

International Nuclear Information System (INIS)

Utsumi, Takayuki; Sasaki, Akira

2000-02-01

The procedures of grasp92 code to calculate accurate (relative error nearly equal 10 -7 ) eigenvalue for the ground sate of helium atom of the Dirac-Coulomb Hamiltonian are presented. The grasp92 code, based on the multi-configuration Dirac-Fock method, is widely used to calculate the atomic properties. However, the main part of the accurate calculations, extended optimal level calculation (EOL), suffer frequently numerical instabilities due to the lack of the confident procedures. The purpose of this report is to illustrate the guideline for stable EOL calculations by calculating the most fundamental atomic system, i.e. the ground sate of helium atom ls 2 1 S 2 . This procedure could be extended for the high-precise eigenfunction calculation of more complex atomic systems, for example highly ionized atoms and high-Z atoms. (author)

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.

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Accurate reactivity void coefficient calculation for the fast spectrum reactor FBR-IME

Energy Technology Data Exchange (ETDEWEB)

Lima, Fabiano P.C.; Vellozo, Sergio de O.; Velozo, Marta J., E-mail: fabianopetruceli@outlook.com, E-mail: vellozo@cbpf.br, E-mail: martajann@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Militar

2017-07-01

This paper aims to present an accurate calculation of the void reactivity coefficient for the FBR-IME, a fast spectrum reactor in development at the Engineering Military Institute (IME). The main design peculiarity lies in using mixed oxide [MOX - PuO{sub 2} + U(natural uranium)O{sub 2}] as fuel core. For this task, SCALE system was used to calculate the reactivity for several voids distributions generated by bubbles in the sodium beyond its boiling point. The results show that although the void reactivity coefficient is positive and location dependent, they are offset by other feedback effects, resulting in a negative overall coefficient. (author)

Handbook of critical experiments benchmarks

International Nuclear Information System (INIS)

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

1978-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-15

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

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

Science.gov (United States)

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

Calculation of accurate small angle X-ray scattering curves from coarse-grained protein models

Directory of Open Access Journals (Sweden)

Stovgaard Kasper

2010-08-01

Full Text Available Abstract Background Genome sequencing projects have expanded the gap between the amount of known protein sequences and structures. The limitations of current high resolution structure determination methods make it unlikely that this gap will disappear in the near future. Small angle X-ray scattering (SAXS is an established low resolution method for routinely determining the structure of proteins in solution. The purpose of this study is to develop a method for the efficient calculation of accurate SAXS curves from coarse-grained protein models. Such a method can for example be used to construct a likelihood function, which is paramount for structure determination based on statistical inference. Results We present a method for the efficient calculation of accurate SAXS curves based on the Debye formula and a set of scattering form factors for dummy atom representations of amino acids. Such a method avoids the computationally costly iteration over all atoms. We estimated the form factors using generated data from a set of high quality protein structures. No ad hoc scaling or correction factors are applied in the calculation of the curves. Two coarse-grained representations of protein structure were investigated; two scattering bodies per amino acid led to significantly better results than a single scattering body. Conclusion We show that the obtained point estimates allow the calculation of accurate SAXS curves from coarse-grained protein models. The resulting curves are on par with the current state-of-the-art program CRYSOL, which requires full atomic detail. Our method was also comparable to CRYSOL in recognizing native structures among native-like decoys. As a proof-of-concept, we combined the coarse-grained Debye calculation with a previously described probabilistic model of protein structure, TorusDBN. This resulted in a significant improvement in the decoy recognition performance. In conclusion, the presented method shows great promise for

Method to Calculate Accurate Top Event Probability in a Seismic PSA

Energy Technology Data Exchange (ETDEWEB)

Jung, Woo Sik [Sejong Univ., Seoul (Korea, Republic of)

2014-05-15

ACUBE(Advanced Cutset Upper Bound Estimator) calculates the top event probability and importance measures from cutsets by dividing cutsets into major and minor groups depending on the cutset probability, where the cutsets that have higher cutset probability are included in the major group and the others in minor cutsets, converting major cutsets into a Binary Decision Diagram (BDD). By applying the ACUBE algorithm to the seismic PSA cutsets, the accuracy of a top event probability and importance measures can be significantly improved. ACUBE works by dividing the cutsets into two groups (higher and lower cutset probability groups), calculating the top event probability and importance measures in each group, and combining the two results from the two groups. Here, ACUBE calculates the top event probability and importance measures of the higher cutset probability group exactly. On the other hand, ACUBE calculates these measures of the lower cutset probability group with an approximation such as MCUB. The ACUBE algorithm is useful for decreasing the conservatism that is caused by approximating the top event probability and importance measure calculations with given cutsets. By applying the ACUBE algorithm to the seismic PSA cutsets, the accuracy of a top event probability and importance measures can be significantly improved. This study shows that careful attention should be paid and an appropriate method be provided in order to avoid the significant overestimation of the top event probability calculation. Due to the strength of ACUBE that is explained in this study, the ACUBE became a vital tool for calculating more accurate CDF of the seismic PSA cutsets than the conventional probability calculation method.

Status on benchmark testing of CENDL-3

CERN Document Server

Liu Ping

2002-01-01

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

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

International Nuclear Information System (INIS)

Bencik, M.; Hadek, J.

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Leal, L.C.

1993-01-01

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

OECD/NEA expert group on uncertainty analysis for criticality safety assessment: Results of benchmark on sensitivity calculation (phase III)

Energy Technology Data Exchange (ETDEWEB)

Ivanova, T.; Laville, C. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay aux Roses (France); Dyrda, J. [Atomic Weapons Establishment AWE, Aldermaston, Reading, RG7 4PR (United Kingdom); Mennerdahl, D. [E Mennerdahl Systems EMS, Starvaegen 12, 18357 Taeby (Sweden); Golovko, Y.; Raskach, K.; Tsiboulia, A. [Inst. for Physics and Power Engineering IPPE, 1, Bondarenko sq., 249033 Obninsk (Russian Federation); Lee, G. S.; Woo, S. W. [Korea Inst. of Nuclear Safety KINS, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Bidaud, A.; Sabouri, P. [Laboratoire de Physique Subatomique et de Cosmologie LPSC, CNRS-IN2P3/UJF/INPG, Grenoble (France); Patel, A. [U.S. Nuclear Regulatory Commission (NRC), Washington, DC 20555-0001 (United States); Bledsoe, K.; Rearden, B. [Oak Ridge National Laboratory ORNL, M.S. 6170, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Gulliford, J.; Michel-Sendis, F. [OECD/NEA, 12, Bd des Iles, 92130 Issy-les-Moulineaux (France)

2012-07-01

The sensitivities of the k{sub eff} eigenvalue to neutron cross sections have become commonly used in similarity studies and as part of the validation algorithm for criticality safety assessments. To test calculations of the sensitivity coefficients, a benchmark study (Phase III) has been established by the OECD-NEA/WPNCS/EG UACSA (Expert Group on Uncertainty Analysis for Criticality Safety Assessment). This paper presents some sensitivity results generated by the benchmark participants using various computational tools based upon different computational methods: SCALE/TSUNAMI-3D and -1D, MONK, APOLLO2-MORET 5, DRAGON-SUSD3D and MMKKENO. The study demonstrates the performance of the tools. It also illustrates how model simplifications impact the sensitivity results and demonstrates the importance of 'implicit' (self-shielding) sensitivities. This work has been a useful step towards verification of the existing and developed sensitivity analysis methods. (authors)

The FLUKA code: An accurate simulation tool for particle therapy

CERN Document Server

Battistoni, Giuseppe; Böhlen, Till T; Cerutti, Francesco; Chin, Mary Pik Wai; Dos Santos Augusto, Ricardo M; Ferrari, Alfredo; Garcia Ortega, Pablo; Kozlowska, Wioletta S; Magro, Giuseppe; Mairani, Andrea; Parodi, Katia; Sala, Paola R; Schoofs, Philippe; Tessonnier, Thomas; Vlachoudis, Vasilis

2016-01-01

Monte Carlo (MC) codes are increasingly spreading in the hadrontherapy community due to their detailed description of radiation transport and interaction with matter. The suitability of a MC code for application to hadrontherapy demands accurate and reliable physical models capable of handling all components of the expected radiation field. This becomes extremely important for correctly performing not only physical but also biologically-based dose calculations, especially in cases where ions heavier than protons are involved. In addition, accurate prediction of emerging secondary radiation is of utmost importance in innovative areas of research aiming at in-vivo treatment verification. This contribution will address the recent developments of the FLUKA MC code and its practical applications in this field. Refinements of the FLUKA nuclear models in the therapeutic energy interval lead to an improved description of the mixed radiation field as shown in the presented benchmarks against experimental data with bot...

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

International Nuclear Information System (INIS)

Liu Guisheng

1999-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

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

International Nuclear Information System (INIS)

Panayotov, Dobromir

2008-01-01

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

CDCC calculations with the Lagrange-mesh technique

International Nuclear Information System (INIS)

Druet, T.; Baye, D.; Descouvemont, P.; Sparenberg, J.-M.

2010-01-01

We apply the Lagrange-mesh technique to the Continuum Discretized Coupled Channel (CDCC) theory. The CDCC equations are solved with the R-matrix method, using Lagrange functions as variational basis. The choice of Lagrange functions is shown to be efficient and accurate for elastic scattering as well as for breakup reactions. We describe the general formalism for two-body projectiles, and apply it to the d+ 58 Ni collision at E d =80 MeV. Various numerical and physical aspects are discussed. Benchmark calculations on elastic scattering and breakup are presented.

Development of common user data model for APOLLO3 and MARBLE and application to benchmark problems

International Nuclear Information System (INIS)

Yokoyama, Kenji

2009-07-01

A Common User Data Model, CUDM, has been developed for the purpose of benchmark calculations between APOLLO3 and MARBLE code systems. The current version of CUDM was designed for core calculation benchmark problems with 3-dimensional Cartesian, 3-D XYZ, geometry. CUDM is able to manage all input/output data such as 3-D XYZ geometry, effective macroscopic cross section, effective multiplication factor and neutron flux. In addition, visualization tools for geometry and neutron flux were included. CUDM was designed by the object-oriented technique and implemented using Python programming language. Based on the CUDM, a prototype system for a benchmark calculation, CUDM-benchmark, was also developed. The CUDM-benchmark supports input/output data conversion for IDT solver in APOLLO3, and TRITAC and SNT solvers in MARBLE. In order to evaluate pertinence of CUDM, the CUDM-benchmark was applied to benchmark problems proposed by T. Takeda, G. Chiba and I. Zmijarevic. It was verified that the CUDM-benchmark successfully reproduced the results calculated with reference input data files, and provided consistent results among all the solvers by using one common input data defined by CUDM. In addition, a detailed benchmark calculation for Chiba benchmark was performed by using the CUDM-benchmark. Chiba benchmark is a neutron transport benchmark problem for fast criticality assembly without homogenization. This benchmark problem consists of 4 core configurations which have different sodium void regions, and each core configuration is defined by more than 5,000 fuel/material cells. In this application, it was found that the results by IDT and SNT solvers agreed well with the reference results by Monte-Carlo code. In addition, model effects such as quadrature set effect, S n order effect and mesh size effect were systematically evaluated and summarized in this report. (author)

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

International Nuclear Information System (INIS)

Ishikawa, M.

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

REVISED STREAM CODE AND WASP5 BENCHMARK

International Nuclear Information System (INIS)

Chen, K

2005-01-01

STREAM is an emergency response code that predicts downstream pollutant concentrations for releases from the SRS area to the Savannah River. The STREAM code uses an algebraic equation to approximate the solution of the one dimensional advective transport differential equation. This approach generates spurious oscillations in the concentration profile when modeling long duration releases. To improve the capability of the STREAM code to model long-term releases, its calculation module was replaced by the WASP5 code. WASP5 is a US EPA water quality analysis program that simulates one-dimensional pollutant transport through surface water. Test cases were performed to compare the revised version of STREAM with the existing version. For continuous releases, results predicted by the revised STREAM code agree with physical expectations. The WASP5 code was benchmarked with the US EPA 1990 and 1991 dye tracer studies, in which the transport of the dye was measured from its release at the New Savannah Bluff Lock and Dam downstream to Savannah. The peak concentrations predicted by the WASP5 agreed with the measurements within ±20.0%. The transport times of the dye concentration peak predicted by the WASP5 agreed with the measurements within ±3.6%. These benchmarking results demonstrate that STREAM should be capable of accurately modeling releases from SRS outfalls

Benchmarking and Performance Management

Directory of Open Access Journals (Sweden)

Adrian TANTAU

2010-12-01

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

FENDL-3 benchmark test with neutronics experiments related to fusion in Japan

International Nuclear Information System (INIS)

Konno, Chikara; Ohta, Masayuki; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi

2014-01-01

Highlights: •We have benchmarked FENDL-3.0 with integral experiments with DT neutron sources in Japan. •The FENDL-3.0 is as accurate as FENDL-2.1 and JENDL-4.0 or more. •Some data in FENDL-3.0 may have some problems. -- Abstract: The IAEA supports and promotes the gathering of the best data from evaluated nuclear data libraries for each nucleus involved in fusion reactor applications and compiles these data as FENDL. In 2012, the IAEA released a major update to FENDL, FENDL-3.0, which extends the neutron energy range from 20 MeV to greater than 60 MeV for 180 nuclei. We have benchmarked FENDL-3.0 versus in situ and TOF experiments using the DT neutron source at FNS at the JAEA and TOF experiments using the DT neutron source at OKTAVIAN at Osaka University in Japan. The Monte Carlo code MCNP-5 and the ACE file of FENDL-3.0 supplied from the IAEA were used for the calculations. The results were compared with measured ones and those obtained using the previous version, FENDL-2.1, and the latest version, JENDL-4.0. It is concluded that FENDL-3.0 is as accurate as or more so than FENDL-2.1 and JENDL-4.0, although some data in FENDL-3.0 may be problematic

Atomic Energy Research benchmark activity

International Nuclear Information System (INIS)

Makai, M.

1998-01-01

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

Present Status and Extensions of the Monte Carlo Performance Benchmark

Science.gov (United States)

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

2014-06-01

The NEA Monte Carlo Performance benchmark started in 2011 aiming to monitor over the years the abilities to perform a full-size Monte Carlo reactor core calculation with a detailed power production for each fuel pin with axial distribution. This paper gives an overview of the contributed results thus far. It shows that reaching a statistical accuracy of 1 % for most of the small fuel zones requires about 100 billion neutron histories. The efficiency of parallel execution of Monte Carlo codes on a large number of processor cores shows clear limitations for computer clusters with common type computer nodes. However, using true supercomputers the speedup of parallel calculations is increasing up to large numbers of processor cores. More experience is needed from calculations on true supercomputers using large numbers of processors in order to predict if the requested calculations can be done in a short time. As the specifications of the reactor geometry for this benchmark test are well suited for further investigations of full-core Monte Carlo calculations and a need is felt for testing other issues than its computational performance, proposals are presented for extending the benchmark to a suite of benchmark problems for evaluating fission source convergence for a system with a high dominance ratio, for coupling with thermal-hydraulics calculations to evaluate the use of different temperatures and coolant densities and to study the correctness and effectiveness of burnup calculations. Moreover, other contemporary proposals for a full-core calculation with realistic geometry and material composition will be discussed.

Present status and extensions of the Monte Carlo performance benchmark

International Nuclear Information System (INIS)

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

2013-01-01

The NEA Monte Carlo Performance benchmark started in 2011 aiming to monitor over the years the abilities to perform a full-size Monte Carlo reactor core calculation with a detailed power production for each fuel pin with axial distribution. This paper gives an overview of the contributed results thus far. It shows that reaching a statistical accuracy of 1 % for most of the small fuel zones requires about 100 billion neutron histories. The efficiency of parallel execution of Monte Carlo codes on a large number of processor cores shows clear limitations for computer clusters with common type computer nodes. However, using true supercomputers the speedup of parallel calculations is increasing up to large numbers of processor cores. More experience is needed from calculations on true supercomputers using large numbers of processors in order to predict if the requested calculations can be done in a short time. As the specifications of the reactor geometry for this benchmark test are well suited for further investigations of full-core Monte Carlo calculations and a need is felt for testing other issues than its computational performance, proposals are presented for extending the benchmark to a suite of benchmark problems for evaluating fission source convergence for a system with a high dominance ratio, for coupling with thermal-hydraulics calculations to evaluate the use of different temperatures and coolant densities and to study the correctness and effectiveness of burnup calculations. Moreover, other contemporary proposals for a full-core calculation with realistic geometry and material composition will be discussed. (authors)

Gamma ray benchmark on the spent fuel shipping cask TN 12

International Nuclear Information System (INIS)

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

1983-05-01

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

Accurate heterogeneous dose calculation for lung cancer patients without high‐resolution CT densities

Science.gov (United States)

Li, Jonathan G.; Liu, Chihray; Olivier, Kenneth R.; Dempsey, James F.

2009-01-01

The aim of this study was to investigate the relative accuracy of megavoltage photon‐beam dose calculations employing either five bulk densities or independent voxel densities determined by calibration of the CT Houndsfield number. Full‐resolution CT and bulk density treatment plans were generated for 70 lung or esophageal cancer tumors (66 cases) using a commercial treatment planning system with an adaptive convolution dose calculation algorithm (Pinnacle3, Philips Medicals Systems). Bulk densities were applied to segmented regions. Individual and population average densities were compared to the full‐resolution plan for each case. Monitor units were kept constant and no normalizations were employed. Dose volume histograms (DVH) and dose difference distributions were examined for all cases. The average densities of the segmented air, lung, fat, soft tissue, and bone for the entire set were found to be 0.14, 0.26, 0.89, 1.02, and 1.12 g/cm3, respectively. In all cases, the normal tissue DVH agreed to better than 2% in dose. In 62 of 70 DVHs of the planning target volume (PTV), agreement to better than 3% in dose was observed. Six cases demonstrated emphysema, one with bullous formations and one with a hiatus hernia having a large volume of gas. These required the additional assignment of density to the emphysemic lung and inflammatory changes to the lung, the regions of collapsed lung, the bullous formations, and the hernia gas. Bulk tissue density dose calculation provides an accurate method of heterogeneous dose calculation. However, patients with advanced emphysema may require high‐resolution CT studies for accurate treatment planning. PACS number: 87.53.Tf

ACCURATELY CALCULATING THE SOLAR ORIENTATION OF THE TIANGONG-2 ULTRAVIOLET FORWARD SPECTROMETER

Directory of Open Access Journals (Sweden)

Z. Liu

2018-04-01

Full Text Available The Ultraviolet Forward Spectrometer is a new type of spectrometer for monitoring the vertical distribution of atmospheric trace gases in the global middle atmosphere. It is on the TianGong-2 space laboratory, which was launched on 15 September 2016. The spectrometer uses a solar calibration mode to modify its irradiance. Accurately calculating the solar orientation is a prerequisite of spectral calibration for the Ultraviolet Forward Spectrometer. In this paper, a method of calculating the solar orientation is proposed according to the imaging geometric characteristics of the spectrometer. Firstly, the solar orientation in the horizontal rectangular coordinate system is calculated based on the solar declination angle algorithm proposed by Bourges and the solar hour angle algorithm proposed by Lamm. Then, the solar orientation in the sensor coordinate system is achieved through several coordinate system transforms. Finally, we calculate the solar orientation in the sensor coordinate system and evaluate its calculation accuracy using actual orbital data of TianGong-2. The results show that the accuracy is close to the simulation method with STK (Satellite Tool Kit, and the error is not more than 2 %. The algorithm we present does not need a lot of astronomical knowledge, but only needs some observation parameters provided by TianGong-2.

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

Calculation of accurate albedo boundary conditions for three-dimensional nodal diffusion codes by the method of characteristics

International Nuclear Information System (INIS)

Petkov, Petko T.

2000-01-01

Most of the few-group three-dimensional nodal diffusion codes used for neutronics calculations of the WWER reactors use albedo type boundary conditions on the core-reflector boundary. The conventional albedo are group-to-group reflection probabilities, defined on each outer node face. The method of characteristics is used to calculate accurate albedo by the following procedure. A many-group two-dimensional heterogeneous core-reflector problem, including a sufficient part of the core and detailed description of the adjacent reflector, is solved first. From this solution the angular flux on the core-reflector boundary is calculated in all groups for all traced neutron directions. Accurate boundary conditions can be calculated for the radial, top and bottom reflectors as well as for the absorber part of the WWER-440 reactor control assemblies. The algorithm can be used to estimate also albedo, coupling outer node faces on the radial reflector in the axial direction. Numerical results for the WWER-440 reactor are presented. (Authors)

A more accurate scheme for calculating Earth's skin temperature

Science.gov (United States)

Tsuang, Ben-Jei; Tu, Chia-Ying; Tsai, Jeng-Lin; Dracup, John A.; Arpe, Klaus; Meyers, Tilden

2009-02-01

The theoretical framework of the vertical discretization of a ground column for calculating Earth’s skin temperature is presented. The suggested discretization is derived from the evenly heat-content discretization with the optimal effective thickness for layer-temperature simulation. For the same level number, the suggested discretization is more accurate in skin temperature as well as surface ground heat flux simulations than those used in some state-of-the-art models. A proposed scheme (“op(3,2,0)”) can reduce the normalized root-mean-square error (or RMSE/STD ratio) of the calculated surface ground heat flux of a cropland site significantly to 2% (or 0.9 W m-2), from 11% (or 5 W m-2) by a 5-layer scheme used in ECMWF, from 19% (or 8 W m-2) by a 5-layer scheme used in ECHAM, and from 74% (or 32 W m-2) by a single-layer scheme used in the UCLA GCM. Better accuracy can be achieved by including more layers to the vertical discretization. Similar improvements are expected for other locations with different land types since the numerical error is inherited into the models for all the land types. The proposed scheme can be easily implemented into state-of-the-art climate models for the temperature simulation of snow, ice and soil.

A PWR Thorium Pin Cell Burnup Benchmark

Energy Technology Data Exchange (ETDEWEB)

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

2000-05-01

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

Benchmarking singlet and triplet excitation energies of molecular semiconductors for singlet fission: Tuning the amount of HF exchange and adjusting local correlation to obtain accurate functionals for singlet-triplet gaps

Science.gov (United States)

Brückner, Charlotte; Engels, Bernd

2017-01-01

Vertical and adiabatic singlet and triplet excitation energies of molecular p-type semiconductors calculated with various DFT functionals and wave-function based approaches are benchmarked against MS-CASPT2/cc-pVTZ reference values. A special focus lies on the singlet-triplet gaps that are very important in the process of singlet fission. Singlet fission has the potential to boost device efficiencies of organic solar cells, but the scope of existing singlet-fission compounds is still limited. A computational prescreening of candidate molecules could enlarge it; yet it requires efficient methods accurately predicting singlet and triplet excitation energies. Different DFT formulations (Tamm-Dancoff approximation, linear response time-dependent DFT, Δ-SCF) and spin scaling schemes along with several ab initio methods (CC2, ADC(2)/MP2, CIS(D), CIS) are evaluated. While wave-function based methods yield rather reliable singlet-triplet gaps, many DFT functionals are shown to systematically underestimate triplet excitation energies. To gain insight, the impact of exact exchange and correlation is in detail addressed.

Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry

International Nuclear Information System (INIS)

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

1993-09-01

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

BN-600 Phase III benchmark calculations

International Nuclear Information System (INIS)

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

2002-01-01

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

Benchmark of WIMS-IST against MCNP for CANDU pressure tube fast fluxes

International Nuclear Information System (INIS)

Donders, R.E.; Douglas, S.R.

2002-01-01

Pressure tube fast-flux data in CANDU are currently calculated using the multi-group neutron transport code WIMS-IST. In this study, the WIMS-IST fast flux calculations are benchmarked against MCNP calculations (a Monte Carlo particle transport code), over the range of fuel burnup and coolant density in CANDU. The comparison shows good agreement between WIMS and MCNP, with WIMS fast fluxes being 1.5% to 4% lower than the MCNP values. The difference is smallest for fresh fuel, and increases with burnup. The fast flux gradient across the pressure tube (factor of 1.23 from inner edge to outer edge) is accurately calculated by WIMS. When reporting fast fluxes in pressure tubes, these are generally given as >1.000 MeV fluxes. For WIMS, this requires an extra conversion step, since the WIMS ENDF/B libraries do not have a group boundary at 1 MeV. The conversion step is based on a fictitious isotope ONEMEV in the WIMS nuclear data library. The conversion factor in WIMS was found to be about one percent too high. When providing >1 MeV fluxes from WIMS, this partially compensates for the slight under prediction of the fast flux. Pressure tube >1 MeV fluxes from WIMS are therefore 0.5% to 3% lower than MCNP values. To obtain accurate fast flux data, neutron transport calculations must be performed on a critical cell. For this study, all calculations were performed with radial albedo boundary conditions giving a critical cell. This required the use of an albedo version of MCNP, developed at AECL. (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

Numerical benchmarking of SPEEDUP trademark against point kinetics solutions

International Nuclear Information System (INIS)

Gregory, M.V.

1993-02-01

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

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

International Nuclear Information System (INIS)

Key, S.W.

1985-01-01

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

INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

Energy Technology Data Exchange (ETDEWEB)

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

2008-09-01

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

International Nuclear Information System (INIS)

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

2008-01-01

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR-06 are highlighted, and the future of the two projects is discussed

Implementation of refined core thermal-hydraulic calculation feature in the MARS/MASTER code

International Nuclear Information System (INIS)

Joo, H. K.; Jung, J. J.; Cho, B. O.; Ji, S. K.; Lee, W. J.; Jang, M. H.

2000-01-01

As an effort to enhance the fidelity of the core thermal/hydraulic calculation in the MARS/MASTER code, a best-estimate system/core coupled code, the COBRA-III module of MASTER is activated that enables refined core T/H calculations. Since the COBRA-III module is capable of using fuel-assembly sized nodes, the resolution of the T/H solution is high so that accurate incorporation of local T/H feedback effects becomes possible. The COBRA-III module is utilized such that the refined core T/H calculation is performed using the coarse-mesh flow boundary conditions specified by MARS at both ends of the core. The results of application to the OECD MSLB benchmark analysis indicate that the local peaking factor can be reduced by upto 15% with the refined calculation through the accurate representation of the local Doppler effect evaluation, although the prediction of the global transient behaviors such as the total core power change remain essentially unaffected

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-15

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

Resolution for the Loviisa benchmark problem

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

Haemaelaeinen, A.; Kyrki-Rajamaeki, R.

2003-01-01

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

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

Discussion of OECD LWR Uncertainty Analysis in Modelling Benchmark

International Nuclear Information System (INIS)

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

2013-01-01

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

Second benchmark problem for WIPP structural computations

International Nuclear Information System (INIS)

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

1980-12-01

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

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

Science.gov (United States)

Bartlett, Philip L.; Stelbovics, Andris T.

2010-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-09-01

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

Shielding benchmark test

International Nuclear Information System (INIS)

Kawai, Masayoshi

1984-01-01

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

Benchmark test of JEF-1 evaluation by calculating fast criticalities

International Nuclear Information System (INIS)

Pelloni, S.

1986-06-01

JEF-1 basic evaluation was tested by calculating fast critical experiments using the cross section discrete-ordinates transport code ONEDANT with P/sub 3/S/sub 16/ approximation. In each computation a spherical one dimensional model was used, together with a 174 neutron group VITAMIN-E structured JEF-1 based nuclear data library, generated at EIR with NJOY and TRANSX-CTR. It is found that the JEF-1 evaluation gives accurate results comparable with ENDF/B-V and that eigenvalues agree well within 10 mk whereas reaction rates deviate by up to 10% from the experiment. U-233 total and fission cross sections seem to be underestimated in the JEF-1 evaluation in the fast energy range between 0.1 and 1 MeV. This confirms previous analysis based on diffusion theory with 71 neutron groups, performed by H. Takano and E. Sartori at NEA Data Bank. (author)

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

International Nuclear Information System (INIS)

Van Der Marck, S. C.

2012-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Benchmark test of MORSE-DD code using double-differential form cross sections

International Nuclear Information System (INIS)

Nakagawa, Masayuki; Mori, Takamasa; Ishiguro, Yukio

1985-02-01

The multi-group double-differential form cross sections (DDX) and the three dimensional Monte Carlo code MORSE-DD devised to utilize the DDX, which were developed for the fusion neutronics analysis, have been validated through many benchmark tests. All the problems tested have a 14 MeV neutron source. To compare the calculated results with the measured values, the following experiments were adopted as the benchmark problems; leakage neutron spectra from spheres composed of nine kinds of materials measured at LLNL, neutron angular spectra from the Li 2 O slab measured at FNS in JAERI, tritium production rate (TPR) in the graphite-reflected Li 2 O sphere measured at FNS and the TPR in the metallic Li sphere measured at KfK. In addition in order to test an accuracy of the calculation method in detail, spectra of neutrons scattered from a small sample and various reaction rates in a Li 2 O cylinder were compared between the present method and the continuous energy Monte Carlo method. The nuclear data files used are mainly ENDF/B4 and partly JENDL-3PR1. The tests were carried out through a comparison with the measured values and also with the results obtained from the conventional Legendre expansion method and the continuous energy Monte Carlo method. It is found that the results by the present method are more accurate than those by the conventional one and agree well with those by the continuous energy Monte Carlo calculations. Discrepancies due to the nuclear data are also discussed. (author)

Benchmark referencing of neutron dosimetry measurements

International Nuclear Information System (INIS)

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

1980-01-01

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

Optimization of extracranial stereotactic radiation therapy of small lung lesions using accurate dose calculation algorithms

International Nuclear Information System (INIS)

Dobler, Barbara; Walter, Cornelia; Knopf, Antje; Fabri, Daniella; Loeschel, Rainer; Polednik, Martin; Schneider, Frank; Wenz, Frederik; Lohr, Frank

2006-01-01

The aim of this study was to compare and to validate different dose calculation algorithms for the use in radiation therapy of small lung lesions and to optimize the treatment planning using accurate dose calculation algorithms. A 9-field conformal treatment plan was generated on an inhomogeneous phantom with lung mimics and a soft tissue equivalent insert, mimicking a lung tumor. The dose distribution was calculated with the Pencil Beam and Collapsed Cone algorithms implemented in Masterplan (Nucletron) and the Monte Carlo system XVMC and validated using Gafchromic EBT films. Differences in dose distribution were evaluated. The plans were then optimized by adding segments to the outer shell of the target in order to increase the dose near the interface to the lung. The Pencil Beam algorithm overestimated the dose by up to 15% compared to the measurements. Collapsed Cone and Monte Carlo predicted the dose more accurately with a maximum difference of -8% and -3% respectively compared to the film. Plan optimization by adding small segments to the peripheral parts of the target, creating a 2-step fluence modulation, allowed to increase target coverage and homogeneity as compared to the uncorrected 9 field plan. The use of forward 2-step fluence modulation in radiotherapy of small lung lesions allows the improvement of tumor coverage and dose homogeneity as compared to non-modulated treatment plans and may thus help to increase the local tumor control probability. While the Collapsed Cone algorithm is closer to measurements than the Pencil Beam algorithm, both algorithms are limited at tissue/lung interfaces, leaving Monte-Carlo the most accurate algorithm for dose prediction

Nonparametric estimation of benchmark doses in environmental risk assessment

Science.gov (United States)

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

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-12-15

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

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.

International benchmark on the natural convection test in Phenix reactor

International Nuclear Information System (INIS)

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

2013-01-01

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

Accurate Bit Error Rate Calculation for Asynchronous Chaos-Based DS-CDMA over Multipath Channel

Science.gov (United States)

Kaddoum, Georges; Roviras, Daniel; Chargé, Pascal; Fournier-Prunaret, Daniele

2009-12-01

An accurate approach to compute the bit error rate expression for multiuser chaosbased DS-CDMA system is presented in this paper. For more realistic communication system a slow fading multipath channel is considered. A simple RAKE receiver structure is considered. Based on the bit energy distribution, this approach compared to others computation methods existing in literature gives accurate results with low computation charge. Perfect estimation of the channel coefficients with the associated delays and chaos synchronization is assumed. The bit error rate is derived in terms of the bit energy distribution, the number of paths, the noise variance, and the number of users. Results are illustrated by theoretical calculations and numerical simulations which point out the accuracy of our approach.

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

Reactor physics calculations on HTR type configurations

Energy Technology Data Exchange (ETDEWEB)

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

1995-04-01

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

Reactor physics calculations on HTR type configurations

International Nuclear Information System (INIS)

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

1995-04-01

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

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

Thermal reactor benchmark tests on JENDL-2

International Nuclear Information System (INIS)

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

1983-11-01

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

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

Science.gov (United States)

Neil, Amanda; Pfeffer, Sally; Burnett, Leslie

2013-01-01

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

A highly accurate algorithm for the solution of the point kinetics equations

International Nuclear Information System (INIS)

Ganapol, B.D.

2013-01-01

Highlights: • Point kinetics equations for nuclear reactor transient analysis are numerically solved to extreme accuracy. • Results for classic benchmarks found in the literature are given to 9-digit accuracy. • Recent results of claimed accuracy are shown to be less accurate than claimed. • Arguably brings a chapter of numerical evaluation of the PKEs to a close. - Abstract: Attempts to resolve the point kinetics equations (PKEs) describing nuclear reactor transients have been the subject of numerous articles and texts over the past 50 years. Some very innovative methods, such as the RTS (Reactor Transient Simulation) and CAC (Continuous Analytical Continuation) methods of G.R. Keepin and J. Vigil respectively, have been shown to be exceptionally useful. Recently however, several authors have developed methods they consider accurate without a clear basis for their assertion. In response, this presentation will establish a definitive set of benchmarks to enable those developing PKE methods to truthfully assess the degree of accuracy of their methods. Then, with these benchmarks, two recently published methods, found in this journal will be shown to be less accurate than claimed and a legacy method from 1984 will be confirmed

The benchmark testing of 9Be of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2002-01-01

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

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

TRX and UO2 criticality benchmarks with SAM-CE

International Nuclear Information System (INIS)

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

1980-01-01

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

Benchmarking of FA2D/PARCS Code Package

International Nuclear Information System (INIS)

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

2006-01-01

FA2D/PARCS code package is used at Faculty of Electrical Engineering and Computing (FER), University of Zagreb, for static and dynamic reactor core analyses. It consists of two codes: FA2D and PARCS. FA2D is a multigroup two dimensional transport theory code for burn-up calculations based on collision probability method, developed at FER. It generates homogenised cross sections both of single pins and entire fuel assemblies. PARCS is an advanced nodal code developed at Purdue University for US NRC and it is based on neutron diffusion theory for three dimensional whole core static and dynamic calculations. It is modified at FER to enable internal 3D depletion calculation and usage of neutron cross section data in a format produced by FA2D and interface codes. The FA2D/PARCS code system has been validated on NPP Krsko operational data (Cycles 1 and 21). As we intend to use this code package for development of IRIS reactor loading patterns the first logical step was to validate the FA2D/PARCS code package on a set of IRIS benchmarks, starting from simple unit fuel cell, via fuel assembly, to full core benchmark. The IRIS 17x17 fuel with erbium burnable absorber was used in last full core benchmark. The results of modelling the IRIS full core benchmark using FA2D/PARCS code package have been compared with reference data showing the adequacy of FA2D/PARCS code package model for IRIS reactor core design.(author)

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

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

Science.gov (United States)

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

2017-10-07

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

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

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2014-01-01

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

Scaled MP3 Non-Covalent Interaction Energies Agree Closely with Accurate CCSD(T) Benchmark Data

Czech Academy of Sciences Publication Activity Database

Pitoňák, Michal; Neogrady, P.; Černý, Jiří; Grimme, S.; Hobza, Pavel

2009-01-01

Roč. 10, č. 1 (2009), s. 282-289 ISSN 1439-4235 R&D Projects: GA MŠk LC512 Institutional research plan: CEZ:AV0Z40550506 Keywords : Scaled MP3 * CCSD(T) Benchmark Data * Extended Data Set Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 3.453, year: 2009

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

A new model for the accurate calculation of natural gas viscosity

Directory of Open Access Journals (Sweden)

Xiaohong Yang

2017-03-01

Full Text Available Viscosity of natural gas is a basic and important parameter, of theoretical and practical significance in the domain of natural gas recovery, transmission and processing. In order to obtain the accurate viscosity data efficiently at a low cost, a new model and its corresponding functional relation are derived on the basis of the relationship among viscosity, temperature and density derived from the kinetic theory of gases. After the model parameters were optimized using a lot of experimental data, the diagram showing the variation of viscosity along with temperature and density is prepared, showing that: ① the gas viscosity increases with the increase of density as well as the increase of temperature in the low density region; ② the gas viscosity increases with the decrease of temperature in high density region. With this new model, the viscosity of 9 natural gas samples was calculated precisely. The average relative deviation between these calculated values and 1539 experimental data measured at 250–450 K and 0.10–140.0 MPa is less than 1.9%. Compared with the 793 experimental data with a measurement error less than 0.5%, the maximum relative deviation is less than 0.98%. It is concluded that this new model is more advantageous than the previous 8 models in terms of simplicity, accuracy, fast calculation, and direct applicability to the CO2 bearing gas samples.

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)

Solution of the new Dukovany benchmark using the new version of the KARATE-440 code

International Nuclear Information System (INIS)

Hegyi, G.; Kereszturi, A.; Maraczy, C.

2008-01-01

A new -so called second generation - fuel type developed by Russian vendor (TVEL) for WWER-440 has been introduced. Even the outer parameters of the assembly remain unchanged, the length of the pins was increased and 6 of the 126 pins have been doped by gadolinium. The above mentioned modifications requiring more accurate calculations have necessitated the further development and validation of our KARATE code system. The paper summarizes the capabilities of the new version of the KARATE program concerning the properties of currently used and newly developed fuel types. On the basis of an international benchmark published from Dukovany NPP validation procedure has been started. (Authors)

Introduction to 'International Handbook of Criticality Safety Benchmark Experiments'

International Nuclear Information System (INIS)

Komuro, Yuichi

1998-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in 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) is now an official activity of the Organization for Economic Cooperation and Development-Nuclear Energy Agency (OECD-NEA). 'International Handbook of Criticality Safety Benchmark Experiments' was prepared and is updated year by year by the working group of the project. 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 calculation techniques used. The author briefly introduces the informative handbook and would like to encourage Japanese engineers who are in charge of nuclear criticality safety to use the handbook. (author)

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.

MoleculeNet: a benchmark for molecular machine learning.

Science.gov (United States)

Wu, Zhenqin; Ramsundar, Bharath; Feinberg, Evan N; Gomes, Joseph; Geniesse, Caleb; Pappu, Aneesh S; Leswing, Karl; Pande, Vijay

2018-01-14

Molecular machine learning has been maturing rapidly over the last few years. Improved methods and the presence of larger datasets have enabled machine learning algorithms to make increasingly accurate predictions about molecular properties. However, algorithmic progress has been limited due to the lack of a standard benchmark to compare the efficacy of proposed methods; most new algorithms are benchmarked on different datasets making it challenging to gauge the quality of proposed methods. This work introduces MoleculeNet, a large scale benchmark for molecular machine learning. MoleculeNet curates multiple public datasets, establishes metrics for evaluation, and offers high quality open-source implementations of multiple previously proposed molecular featurization and learning algorithms (released as part of the DeepChem open source library). MoleculeNet benchmarks demonstrate that learnable representations are powerful tools for molecular machine learning and broadly offer the best performance. However, this result comes with caveats. Learnable representations still struggle to deal with complex tasks under data scarcity and highly imbalanced classification. For quantum mechanical and biophysical datasets, the use of physics-aware featurizations can be more important than choice of particular learning algorithm.

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

Energy Technology Data Exchange (ETDEWEB)

Hannstein, Volker; Sommer, Fabian

2017-05-15

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

Monte Carlo burnup simulation of the TAKAHAMA-3 benchmark experiment

International Nuclear Information System (INIS)

Dalle, Hugo M.

2009-01-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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.

Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark

International Nuclear Information System (INIS)

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

2008-01-01

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

The International Criticality Safety Benchmark Evaluation Project (ICSBEP)

International Nuclear Information System (INIS)

Briggs, J.B.

2003-01-01

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 by the United States Department of Energy. The ICSBEP became an official activity of the Organisation for Economic Cooperation and Development (OECD) - Nuclear Energy Agency (NEA) in 1995. Representatives from the United States, United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Yugoslavia, Kazakhstan, Israel, Spain, and Brazil are now participating. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an OECD handbook entitled 'International Handbook of Evaluated Criticality Safety Benchmark Experiments.' The 2003 Edition of the Handbook contains benchmark model specifications for 3070 critical or subcritical configurations that are intended for validating computer codes that calculate effective neutron multiplication and for testing basic nuclear data. (author)

Accurate protein structure modeling using sparse NMR data and homologous structure information.

Science.gov (United States)

Thompson, James M; Sgourakis, Nikolaos G; Liu, Gaohua; Rossi, Paolo; Tang, Yuefeng; Mills, Jeffrey L; Szyperski, Thomas; Montelione, Gaetano T; Baker, David

2012-06-19

While information from homologous structures plays a central role in X-ray structure determination by molecular replacement, such information is rarely used in NMR structure determination because it can be incorrect, both locally and globally, when evolutionary relationships are inferred incorrectly or there has been considerable evolutionary structural divergence. Here we describe a method that allows robust modeling of protein structures of up to 225 residues by combining (1)H(N), (13)C, and (15)N backbone and (13)Cβ chemical shift data, distance restraints derived from homologous structures, and a physically realistic all-atom energy function. Accurate models are distinguished from inaccurate models generated using incorrect sequence alignments by requiring that (i) the all-atom energies of models generated using the restraints are lower than models generated in unrestrained calculations and (ii) the low-energy structures converge to within 2.0 Å backbone rmsd over 75% of the protein. Benchmark calculations on known structures and blind targets show that the method can accurately model protein structures, even with very remote homology information, to a backbone rmsd of 1.2-1.9 Å relative to the conventional determined NMR ensembles and of 0.9-1.6 Å relative to X-ray structures for well-defined regions of the protein structures. This approach facilitates the accurate modeling of protein structures using backbone chemical shift data without need for side-chain resonance assignments and extensive analysis of NOESY cross-peak assignments.

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-15

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

An accurate algorithm to calculate the Hurst exponent of self-similar processes

International Nuclear Information System (INIS)

Fernández-Martínez, M.; Sánchez-Granero, M.A.; Trinidad Segovia, J.E.; Román-Sánchez, I.M.

2014-01-01

In this paper, we introduce a new approach which generalizes the GM2 algorithm (introduced in Sánchez-Granero et al. (2008) [52]) as well as fractal dimension algorithms (FD1, FD2 and FD3) (first appeared in Sánchez-Granero et al. (2012) [51]), providing an accurate algorithm to calculate the Hurst exponent of self-similar processes. We prove that this algorithm performs properly in the case of short time series when fractional Brownian motions and Lévy stable motions are considered. We conclude the paper with a dynamic study of the Hurst exponent evolution in the S and P500 index stocks. - Highlights: • We provide a new approach to properly calculate the Hurst exponent. • This generalizes FD algorithms and GM2, introduced previously by the authors. • This method (FD4) results especially appropriate for short time series. • FD4 may be used in both unifractal and multifractal contexts. • As an empirical application, we show that S and P500 stocks improved their efficiency

An accurate algorithm to calculate the Hurst exponent of self-similar processes

Energy Technology Data Exchange (ETDEWEB)

Fernández-Martínez, M., E-mail: fmm124@ual.es [Department of Mathematics, Faculty of Science, Universidad de Almería, 04120 Almería (Spain); Sánchez-Granero, M.A., E-mail: misanche@ual.es [Department of Mathematics, Faculty of Science, Universidad de Almería, 04120 Almería (Spain); Trinidad Segovia, J.E., E-mail: jetrini@ual.es [Department of Accounting and Finance, Faculty of Economics and Business, Universidad de Almería, 04120 Almería (Spain); Román-Sánchez, I.M., E-mail: iroman@ual.es [Department of Accounting and Finance, Faculty of Economics and Business, Universidad de Almería, 04120 Almería (Spain)

2014-06-27

In this paper, we introduce a new approach which generalizes the GM2 algorithm (introduced in Sánchez-Granero et al. (2008) [52]) as well as fractal dimension algorithms (FD1, FD2 and FD3) (first appeared in Sánchez-Granero et al. (2012) [51]), providing an accurate algorithm to calculate the Hurst exponent of self-similar processes. We prove that this algorithm performs properly in the case of short time series when fractional Brownian motions and Lévy stable motions are considered. We conclude the paper with a dynamic study of the Hurst exponent evolution in the S and P500 index stocks. - Highlights: • We provide a new approach to properly calculate the Hurst exponent. • This generalizes FD algorithms and GM2, introduced previously by the authors. • This method (FD4) results especially appropriate for short time series. • FD4 may be used in both unifractal and multifractal contexts. • As an empirical application, we show that S and P500 stocks improved their efficiency.

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.

CompaRNA: a server for continuous benchmarking of automated methods for RNA secondary structure prediction

Science.gov (United States)

Puton, Tomasz; Kozlowski, Lukasz P.; Rother, Kristian M.; Bujnicki, Janusz M.

2013-01-01

We present a continuous benchmarking approach for the assessment of RNA secondary structure prediction methods implemented in the CompaRNA web server. As of 3 October 2012, the performance of 28 single-sequence and 13 comparative methods has been evaluated on RNA sequences/structures released weekly by the Protein Data Bank. We also provide a static benchmark generated on RNA 2D structures derived from the RNAstrand database. Benchmarks on both data sets offer insight into the relative performance of RNA secondary structure prediction methods on RNAs of different size and with respect to different types of structure. According to our tests, on the average, the most accurate predictions obtained by a comparative approach are generated by CentroidAlifold, MXScarna, RNAalifold and TurboFold. On the average, the most accurate predictions obtained by single-sequence analyses are generated by CentroidFold, ContextFold and IPknot. The best comparative methods typically outperform the best single-sequence methods if an alignment of homologous RNA sequences is available. This article presents the results of our benchmarks as of 3 October 2012, whereas the rankings presented online are continuously updated. We will gladly include new prediction methods and new measures of accuracy in the new editions of CompaRNA benchmarks. PMID:23435231

CompaRNA: a server for continuous benchmarking of automated methods for RNA secondary structure prediction.

Science.gov (United States)

Puton, Tomasz; Kozlowski, Lukasz P; Rother, Kristian M; Bujnicki, Janusz M

2013-04-01

We present a continuous benchmarking approach for the assessment of RNA secondary structure prediction methods implemented in the CompaRNA web server. As of 3 October 2012, the performance of 28 single-sequence and 13 comparative methods has been evaluated on RNA sequences/structures released weekly by the Protein Data Bank. We also provide a static benchmark generated on RNA 2D structures derived from the RNAstrand database. Benchmarks on both data sets offer insight into the relative performance of RNA secondary structure prediction methods on RNAs of different size and with respect to different types of structure. According to our tests, on the average, the most accurate predictions obtained by a comparative approach are generated by CentroidAlifold, MXScarna, RNAalifold and TurboFold. On the average, the most accurate predictions obtained by single-sequence analyses are generated by CentroidFold, ContextFold and IPknot. The best comparative methods typically outperform the best single-sequence methods if an alignment of homologous RNA sequences is available. This article presents the results of our benchmarks as of 3 October 2012, whereas the rankings presented online are continuously updated. We will gladly include new prediction methods and new measures of accuracy in the new editions of CompaRNA benchmarks.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Calculation of accurate small angle X-ray scattering curves from coarse-grained protein models

DEFF Research Database (Denmark)

Stovgaard, Kasper; Andreetta, Christian; Ferkinghoff-Borg, Jesper

2010-01-01

, which is paramount for structure determination based on statistical inference. Results: We present a method for the efficient calculation of accurate SAXS curves based on the Debye formula and a set of scattering form factors for dummy atom representations of amino acids. Such a method avoids......DBN. This resulted in a significant improvement in the decoy recognition performance. In conclusion, the presented method shows great promise for use in statistical inference of protein structures from SAXS data....

Coulomb-Sturmian separable expansion approach: Three-body Faddeev calculations for Coulomb-like interactions

International Nuclear Information System (INIS)

Papp, Z.; Plessas, W.

1996-01-01

We demonstrate the feasibility and efficiency of the Coulomb-Sturmian separable expansion method for generating accurate solutions of the Faddeev equations. Results obtained with this method are reported for several benchmark cases of bosonic and fermionic three-body systems. Correct bound-state results in agreement with the ones established in the literature are achieved for short-range interactions. We outline the formalism for the treatment of three-body Coulomb systems and present a bound-state calculation for a three-boson system interacting via Coulomb plus short-range forces. The corresponding result is in good agreement with the answer from a recent stochastic-variational-method calculation. copyright 1996 The American Physical Society

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

Science.gov (United States)

Heuel-Fabianek, Burkhard; Hille, Ralf

2005-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

MAAP4 hot leg and lower head failure benchmarking

International Nuclear Information System (INIS)

Lee, S.J.; Henry, R.E.; Paik, C.Y.; Conzen, J.; Luangdilok, W.

2009-01-01

The MAAP4 material creep calculation was compared with the experiments reported by Maile, et al., for a 0.7 m diameter hot leg, with a thickness of 47 mm, which is pressurized to 16.3 MPa and heated to temperatures in excess of 700degC. These experiments showed that the carbon steel hot leg would undergo material creep to a failure state in approximately 1,100 seconds. In addition, the MAAP4 creep calculation was compared with the lower head failure tests performed at the Sandia National Laboratories (SNL). These experiments were performed using scaled models of a typical Reactor Pressure Vessel lower head. The test vessel was fabricated from SA533B1 steel with an inner diameter of 0.91 m and a nominal thickness of 30 mm. The experiments were performed at around 10 MPa internal pressure with various imposed heat flux distributions. The onset of creep was observed to occur between 660degC and 705degC. The MAAP4 model provides a good characterization of the material creep behavior. For the hot leg test benchmark, the key is determining the correct equivalent stress when the stress is multi-axial. A good agreement was obtained when a multiplier of 1.09 to the hoop stress was used. For the lower head failure benchmark, using correct creep properties is important. The SNL test vessel material was fabricated as SA533B1 steel. However, when the experimental vessel material was tested for creep properties it turned out to be significantly weaker than the reactor vessel steel which has the same identification. Also, the material undergoing phase transition and becoming stronger at high temperatures has to be considered for accurate prediction of the failure time. A good agreement was obtained when the creep data of Jeong, et al., was used. (author)

Accurate Holdup Calculations with Predictive Modeling & Data Integration

Energy Technology Data Exchange (ETDEWEB)

Azmy, Yousry [North Carolina State Univ., Raleigh, NC (United States). Dept. of Nuclear Engineering; Cacuci, Dan [Univ. of South Carolina, Columbia, SC (United States). Dept. of Mechanical Engineering

2017-04-03

In facilities that process special nuclear material (SNM) it is important to account accurately for the fissile material that enters and leaves the plant. Although there are many stages and processes through which materials must be traced and measured, the focus of this project is material that is “held-up” in equipment, pipes, and ducts during normal operation and that can accumulate over time into significant quantities. Accurately estimating the holdup is essential for proper SNM accounting (vis-à-vis nuclear non-proliferation), criticality and radiation safety, waste management, and efficient plant operation. Usually it is not possible to directly measure the holdup quantity and location, so these must be inferred from measured radiation fields, primarily gamma and less frequently neutrons. Current methods to quantify holdup, i.e. Generalized Geometry Holdup (GGH), primarily rely on simple source configurations and crude radiation transport models aided by ad hoc correction factors. This project seeks an alternate method of performing measurement-based holdup calculations using a predictive model that employs state-of-the-art radiation transport codes capable of accurately simulating such situations. Inverse and data assimilation methods use the forward transport model to search for a source configuration that best matches the measured data and simultaneously provide an estimate of the level of confidence in the correctness of such configuration. In this work the holdup problem is re-interpreted as an inverse problem that is under-determined, hence may permit multiple solutions. A probabilistic approach is applied to solving the resulting inverse problem. This approach rates possible solutions according to their plausibility given the measurements and initial information. This is accomplished through the use of Bayes’ Theorem that resolves the issue of multiple solutions by giving an estimate of the probability of observing each possible solution. To use

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

Science.gov (United States)

Han, Jeong-Hwan; Oda, Takuji

2018-04-01

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

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

International Nuclear Information System (INIS)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

2002-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-02-01

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

Accurate convolution/superposition for multi-resolution dose calculation using cumulative tabulated kernels

International Nuclear Information System (INIS)

Lu Weiguo; Olivera, Gustavo H; Chen Mingli; Reckwerdt, Paul J; Mackie, Thomas R

2005-01-01

is demonstrated to be the most accurate one for multi-resolution dose calculations

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

International Nuclear Information System (INIS)

Grundmann, U.; Rohde, U.

1997-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

Use of Monte Carlo computation in benchmarking radiotherapy treatment planning system algorithms

International Nuclear Information System (INIS)

Lewis, R.D.; Ryde, S.J.S.; Seaby, A.W.; Hancock, D.A.; Evans, C.J.

2000-01-01

Radiotherapy treatments are becoming more complex, often requiring the dose to be calculated in three dimensions and sometimes involving the application of non-coplanar beams. The ability of treatment planning systems to accurately calculate dose under a range of these and other irradiation conditions requires evaluation. Practical assessment of such arrangements can be problematical, especially when a heterogeneous medium is used. This work describes the use of Monte Carlo computation as a benchmarking tool to assess the dose distribution of external photon beam plans obtained in a simple heterogeneous phantom by several commercially available 3D and 2D treatment planning system algorithms. For comparison, practical measurements were undertaken using film dosimetry. The dose distributions were calculated for a variety of irradiation conditions designed to show the effects of surface obliquity, inhomogeneities and missing tissue above tangential beams. The results show maximum dose differences of 47% between some planning algorithms and film at a point 1 mm below a tangentially irradiated surface. Overall, the dose distribution obtained from film was most faithfully reproduced by the Monte Carlo N-Particle results illustrating the potential of Monte Carlo computation in evaluating treatment planning system algorithms. (author)

Benchmarking of MCAM 4.0 with the ITER 3D Model

International Nuclear Information System (INIS)

Ying Li; Lei Lu; Aiping Ding; Haimin Hu; Qin Zeng; Shanliang Zheng; Yican Wu

2006-01-01

Monte Carlo particle transport simulations are widely employed in fields such as nuclear engineering, radio-therapy and space science. Describing and verifying the 3D geometry of fusion devices, however, are among the most complex tasks of MCNP calculation problems in nuclear analysis. The manual modeling of a complex geometry for MCNP code, though a common practice, is an extensive, time-consuming, and error prone task. An efficient solution is to shift the geometric modeling into Computer Aided Design(CAD) systems and to use an interface for MCNP to convert the CAD model to MCNP file. The advantage of this approach lies in the fact that it allows access to full features of modern CAD systems facilitating the geometric modeling and utilizing the existing CAD models. MCAM(MCNP Automatic Modeling System) is an integrated tool for CAD model preprocessing, accurate bi-directional conversion between CAD/MCNP models, neutronics property processing and geometric modeling developed by FDS team in ASIPP and Hefei University of Technology. MCAM4.0 has been extended and enhanced to support various CAD file formats and the preprocessing of CAD model, such as healing, automatic model reconstruction, overlap detection and correction, automatic void modeling. The ITER international benchmark model is provided by ITER international team to compare the CAD/MCNP programs being developed in the ITER participant teams. It is created in CATIA/V5, which has been chosen as the CAD system for ITER design, including all the important parts and components of the ITER device. The benchmark model contains vast curve surfaces, which can fully test the ability of MCNP/CAD codes. The whole processing procedure of this model will be presented in this paper, which includes the geometric model processing, neutroics property processing, converting to MCNP input file, calculating with MCNP and analysis. The nuclear analysis results of the model will be given in the end. Although these preliminary

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.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Pericles and Attila results for the C5G7 MOX benchmark problems

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Accurate calculation of field and carrier distributions in doped semiconductors

Directory of Open Access Journals (Sweden)

Wenji Yang

2012-06-01

Full Text Available We use the numerical squeezing algorithm(NSA combined with the shooting method to accurately calculate the built-in fields and carrier distributions in doped silicon films (SFs in the micron and sub-micron thickness range and results are presented in graphical form for variety of doping profiles under different boundary conditions. As a complementary approach, we also present the methods and the results of the inverse problem (IVP - finding out the doping profile in the SFs for given field distribution. The solution of the IVP provides us the approach to arbitrarily design field distribution in SFs - which is very important for low dimensional (LD systems and device designing. Further more, the solution of the IVP is both direct and much easy for all the one-, two-, and three-dimensional semiconductor systems. With current efforts focused on the LD physics, knowing of the field and carrier distribution details in the LD systems will facilitate further researches on other aspects and hence the current work provides a platform for those researches.

Reconstruction of pin burnup characteristics from nodal calculations in hexagonal geometry

International Nuclear Information System (INIS)

Yang, W.S.; Finck, P.J.; Khalil, H.S.

1990-01-01

A reconstruction method has been developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intra-modal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method has been tested by performing several numerical benchmark calculations and by comparing predicted local burnups to values measured for experimental assemblies in EBR-11. The results indicate that the reconstruction methods are quite accurate, yielding maximum errors in power and nuclide densities that are less than 2% for driver assemblies and typically less than 5% for blanket assemblies. 14 refs., 2 figs., 5 tabs

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

International Nuclear Information System (INIS)

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

2000-01-01

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

An automated protocol for performance benchmarking a widefield fluorescence microscope.

Science.gov (United States)

Halter, Michael; Bier, Elianna; DeRose, Paul C; Cooksey, Gregory A; Choquette, Steven J; Plant, Anne L; Elliott, John T

2014-11-01

Widefield fluorescence microscopy is a highly used tool for visually assessing biological samples and for quantifying cell responses. Despite its widespread use in high content analysis and other imaging applications, few published methods exist for evaluating and benchmarking the analytical performance of a microscope. Easy-to-use benchmarking methods would facilitate the use of fluorescence imaging as a quantitative analytical tool in research applications, and would aid the determination of instrumental method validation for commercial product development applications. We describe and evaluate an automated method to characterize a fluorescence imaging system's performance by benchmarking the detection threshold, saturation, and linear dynamic range to a reference material. The benchmarking procedure is demonstrated using two different materials as the reference material, uranyl-ion-doped glass and Schott 475 GG filter glass. Both are suitable candidate reference materials that are homogeneously fluorescent and highly photostable, and the Schott 475 GG filter glass is currently commercially available. In addition to benchmarking the analytical performance, we also demonstrate that the reference materials provide for accurate day to day intensity calibration. Published 2014 Wiley Periodicals Inc. Published 2014 Wiley Periodicals Inc. This article is a US government work and, as such, is in the public domain in the United States of America.

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

Audit materiality and risk : benchmarks and the impact on the audit process / J.J. Swart

OpenAIRE

Swart, Jacobus Johannes

2013-01-01

The objective of this study is to address the gap that exists in the literature regarding quantifiable guidelines, benchmarks and consistency of applications. During the research acceptable benchmarks for the calculation or quantification of the elements linked to materiality and audit risk were found. The benchmarks are in compliance with the practices and the requirements of the ISAs and regulations. Models and benchmarks based on literature were used as a basis and modified for application...

Integral benchmark test of JENDL-4.0 for U-233 systems with ICSBEP handbook

International Nuclear Information System (INIS)

Kuwagaki, Kazuki; Nagaya, Yasunobu

2017-03-01

The integral benchmark test of JENDL-4.0 for U-233 systems using the continuous-energy Monte Carlo code MVP was conducted. The previous benchmark test was performed only for U-233 thermal solution and fast metallic systems in the ICSBEP handbook. In this study, MVP input files were prepared for uninvestigated benchmark problems in the handbook including compound thermal systems (mainly lattice systems) and integral benchmark test was performed. The prediction accuracy of JENDL-4.0 was evaluated for effective multiplication factors (k eff 's) of the U-233 systems. As a result, a trend of underestimation was observed for all the categories of U-233 systems. In the benchmark test of ENDF/B-VII.1 for U-233 systems with the ICSBEP handbook, it is reported that a decreasing trend of calculated k eff values in association with a parameter ATFF (Above-Thermal Fission Fraction) is observed. The ATFF values were also calculated in this benchmark test of JENDL-4.0 and the same trend as ENDF/B-VII.1 was observed. A CD-ROM is attached as an appendix. (J.P.N.)

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

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-01-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

SU-F-T-152: Experimental Validation and Calculation Benchmark for a Commercial Monte Carlo Pencil BeamScanning Proton Therapy Treatment Planning System in Heterogeneous Media

Energy Technology Data Exchange (ETDEWEB)

Lin, L; Huang, S; Kang, M; Ainsley, C; Simone, C; McDonough, J; Solberg, T [University of Pennsylvania, Philadelphia, PA (United States)

2016-06-15

Purpose: Eclipse AcurosPT 13.7, the first commercial Monte Carlo pencil beam scanning (PBS) proton therapy treatment planning system (TPS), was experimentally validated for an IBA dedicated PBS nozzle in the CIRS 002LFC thoracic phantom. Methods: A two-stage procedure involving the use of TOPAS 1.3 simulations was performed. First, Geant4-based TOPAS simulations in this phantom were experimentally validated for single and multi-spot profiles at several depths for 100, 115, 150, 180, 210 and 225 MeV proton beams, using the combination of a Lynx scintillation detector and a MatriXXPT ionization chamber array. Second, benchmark calculations were performed with both AcurosPT and TOPAS in a phantom identical to the CIRS 002LFC, with the exception that the CIRS bone/mediastinum/lung tissues were replaced with similar tissues that are predefined in AcurosPT (a limitation of this system which necessitates the two stage procedure). Results: Spot sigmas measured in tissue were in agreement within 0.2 mm of TOPAS simulation for all six energies, while AcurosPT was consistently found to have larger spot sigma (<0.7 mm) than TOPAS. Using absolute dose calibration by MatriXXPT, the agreements between profiles measurements and TOPAS simulation, and calculation benchmarks are over 97% except near the end of range using 2 mm/2% gamma criteria. Overdosing and underdosing were observed at the low and high density side of tissue interfaces, respectively, and these increased with increasing depth and decreasing energy. Near the mediastinum/lung interface, the magnitude can exceed 5 mm/10%. Furthermore, we observed >5% quenching effect in the conversion of Lynx measurements to dose. Conclusion: We recommend the use of an ionization chamber array in combination with the scintillation detector to measure absolute dose and relative PBS spot characteristics. We also recommend the use of an independent Monte Carlo calculation benchmark for the commissioning of a commercial TPS. Partially

Adjoint electron Monte Carlo calculations

International Nuclear Information System (INIS)

Jordan, T.M.

1986-01-01

Adjoint Monte Carlo is the most efficient method for accurate analysis of space systems exposed to natural and artificially enhanced electron environments. Recent adjoint calculations for isotropic electron environments include: comparative data for experimental measurements on electronics boxes; benchmark problem solutions for comparing total dose prediction methodologies; preliminary assessment of sectoring methods used during space system design; and total dose predictions on an electronics package. Adjoint Monte Carlo, forward Monte Carlo, and experiment are in excellent agreement for electron sources that simulate space environments. For electron space environments, adjoint Monte Carlo is clearly superior to forward Monte Carlo, requiring one to two orders of magnitude less computer time for relatively simple geometries. The solid-angle sectoring approximations used for routine design calculations can err by more than a factor of 2 on dose in simple shield geometries. For critical space systems exposed to severe electron environments, these potential sectoring errors demand the establishment of large design margins and/or verification of shield design by adjoint Monte Carlo/experiment

Benchmarking of SIMULATE-3 on engineering workstations

International Nuclear Information System (INIS)

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

1990-01-01

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

Benchmarking the cost efficiency of community care in Australian child and adolescent mental health services: implications for future benchmarking.

Science.gov (United States)

Furber, Gareth; Brann, Peter; Skene, Clive; Allison, Stephen

2011-06-01

The purpose of this study was to benchmark the cost efficiency of community care across six child and adolescent mental health services (CAMHS) drawn from different Australian states. Organizational, contact and outcome data from the National Mental Health Benchmarking Project (NMHBP) data-sets were used to calculate cost per "treatment hour" and cost per episode for the six participating organizations. We also explored the relationship between intake severity as measured by the Health of the Nations Outcome Scales for Children and Adolescents (HoNOSCA) and cost per episode. The average cost per treatment hour was $223, with cost differences across the six services ranging from a mean of $156 to $273 per treatment hour. The average cost per episode was $3349 (median $1577) and there were significant differences in the CAMHS organizational medians ranging from $388 to $7076 per episode. HoNOSCA scores explained at best 6% of the cost variance per episode. These large cost differences indicate that community CAMHS have the potential to make substantial gains in cost efficiency through collaborative benchmarking. Benchmarking forums need considerable financial and business expertise for detailed comparison of business models for service provision.

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-11-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

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

Benchmark calculations with correlated molecular wave functions. VI. Second row A2 and first row/second row AB diatomic molecules

International Nuclear Information System (INIS)

Woon, D.E.; Dunning, T.H. Jr.

1994-01-01

Benchmark calculations employing the correlation consistent basis sets of Dunning and co-workers are reported for the following diatomic species: Al 2 , Si 2 , P 2 , S 2 , Cl 2 , SiS, PS, PN, PO, and SO. Internally contracted multireference configuration interaction (CMRCI) calculations (correlating valence electrons only) have been performed for each species. For Cl 2 , P 2 , and PN, calculations have also been carried out using Moller--Plesset perturbation theory (MP2, MP3, MP4) and the singles and doubles coupled-cluster method with and without perturbative triples [CCSD, CCSD(T)]. Spectroscopic constants and dissociation energies are reported for the ground state of each species. In addition, the low-lying excited states of Al 2 and Si 2 have been investigated. Estimated complete basis set (CBS) limits for the dissociation energies, D e , and other spectroscopic constants are obtained from simple exponential extrapolations of the computed quantities. At the CBS limit the root-mean-square (rms) error in D e for the CMRCI calculations, the intrinsic error, on the ten species considered here is 3.9 kcal/mol; for r e the rms intrinsic error is 0.009 A, and for ω e it is 5.1 cm -1

SU-E-T-22: A Deterministic Solver of the Boltzmann-Fokker-Planck Equation for Dose Calculation

Energy Technology Data Exchange (ETDEWEB)

Hong, X; Gao, H [Shanghai Jiao Tong University, Shanghai, Shanghai (China); Paganetti, H [Massachusetts General Hospital, Boston, MA (United States)

2015-06-15

Purpose: The Boltzmann-Fokker-Planck equation (BFPE) accurately models the migration of photons/charged particles in tissues. While the Monte Carlo (MC) method is popular for solving BFPE in a statistical manner, we aim to develop a deterministic BFPE solver based on various state-of-art numerical acceleration techniques for rapid and accurate dose calculation. Methods: Our BFPE solver is based on the structured grid that is maximally parallelizable, with the discretization in energy, angle and space, and its cross section coefficients are derived or directly imported from the Geant4 database. The physical processes that are taken into account are Compton scattering, photoelectric effect, pair production for photons, and elastic scattering, ionization and bremsstrahlung for charged particles.While the spatial discretization is based on the diamond scheme, the angular discretization synergizes finite element method (FEM) and spherical harmonics (SH). Thus, SH is used to globally expand the scattering kernel and FFM is used to locally discretize the angular sphere. As a Result, this hybrid method (FEM-SH) is both accurate in dealing with forward-peaking scattering via FEM, and efficient for multi-energy-group computation via SH. In addition, FEM-SH enables the analytical integration in energy variable of delta scattering kernel for elastic scattering with reduced truncation error from the numerical integration based on the classic SH-based multi-energy-group method. Results: The accuracy of the proposed BFPE solver was benchmarked against Geant4 for photon dose calculation. In particular, FEM-SH had improved accuracy compared to FEM, while both were within 2% of the results obtained with Geant4. Conclusion: A deterministic solver of the Boltzmann-Fokker-Planck equation is developed for dose calculation, and benchmarked against Geant4. Xiang Hong and Hao Gao were partially supported by the NSFC (#11405105), the 973 Program (#2015CB856000) and the Shanghai Pujiang

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.

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

International Nuclear Information System (INIS)

Ganda, F.; Greenspan, A.

2005-01-01

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

Benchmarking Continuum Solvent Models for Keto-Enol Tautomerizations.

Science.gov (United States)

McCann, Billy W; McFarland, Stuart; Acevedo, Orlando

2015-08-13

Experimental free energies of tautomerization, ΔGT, were used to benchmark the gas-phase predictions of 17 different quantum mechanical methods and eight basis sets for seven keto-enol tautomer pairs dominated by their enolic form. The G4 method and M06/6-31+G(d,p) yielded the most accurate results, with mean absolute errors (MAE's) of 0.95 and 0.71 kcal/mol, respectively. Using these two theory levels, the solution-phase ΔGT values for 23 unique tautomer pairs composed of aliphatic ketones, β-dicarbonyls, and heterocycles were computed in multiple protic and aprotic solvents. The continuum solvation models, namely, polarizable continuum model (PCM), polarizable conductor calculation model (CPCM), and universal solvation model (SMD), gave relatively similar MAE's of ∼1.6-1.7 kcal/mol for G4 and ∼1.9-2.0 kcal/mol with M06/6-31+G(d,p). Partitioning the tautomer pairs into their respective molecular types, that is, aliphatic ketones, β-dicarbonyls, and heterocycles, and separating out the aqueous versus nonaqueous results finds G4/PCM utilizing the UA0 cavity to be the overall most accurate combination. Free energies of activation, ΔG(‡), for the base-catalyzed keto-enol interconversion of 2-nitrocyclohexanone were also computed using six bases and five solvents. The M06/6-31+G(d,p) reproduced the ΔG(‡) with MAE's of 1.5 and 1.8 kcal/mol using CPCM and SMD, respectively, for all combinations of base and solvent. That specific enolization was previously proposed to proceed via a concerted mechanism in less polar solvents but shift to a stepwise mechanism in more polar solvents. However, the current calculations suggest that the stepwise mechanism operates in all solvents.

Accurate line intensities of methane from first-principles calculations

Science.gov (United States)

Nikitin, Andrei V.; Rey, Michael; Tyuterev, Vladimir G.

2017-10-01

In this work, we report first-principle theoretical predictions of methane spectral line intensities that are competitive with (and complementary to) the best laboratory measurements. A detailed comparison with the most accurate data shows that discrepancies in integrated polyad intensities are in the range of 0.4%-2.3%. This corresponds to estimations of the best available accuracy in laboratory Fourier Transform spectra measurements for this quantity. For relatively isolated strong lines the individual intensity deviations are in the same range. A comparison with the most precise laser measurements of the multiplet intensities in the 2ν3 band gives an agreement within the experimental error margins (about 1%). This is achieved for the first time for five-atomic molecules. In the Supplementary Material we provide the lists of theoretical intensities at 269 K for over 5000 strongest transitions in the range below 6166 cm-1. The advantage of the described method is that this offers a possibility to generate fully assigned exhaustive line lists at various temperature conditions. Extensive calculations up to 12,000 cm-1 including high-T predictions will be made freely available through the TheoReTS information system (http://theorets.univ-reims.fr, http://theorets.tsu.ru) that contains ab initio born line lists and provides a user-friendly graphical interface for a fast simulation of the absorption cross-sections and radiance.

Three-body calculations at Los Alamos

International Nuclear Information System (INIS)

Friar, J.L.

1986-01-01

This work was motivated by four goals: (1) by working in configuration space, where intuition is greatest, investigate graphically those trinucleon properties which are determined by specific features of wave functions; (2) produce benchmark calculations against which new techniques and numerical methods can be measured; (3) investigate the effect of the Coulomb interaction between the two protons in 3 He and in the p-d system; (4) systematically investigate the various trinucleon observables. Configuration space is particularly well-suited for investigating the Coulomb problem. The singularity and discontinuity problems associated with the Coulomb (momentum space) t-matrix are transformed into boundary condition problems in configuration space. One simply adds the Coulomb potential to the strong interaction. In order to produce accurate numerical solutions powerful techniques were adopted which have not frequently been used in nuclear physics. These spline methods together with collocation techniques combine the power of Gaussian quadrature procedures with the flexibility and strength of finite element approaches to solving partial differential equations. The union of these methods allows one to calculate wavefunctions at the same qualitative level of accuracy as the eigenvalues. Observables can therefore be calculated with considerable confidence. 30 refs., 6 figs

Benchmark study of some thermal and structural computer codes for nuclear shipping casks

International Nuclear Information System (INIS)

Ikushima, Takeshi; Kanae, Yoshioki; Shimada, Hirohisa; Shimoda, Atsumu; Halliquist, J.O.

1984-01-01

There are many computer codes which could be applied to the design and analysis of nuclear material shipping casks. One of problems which the designer of shipping cask faces is the decision regarding the choice of the computer codes to be used. For this situation, the thermal and structural benchmark tests for nuclear shipping casks are carried out to clarify adequacy of the calculation results. The calculation results are compared with the experimental ones. This report describes the results and discussion of the benchmark test. (author)

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.

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.

Fast and accurate inductance and coupling calculation for a multi-layer Nb process

International Nuclear Information System (INIS)

Fourie, Coenrad J; Takahashi, Akitomo; Yoshikawa, Nobuyuki

2015-01-01

Currently, fabrication processes for superconductive integrated circuits are moving to multiple wiring and shielding layers, some of which are placed below the main ground plane (GP) and device layers. The Advanced Industrial Science and Technology advanced process (ADP2) was the first such multi-layer Nb process with planarized passive transmission line and GP layers below the junction layer, and is at the time of writing still the most developed. This process allows complex circuit designs, and accurate inductance extraction helps to push the boundaries of the layouts possible. We show that the position of ground connections between ground layers influences the inductance of structures for which these GPs act as return path, and that this needs to be accounted for in modelling. However, due to the number of wiring layers and GPs, full layout modelling of large cells causes long calculation times. In this paper we discuss methods with which to reduce model size, and calibrate InductEx calculations using these methods against measured results. We show that model reduction followed by calibration results in fast calculation times while good accuracy is maintained. We also show that InductEx correctly handles coupling between conductors in a multi-layer layout, and how to model layouts to gauge unwanted coupling between power lines and single flux quantum electronics. (paper)

Parameters calculation of fuel assembly with complex geometry

International Nuclear Information System (INIS)

Wu Hongchun; Ju Haitao; Yao Dong

2006-01-01

The code DRAGON was developed for CANDU reactor by Ecole Polytechnique de Montreal of Canada. In order to validate the DRAGON code's applicability for complex geometry fuel assembly calculation, the rod shape fuel assembly of PWR benchmark problem and the plate shape fuel assembly of MTR benchmark problem were analyzed by DRAGON code. Some other shape fuel assemblies were also discussed simply. Calculation results show that the DRAGON code can be used to calculate variform fuel assembly and the precision is high. (authors)

NetBenchmark: a bioconductor package for reproducible benchmarks of gene regulatory network inference.

Science.gov (United States)

Bellot, Pau; Olsen, Catharina; Salembier, Philippe; Oliveras-Vergés, Albert; Meyer, Patrick E

2015-09-29

In the last decade, a great number of methods for reconstructing gene regulatory networks from expression data have been proposed. However, very few tools and datasets allow to evaluate accurately and reproducibly those methods. Hence, we propose here a new tool, able to perform a systematic, yet fully reproducible, evaluation of transcriptional network inference methods. Our open-source and freely available Bioconductor package aggregates a large set of tools to assess the robustness of network inference algorithms against different simulators, topologies, sample sizes and noise intensities. The benchmarking framework that uses various datasets highlights the specialization of some methods toward network types and data. As a result, it is possible to identify the techniques that have broad overall performances.

Benchmarking of HEU mental annuli critical assemblies with internally reflected graphite cylinder

Directory of Open Access Journals (Sweden)

Xiaobo Liu

2017-01-01

Full Text Available Three experimental configurations of critical assemblies, performed in 1963 at the Oak Ridge Critical Experiment Facility, which are assembled using three different diameter HEU annuli (15-9 inches, 15-7 inches and 13-7 inches metal annuli with internally reflected graphite cylinder are evaluated and benchmarked. The experimental uncertainties which are 0.00057, 0.00058 and 0.00057 respectively, and biases to the benchmark models which are − 0.00286, − 0.00242 and − 0.00168 respectively, were determined, and the experimental benchmark keff results were obtained for both detailed and simplified models. The calculation results for both detailed and simplified models using MCNP6-1.0 and ENDF/B-VII.1 agree well to the benchmark experimental results within difference less than 0.2%. The benchmarking results were accepted for the inclusion of ICSBEP Handbook.

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Analytical Radiation Transport Benchmarks for The Next Century

International Nuclear Information System (INIS)

Ganapol, B.D.

2005-01-01

Verification of large-scale computational algorithms used in nuclear engineering and radiological applications is an essential element of reliable code performance. For this reason, the development of a suite of multidimensional semi-analytical benchmarks has been undertaken to provide independent verification of proper operation of codes dealing with the transport of neutral particles. The benchmarks considered cover several one-dimensional, multidimensional, monoenergetic and multigroup, fixed source and critical transport scenarios. The first approach, called the Green's Function. In slab geometry, the Green's function is incorporated into a set of integral equations for the boundary fluxes. Through a numerical Fourier transform inversion and subsequent matrix inversion for the boundary fluxes, a semi-analytical benchmark emerges. Multidimensional solutions in a variety of infinite media are also based on the slab Green's function. In a second approach, a new converged SN method is developed. In this method, the SN solution is ''minded'' to bring out hidden high quality solutions. For this case multigroup fixed source and criticality transport problems are considered. Remarkably accurate solutions can be obtained with this new method called the Multigroup Converged SN (MGCSN) method as will be demonstrated

Review and comparison of effective delayed neutron fraction calculation methods with Monte Carlo codes

International Nuclear Information System (INIS)

Bécares, V.; Pérez-Martín, S.; Vázquez-Antolín, M.; Villamarín, D.; Martín-Fuertes, F.; González-Romero, E.M.; Merino, I.

2014-01-01

Highlights: • Review of several Monte Carlo effective delayed neutron fraction calculation methods. • These methods have been implemented with the Monte Carlo code MCNPX. • They have been benchmarked against against some critical and subcritical systems. • Several nuclear data libraries have been used. - Abstract: The calculation of the effective delayed neutron fraction, β eff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for β eff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of β eff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of β eff

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.

Pediatric Academic Productivity: Pediatric Benchmarks for the h- and g-Indices.

Science.gov (United States)

Tschudy, Megan M; Rowe, Tashi L; Dover, George J; Cheng, Tina L

2016-02-01

To describe h- and g-indices benchmarks in pediatric subspecialties and general academic pediatrics. Academic productivity is measured increasingly through bibliometrics that derive a statistical enumeration of academic output and impact. The h- and g-indices incorporate the number of publications and citations. Benchmarks for pediatrics have not been reported. Thirty programs were selected randomly from pediatric residency programs accredited by the Accreditation Council for Graduate Medical Education. The h- and g-indices of department chairs were calculated. For general academic pediatrics, pediatric gastroenterology, and pediatric nephrology, a random sample of 30 programs with fellowships were selected. Within each program, an MD faculty member from each academic rank was selected randomly. Google Scholar via Harzing's Publish or Perish was used to calculate the h-index, g-index, and total manuscripts. Only peer-reviewed and English language publications were included. For Chairs, calculations from Google Scholar were compared with Scopus. For all specialties, the mean h- and g-indices significantly increased with academic rank (all P calculation using different bibliographic databases only differed by ±1. Mean h-indices increased with academic rank and were not significantly different across the pediatric specialties. Benchmarks for h- and g-indices in pediatrics are provided and may be one measure of academic productivity and impact. Copyright © 2016 Elsevier Inc. All rights reserved.

Benchmarks on neutron leakage from iron and beryllium slabs and spheres

International Nuclear Information System (INIS)

Belousov, S.; Ilieva, K.; Popova, I.; Antonov, S.

1998-01-01

Five benchmarks, recommended by the IAEA for nuclear power engineering have been calculated for assessment of Iron and Beryllium neutron data from the recent FENDL version. The FENDL multigroup data processed in IAEA by NJOY code are in VITAMIN-J energy structure in MATXS format. These data have been converted to ANISN format by TRANSX code, ported to PC version and collected to binary library by LIBFENDL code, that we have created especially for this purpose. Neutron transport calculations have been carried out by the codes ANISN, GRTUNCL and DORT. The benchmark for neutron leakage from 25 cm radius Iron sphere with 252 Cf source posses to test FENDL Iron data for LWRs application. Other two benchmarks important for fusion application correspond to the 14 MeV neutron transmission through Iron sphere shell (Simakov S.P. et al., IPPE, Obninsk) and Iron slabs (Y. Oyama, H. Maekawa, FNS/ JAERI). The comparison of measured and calculation results demonstrates discouraged inconsistency when material thickness exceeds 20 cm. Modelling of the angular neutron leakage from 14 MeV source transmitted through Beryllium slabs (H. Maekawa and Y. Oyama at FNS/JAERI) and scalar neutron leakage spectra through sphere shell (Simakov S.P. at al in IPPE, Obninsk) is important for studding the multiplication properties of Beryllium as a fusion blanket material. The measured and calculated results are in relatively good consistency

Calculus of a reactor VVER-1000 benchmark

International Nuclear Information System (INIS)

Dourougie, C.

1998-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-09-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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)

Doppler Temperature Coefficient Calculations Using Adjoint-Weighted Tallies and Continuous Energy Cross Sections in MCNP6

Science.gov (United States)

Gonzales, Matthew Alejandro

version of MCNP6. Temperature feedback results from the cross sections themselves, changes in the probability density functions, as well as changes in the density of the materials. The focus of this work is specific to the Doppler temperature feedback which result from Doppler broadening of cross sections as well as changes in the probability density function within the scattering kernel. This method is compared against published results using Mosteller's numerical benchmark to show accurate evaluations of the Doppler temperature coefficient, fuel assembly calculations, and a benchmark solution based on the heavy gas model for free-gas elastic scattering. An infinite medium benchmark for neutron free gas elastic scattering for large scattering ratios and constant absorption cross section has been developed using the heavy gas model. An exact closed form solution for the neutron energy spectrum is obtained in terms of the confluent hypergeometric function and compared against spectra for the free gas scattering model in MCNP6. Results show a quick increase in convergence of the analytic energy spectrum to the MCNP6 code with increasing target size, showing absolute relative differences of less than 5% for neutrons scattering with carbon. The analytic solution has been generalized to accommodate piecewise constant in energy absorption cross section to produce temperature feedback. Results reinforce the constraints in which heavy gas theory may be applied resulting in a significant target size to accommodate increasing cross section structure. The energy dependent piecewise constant cross section heavy gas model was used to produce a benchmark calculation of the Doppler temperature coefficient to show accurate calculations when using the adjoint-weighted method. Results show the Doppler temperature coefficient using adjoint weighting and cross section derivatives accurately obtains the correct solution within statistics as well as reduce computer runtimes by a factor of 50.

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.

Pynamic: the Python Dynamic Benchmark

Energy Technology Data Exchange (ETDEWEB)

Lee, G L; Ahn, D H; de Supinksi, B R; Gyllenhaal, J C; Miller, P J

2007-07-10

Python is widely used in scientific computing to facilitate application development and to support features such as computational steering. Making full use of some of Python's popular features, which improve programmer productivity, leads to applications that access extremely high numbers of dynamically linked libraries (DLLs). As a result, some important Python-based applications severely stress a system's dynamic linking and loading capabilities and also cause significant difficulties for most development environment tools, such as debuggers. Furthermore, using the Python paradigm for large scale MPI-based applications can create significant file IO and further stress tools and operating systems. In this paper, we present Pynamic, the first benchmark program to support configurable emulation of a wide-range of the DLL usage of Python-based applications for large scale systems. Pynamic has already accurately reproduced system software and tool issues encountered by important large Python-based scientific applications on our supercomputers. Pynamic provided insight for our system software and tool vendors, and our application developers, into the impact of several design decisions. As we describe the Pynamic benchmark, we will highlight some of the issues discovered in our large scale system software and tools using Pynamic.

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

International Nuclear Information System (INIS)

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

2013-01-01

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

CFD validation in OECD/NEA t-junction benchmark.

Energy Technology Data Exchange (ETDEWEB)

Obabko, A. V.; Fischer, P. F.; Tautges, T. J.; Karabasov, S.; Goloviznin, V. M.; Zaytsev, M. A.; Chudanov, V. V.; Pervichko, V. A.; Aksenova, A. E. (Mathematics and Computer Science); (Cambridge Univ.); (Moscow Institute of Nuclar Energy Safety)

2011-08-23

When streams of rapidly moving flow merge in a T-junction, the potential arises for large oscillations at the scale of the diameter, D, with a period scaling as O(D/U), where U is the characteristic flow velocity. If the streams are of different temperatures, the oscillations result in experimental fluctuations (thermal striping) at the pipe wall in the outlet branch that can accelerate thermal-mechanical fatigue and ultimately cause pipe failure. The importance of this phenomenon has prompted the nuclear energy modeling and simulation community to establish a benchmark to test the ability of computational fluid dynamics (CFD) codes to predict thermal striping. The benchmark is based on thermal and velocity data measured in an experiment designed specifically for this purpose. Thermal striping is intrinsically unsteady and hence not accessible to steady state simulation approaches such as steady state Reynolds-averaged Navier-Stokes (RANS) models.1 Consequently, one must consider either unsteady RANS or large eddy simulation (LES). This report compares the results for three LES codes: Nek5000, developed at Argonne National Laboratory (USA), and Cabaret and Conv3D, developed at the Moscow Institute of Nuclear Energy Safety at (IBRAE) in Russia. Nek5000 is based on the spectral element method (SEM), which is a high-order weighted residual technique that combines the geometric flexibility of the finite element method (FEM) with the tensor-product efficiencies of spectral methods. Cabaret is a 'compact accurately boundary-adjusting high-resolution technique' for fluid dynamics simulation. The method is second-order accurate on nonuniform grids in space and time, and has a small dispersion error and computational stencil defined within one space-time cell. The scheme is equipped with a conservative nonlinear correction procedure based on the maximum principle. CONV3D is based on the immersed boundary method and is validated on a wide set of the experimental

Monte Carlo benchmark calculations of energy deposition by electron/photon showers up to 1 GeV

International Nuclear Information System (INIS)

Mehlhorn, T.A.; Halbleib, J.A.

1983-01-01

Over the past several years the TIGER series of coupled electron/photon Monte Carlo transport codes has been applied to a variety of problems involving nuclear and space radiations, electron accelerators, and radioactive sources. In particular, they have been used at Sandia to simulate the interaction of electron beams, generated by pulsed-power accelerators, with various target materials for weapons effect simulation, and electron beam fusion. These codes are based on the ETRAN system which was developed for an energy range from about 10 keV up to a few tens of MeV. In this paper we will discuss the modifications that were made to the TIGER series of codes in order to extend their applicability to energies of interest to the high energy physics community (up to 1 GeV). We report the results of a series of benchmark calculations of the energy deposition by high energy electron beams in various materials using the modified codes. These results are then compared with the published results of various experimental measurements and other computational models

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…

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

Thermal hydraulic calculation of STORM facility using GOTHIC code

International Nuclear Information System (INIS)

Pevec, D.; Grgic, D.; Prah, M.

1995-01-01

Benchmark calculation CTI defined in frame of STORM experimental programme is used to prove that the GOTHIC code is capable to predict behaviour of experimental facility with reasonable accuracy. GOTHIC code is developed mainly for containment calculation. In this situation it is successfully used for calculation of one dimensional flow of steam and noncondensable mixture. Steady state distributions of pressure, temperature and the velocity of gas along facility are consistent with results obtained by other benchmark participants. (author)

Accurate Calculations of Rotationally Inelastic Scattering Cross Sections Using Mixed Quantum/Classical Theory.

Science.gov (United States)

Semenov, Alexander; Babikov, Dmitri

2014-01-16

For computational treatment of rotationally inelastic scattering of molecules, we propose to use the mixed quantum/classical theory, MQCT. The old idea of treating translational motion classically, while quantum mechanics is used for rotational degrees of freedom, is developed to the new level and is applied to Na + N2 collisions in a broad range of energies. Comparison with full-quantum calculations shows that MQCT accurately reproduces all, even minor, features of energy dependence of cross sections, except scattering resonances at very low energies. The remarkable success of MQCT opens up wide opportunities for computational predictions of inelastic scattering cross sections at higher temperatures and/or for polyatomic molecules and heavier quenchers, which is computationally close to impossible within the full-quantum framework.

Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

International Nuclear Information System (INIS)

Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

2003-01-01

All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important

Benchmarking of nuclear economics tools

International Nuclear Information System (INIS)

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

2017-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-15

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

Computational scheme for pH-dependent binding free energy calculation with explicit solvent.

Science.gov (United States)

Lee, Juyong; Miller, Benjamin T; Brooks, Bernard R

2016-01-01

We present a computational scheme to compute the pH-dependence of binding free energy with explicit solvent. Despite the importance of pH, the effect of pH has been generally neglected in binding free energy calculations because of a lack of accurate methods to model it. To address this limitation, we use a constant-pH methodology to obtain a true ensemble of multiple protonation states of a titratable system at a given pH and analyze the ensemble using the Bennett acceptance ratio (BAR) method. The constant pH method is based on the combination of enveloping distribution sampling (EDS) with the Hamiltonian replica exchange method (HREM), which yields an accurate semi-grand canonical ensemble of a titratable system. By considering the free energy change of constraining multiple protonation states to a single state or releasing a single protonation state to multiple states, the pH dependent binding free energy profile can be obtained. We perform benchmark simulations of a host-guest system: cucurbit[7]uril (CB[7]) and benzimidazole (BZ). BZ experiences a large pKa shift upon complex formation. The pH-dependent binding free energy profiles of the benchmark system are obtained with three different long-range interaction calculation schemes: a cutoff, the particle mesh Ewald (PME), and the isotropic periodic sum (IPS) method. Our scheme captures the pH-dependent behavior of binding free energy successfully. Absolute binding free energy values obtained with the PME and IPS methods are consistent, while cutoff method results are off by 2 kcal mol(-1) . We also discuss the characteristics of three long-range interaction calculation methods for constant-pH simulations. © 2015 The Protein Society.

Cross sections, benchmarks, etc.: What is data testing all about

International Nuclear Information System (INIS)

Wagschal, J.; Yeivin, Y.

1985-01-01

In order to determine the consistency of two distinct measurements of a physical quantity, the discrepancy d between the two should be compared with its own standard deviation, σ = √(σ/sub 1//sup 2/+σ/sub 2//sup 2/). To properly test a given cross-section library by a set of benchmark (integral) measurements, the quantity corresponding to (d/σ)/sup 2/ is the quadratic d/sup dagger/C/sup -1/d. Here d is the vector of which the components are the discrepancies between the calculated values of the integral parameters and their corresponding measured values, and C is the uncertainty matrix of these discrepancies. This quadratic form is the only true measure of the joint consistency of the library and benchmarks. On the other hand, the very matrix C is essentially all one needs to adjust the library by the benchmarks. Therefore, any argument against adjustment simultaneously disqualifies all serious attempts to test cross-section libraries against integral benchmarks

CONTAIN calculations

International Nuclear Information System (INIS)

Scholtyssek, W.

1995-01-01

In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident 'medium-sized leak in the cold leg', especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

Accurate calculation of high harmonics generated by relativistic Thomson scattering

International Nuclear Information System (INIS)

Popa, Alexandru

2008-01-01

The recent emergence of the field of ultraintense laser pulses, corresponding to beam intensities higher than 10 18 W cm -2 , brings about the problem of the high harmonic generation (HHG) by the relativistic Thomson scattering of the electromagnetic radiation by free electrons. Starting from the equations of the relativistic motion of the electron in the electromagnetic field, we give an exact solution of this problem. Taking into account the Lienard-Wiechert equations, we obtain a periodic scattered electromagnetic field. Without loss of generality, the solution is strongly simplified by observing that the electromagnetic field is always normal to the direction electron-detector. The Fourier series expansion of this field leads to accurate expressions of the high harmonics generated by the Thomson scattering. Our calculations lead to a discrete HHG spectrum, whose shape and angular distribution are in agreement with the experimental data from the literature. Since no approximations were made, our approach is also valid in the ultrarelativistic regime, corresponding to intensities higher than 10 23 W cm -2 , where it predicts a strong increase of the HHG intensities and of the order of harmonics. In this domain, the nonlinear Thomson scattering could be an efficient source of hard x-rays

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)

Benchmark comparisons of evaluated nuclear data files

International Nuclear Information System (INIS)

Resler, D.A.; Howerton, R.J.; White, R.M.

1994-05-01

With the availability and maturity of several evaluated nuclear data files, it is timely to compare the results of integral tests with calculations using these different files. We discuss here our progress in making integral benchmark tests of the following nuclear data files: ENDL-94, ENDF/B-V and -VI, JENDL-3, JEF-2, and BROND-2. The methods used to process these evaluated libraries in a consistent way into applications files for use in Monte Carlo calculations is presented. Using these libraries, we are calculating and comparing to experiment k eff for 68 fast critical assemblies of 233,235 U and 239 Pu with reflectors of various material and thickness

Specification for the VERA Depletion Benchmark Suite

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-12-17

CASL-X-2015-1014-000 iii Consortium for Advanced Simulation of LWRs EXECUTIVE SUMMARY The CASL neutronics simulator MPACT is under development for the neutronics and T-H coupled simulation for the pressurized water reactor. MPACT includes the ORIGEN-API and internal depletion module to perform depletion calculations based upon neutron-material reaction and radioactive decay. It is a challenge to validate the depletion capability because of the insufficient measured data. One of the detoured methods to validate it is to perform a code-to-code comparison for benchmark problems. In this study a depletion benchmark suite has been developed and a detailed guideline has been provided to obtain meaningful computational outcomes which can be used in the validation of the MPACT depletion capability.

Experimental and computational benchmark tests

International Nuclear Information System (INIS)

Gilliam, D.M.; Briesmeister, J.F.

1994-01-01

A program involving principally NIST, LANL, and ORNL has been in progress for about four years now to establish a series of benchmark measurements and calculations related to the moderation and leakage of 252 Cf neutrons from a source surrounded by spherical aqueous moderators of various thicknesses and compositions. The motivation for these studies comes from problems in criticality calculations concerning arrays of multiplying components, where the leakage from one component acts as a source for the other components. This talk compares experimental and calculated values for the fission rates of four nuclides - 235 U, 239 Pu, 238 U, and 237 Np - in the leakage spectrum from moderator spheres of diameters 76.2 mm, 101.6 mm, and 127.0 mm, with either pure water or enriched B-10 solutions as the moderator. Very detailed Monte Carlo calculations were done with the MCNP code, using a open-quotes light waterclose quotes S(α,β) scattering kernel

International benchmark tests of the FENDL-1 Nuclear Data Library

International Nuclear Information System (INIS)

Fischer, U.

1997-01-01

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

Benchmarking electricity distribution

Energy Technology Data Exchange (ETDEWEB)

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

1995-12-31

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

World-Wide Benchmarking of ITER Nb$_{3}$Sn Strand Test Facilities

CERN Document Server

Jewell, MC; Takahashi, Yoshikazu; Shikov, Alexander; Devred, Arnaud; Vostner, Alexander; Liu, Fang; Wu, Yu; Jewell, Matthew C; Boutboul, Thierry; Bessette, Denis; Park, Soo-Hyeon; Isono, Takaaki; Vorobieva, Alexandra; Martovetsky, Nicolai; Seo, Kazutaka

2010-01-01

The world-wide procurement of Nb$_{3}$Sn and NbTi for the ITER superconducting magnet systems will involve eight to ten strand suppliers from six Domestic Agencies (DAs) on three continents. To ensure accurate and consistent measurement of the physical and superconducting properties of the composite strand, a strand test facility benchmarking effort was initiated in August 2008. The objectives of this effort are to assess and improve the superconducting strand test and sample preparation technologies at each DA and supplier, in preparation for the more than ten thousand samples that will be tested during ITER procurement. The present benchmarking includes tests for critical current (I-c), n-index, hysteresis loss (Q(hys)), residual resistivity ratio (RRR), strand diameter, Cu fraction, twist pitch, twist direction, and metal plating thickness (Cr or Ni). Nineteen participants from six parties (China, EU, Japan, South Korea, Russia, and the United States) have participated in the benchmarking. This round, cond...

Geometric constraints in semiclassical initial value representation calculations in Cartesian coordinates: accurate reduction in zero-point energy.

Science.gov (United States)

Issack, Bilkiss B; Roy, Pierre-Nicholas

2005-08-22

An approach for the inclusion of geometric constraints in semiclassical initial value representation calculations is introduced. An important aspect of the approach is that Cartesian coordinates are used throughout. We devised an algorithm for the constrained sampling of initial conditions through the use of multivariate Gaussian distribution based on a projected Hessian. We also propose an approach for the constrained evaluation of the so-called Herman-Kluk prefactor in its exact log-derivative form. Sample calculations are performed for free and constrained rare-gas trimers. The results show that the proposed approach provides an accurate evaluation of the reduction in zero-point energy. Exact basis set calculations are used to assess the accuracy of the semiclassical results. Since Cartesian coordinates are used, the approach is general and applicable to a variety of molecular and atomic systems.

The International Criticality Safety Benchmark Evaluation Project

International Nuclear Information System (INIS)

Briggs, B. J.; Dean, V. F.; Pesic, M. P.

2001-01-01

In order to properly manage the risk of a nuclear criticality accident, it is important to establish the conditions for which such an accident becomes possible for any activity involving fissile material. Only when this information is known is it possible to establish the likelihood of actually achieving such conditions. It is therefore important that criticality safety analysts have confidence in the accuracy of their calculations. Confidence in analytical results can only be gained through comparison of those results with experimental data. The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the US Department of Energy. The project was managed through the Idaho National Engineering and Environmental Laboratory (INEEL), but involved nationally known criticality safety experts from Los Alamos National Laboratory, Lawrence Livermore National Laboratory, Savannah River Technology Center, Oak Ridge National Laboratory and the Y-12 Plant, Hanford, Argonne National Laboratory, and the Rocky Flats Plant. An International Criticality Safety Data Exchange component was added to the project during 1994 and the project became what is currently known as the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Representatives from the United Kingdom, France, Japan, the Russian Federation, Hungary, Kazakhstan, Korea, Slovenia, Yugoslavia, Spain, and Israel are now participating on the project In December of 1994, the ICSBEP became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency's (OECD-NEA) Nuclear Science Committee. The United States currently remains the lead country, providing most of the administrative support. The purpose of the ICSBEP is to: (1) identify and evaluate a comprehensive set of critical benchmark data; (2) verify the data, to the extent possible, by reviewing original and subsequently revised documentation, and by talking with the

Development of an ICSBEP Benchmark Evaluation, Nearly 20 Years of Experience

International Nuclear Information System (INIS)

Briggs, J. Blair; Bess, John D.

2011-01-01

The basic structure of all ICSBEP benchmark evaluations is essentially the same and includes (1) a detailed description of the experiment; (2) an evaluation of the experiment, including an exhaustive effort to quantify the effects of uncertainties on measured quantities; (3) a concise presentation of benchmark-model specifications; (4) sample calculation results; and (5) a summary of experimental references. Computer code input listings and other relevant information are generally preserved in appendixes. Details of an ICSBEP evaluation is presented.

Accurate effective temperatures of the metal-poor benchmark stars HD 140283, HD 122563, and HD 103095 from CHARA interferometry

Science.gov (United States)

Karovicova, I.; White, T. R.; Nordlander, T.; Lind, K.; Casagrande, L.; Ireland, M. J.; Huber, D.; Creevey, O.; Mourard, D.; Schaefer, G. H.; Gilmore, G.; Chiavassa, A.; Wittkowski, M.; Jofré, P.; Heiter, U.; Thévenin, F.; Asplund, M.

2018-03-01

Large stellar surveys of the Milky Way require validation with reference to a set of `benchmark' stars whose fundamental properties are well determined. For metal-poor benchmark stars, disagreement between spectroscopic and interferometric effective temperatures has called the reliability of the temperature scale into question. We present new interferometric measurements of three metal-poor benchmark stars, HD 140283, HD 122563, and HD 103095, from which we determine their effective temperatures. The angular sizes of all the stars were determined from observations with the PAVO beam combiner at visible wavelengths at the CHARA array, with additional observations of HD 103095 made with the VEGA instrument, also at the CHARA array. Together with photometrically derived bolometric fluxes, the angular diameters give a direct measurement of the effective temperature. For HD 140283, we find θLD = 0.324 ± 0.005 mas, Teff = 5787 ± 48 K; for HD 122563, θLD = 0.926 ± 0.011 mas, Teff = 4636 ± 37 K; and for HD 103095, θLD = 0.595 ± 0.007 mas, Teff = 5140 ± 49 K. Our temperatures for HD 140283 and HD 103095 are hotter than the previous interferometric measurements by 253 and 322 K, respectively. We find good agreement between our temperatures and recent spectroscopic and photometric estimates. We conclude some previous interferometric measurements have been affected by systematic uncertainties larger than their quoted errors.

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:

Accurate thermodynamic relations of the melting temperature of nanocrystals with different shapes and pure theoretical calculation

Energy Technology Data Exchange (ETDEWEB)

Zhu, Jinhua; Fu, Qingshan; Xue, Yongqiang, E-mail: xyqlw@126.com; Cui, Zixiang

2017-05-01

Based on the surface pre-melting model, accurate thermodynamic relations of the melting temperature of nanocrystals with different shapes (tetrahedron, cube, octahedron, dodecahedron, icosahedron, nanowire) were derived. The theoretically calculated melting temperatures are in relative good agreements with experimental, molecular dynamic simulation and other theoretical results for nanometer Au, Ag, Al, In and Pb. It is found that the particle size and shape have notable effects on the melting temperature of nanocrystals, and the smaller the particle size, the greater the effect of shape. Furthermore, at the same equivalent radius, the more the shape deviates from sphere, the lower the melting temperature is. The value of melting temperature depression of cylindrical nanowire is just half of that of spherical nanoparticle with an identical radius. The theoretical relations enable one to quantitatively describe the influence regularities of size and shape on the melting temperature and to provide an effective way to predict and interpret the melting temperature of nanocrystals with different sizes and shapes. - Highlights: • Accurate relations of T{sub m} of nanocrystals with various shapes are derived. • Calculated T{sub m} agree with literature results for nano Au, Ag, Al, In and Pb. • ΔT{sub m} (nanowire) = 0.5ΔT{sub m} (spherical nanocrystal). • The relations apply to predict and interpret the melting behaviors of nanocrystals.

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

Energy Technology Data Exchange (ETDEWEB)

Kuijper, J.C.

1994-10-01

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

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

International Nuclear Information System (INIS)

Kuijper, J.C.

1996-10-01

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

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

International Nuclear Information System (INIS)

Kuijper, J.C.

1994-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Kuijper, J.C.

1996-10-01

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

RUNE benchmarks

DEFF Research Database (Denmark)

Peña, Alfredo

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

Benchmarking of Monte Carlo simulation of bremsstrahlung from thick targets at radiotherapy energies

International Nuclear Information System (INIS)

Faddegon, Bruce A.; Asai, Makoto; Perl, Joseph; Ross, Carl; Sempau, Josep; Tinslay, Jane; Salvat, Francesc

2008-01-01

Several Monte Carlo systems were benchmarked against published measurements of bremsstrahlung yield from thick targets for 10-30 MV beams. The quantity measured was photon fluence at 1 m per unit energy per incident electron (spectra), and total photon fluence, integrated over energy, per incident electron (photon yield). Results were reported at 10-30 MV on the beam axis for Al and Pb targets and at 15 MV at angles out to 90 degree sign for Be, Al, and Pb targets. Beam energy was revised with improved accuracy of 0.5% using an improved energy calibration of the accelerator. Recently released versions of the Monte Carlo systems EGSNRC, GEANT4, and PENELOPE were benchmarked against the published measurements using the revised beam energies. Monte Carlo simulation was capable of calculation of photon yield in the experimental geometry to 5% out to 30 degree sign , 10% at wider angles, and photon spectra to 10% at intermediate photon energies, 15% at lower energies. Accuracy of measured photon yield from 0 to 30 degree sign was 5%, 1 s.d., increasing to 7% for the larger angles. EGSNRC and PENELOPE results were within 2 s.d. of the measured photon yield at all beam energies and angles, GEANT4 within 3 s.d. Photon yield at nonzero angles for angles covering conventional field sizes used in radiotherapy (out to 10 degree sign ), measured with an accuracy of 3%, was calculated within 1 s.d. of measurement for EGSNRC, 2 s.d. for PENELOPE and GEANT4. Calculated spectra closely matched measurement at photon energies over 5 MeV. Photon spectra near 5 MeV were underestimated by as much as 10% by all three codes. The photon spectra below 2-3 MeV for the Be and Al targets and small angles were overestimated by up to 15% when using EGSNRC and PENELOPE, 20% with GEANT4. EGSNRC results with the NIST option for the bremsstrahlung cross section were preferred over the alternative cross section available in EGSNRC and over EGS4. GEANT4 results calculated with the ''low energy

Benchmarking of Monte Carlo simulation of bremsstrahlung from thick targets at radiotherapy energies

Energy Technology Data Exchange (ETDEWEB)

Faddegon, Bruce A.; Asai, Makoto; Perl, Joseph; Ross, Carl; Sempau, Josep; Tinslay, Jane; Salvat, Francesc [Department of Radiation Oncology, University of California at San Francisco, San Francisco, California 94143 (United States); Stanford Linear Accelerator Center, 2575 Sand Hill Road, Menlo Park, California 94025 (United States); National Research Council Canada, Institute for National Measurement Standards, 1200 Montreal Road, Building M-36, Ottawa, Ontario K1A 0R6 (Canada); Institut de Tecniques Energetiques, Universitat Politecnica de Catalunya and Centro de Investigacion Biomedica en Red en Bioingenieria, Biomateriales y Nanomedicina (CIBER-BBN), Diagonal 647, 08028 Barcelona (Spain); Stanford Linear Accelerator Center, 2575 Sand Hill Road, Menlo Park, California 94025 (United States); Facultat de Fisica (ECM), Universitat de Barcelona, Societat Catalana de Fisica (IEC), Diagonal 647, 08028 Barcelona (Spain)

2008-10-15

Several Monte Carlo systems were benchmarked against published measurements of bremsstrahlung yield from thick targets for 10-30 MV beams. The quantity measured was photon fluence at 1 m per unit energy per incident electron (spectra), and total photon fluence, integrated over energy, per incident electron (photon yield). Results were reported at 10-30 MV on the beam axis for Al and Pb targets and at 15 MV at angles out to 90 degree sign for Be, Al, and Pb targets. Beam energy was revised with improved accuracy of 0.5% using an improved energy calibration of the accelerator. Recently released versions of the Monte Carlo systems EGSNRC, GEANT4, and PENELOPE were benchmarked against the published measurements using the revised beam energies. Monte Carlo simulation was capable of calculation of photon yield in the experimental geometry to 5% out to 30 degree sign , 10% at wider angles, and photon spectra to 10% at intermediate photon energies, 15% at lower energies. Accuracy of measured photon yield from 0 to 30 degree sign was 5%, 1 s.d., increasing to 7% for the larger angles. EGSNRC and PENELOPE results were within 2 s.d. of the measured photon yield at all beam energies and angles, GEANT4 within 3 s.d. Photon yield at nonzero angles for angles covering conventional field sizes used in radiotherapy (out to 10 degree sign ), measured with an accuracy of 3%, was calculated within 1 s.d. of measurement for EGSNRC, 2 s.d. for PENELOPE and GEANT4. Calculated spectra closely matched measurement at photon energies over 5 MeV. Photon spectra near 5 MeV were underestimated by as much as 10% by all three codes. The photon spectra below 2-3 MeV for the Be and Al targets and small angles were overestimated by up to 15% when using EGSNRC and PENELOPE, 20% with GEANT4. EGSNRC results with the NIST option for the bremsstrahlung cross section were preferred over the alternative cross section available in EGSNRC and over EGS4. GEANT4 results calculated with the &apos

Within-Group Effect-Size Benchmarks for Trauma-Focused Cognitive Behavioral Therapy with Children and Adolescents

Science.gov (United States)

Rubin, Allen; Washburn, Micki; Schieszler, Christine

2017-01-01

Purpose: This article provides benchmark data on within-group effect sizes from published randomized clinical trials (RCTs) supporting the efficacy of trauma-focused cognitive behavioral therapy (TF-CBT) for traumatized children. Methods: Within-group effect-size benchmarks for symptoms of trauma, anxiety, and depression were calculated via the…

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-15

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

Check and visualization of input geometry data using the geometrical module of the Monte Carlo code MCU: WWER-440 pressure vessel dosimetry benchmarks

International Nuclear Information System (INIS)

Gurevich, M.; Zaritsky, S.; Osmera, B.; Mikus, J.

1997-01-01

The Monte Carlo method gives the opportunity to conduct the calculations of neutron and photon flux without any simplifications of the 3-D geometry of the nuclear power and experimental devices. So, each graduated Monte Carlo code includes the combinatorial geometry module and tools for the geometry description giving a possibility to describe very complex systems with a number of hierarchy levels of the geometrical objects. Such codes as usual have special modules for the visual checking of geometry input information. These geometry opportunities could be used for all cases when the accurate 3-D description of the complex geometry becomes a necessity. The description (specification) of benchmark experiments is one of the such cases. Such accurate and uniform description detects all mistakes and ambiguities in the starting information of various kinds (drawings, reports etc.). Usually the quality of different parts of the starting information (generally produced by different persons during the different stages of the device elaboration and operation) is different. After using the above mentioned modules and tools, the resultant geometry description can be used as a standard for this device. One can automatically produce any type of the device figure. The detail geometry description can be used as input for different calculation models carrying out (not only for Monte Carlo). The application of that method to the description of the WWER-440 mock-ups is represented in the report. The mock-ups were created on the reactor LR-O (NRI) and the reactor vessel dosimetry benchmarks were developed on the basis of these mock-up experiments. The NCG-8 module of the Russian Monte Carlo code MCU was used. It is the combinatorial multilingual universal geometrical module. The MCU code was certified by Russian Nuclear Regulatory Body. Almost all figures for mentioned benchmarks specifications were made by the MCU visualization code. The problem of the automatic generation of the

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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