A photon dominated region code comparison study

NARCIS (Netherlands)

Roellig, M.; Abel, N. P.; Bell, T.; Bensch, F.; Black, J.; Ferland, G. J.; Jonkheid, B.; Kamp, I.; Kaufman, M. J.; Le Bourlot, J.; Le Petit, F.; Meijerink, R.; Morata, O.; Ossenkopf, Volker; Roueff, E.; Shaw, G.; Spaans, M.; Sternberg, A.; Stutzki, J.; Thi, W.-F.; van Dishoeck, E. F.; van Hoof, P. A. M.; Viti, S.; Wolfire, M. G.

Aims. We present a comparison between independent computer codes, modeling the physics and chemistry of interstellar photon dominated regions (PDRs). Our goal was to understand the mutual differences in the PDR codes and their effects on the physical and chemical structure of the model clouds, and

Offshore Code Comparison Collaboration within IEA Wind Task 23: Phase IV Results Regarding Floating Wind Turbine Modeling; Preprint

Energy Technology Data Exchange (ETDEWEB)

Jonkman, J.; Larsen, T.; Hansen, A.; Nygaard, T.; Maus, K.; Karimirad, M.; Gao, Z.; Moan, T.; Fylling, I.

2010-04-01

Offshore wind turbines are designed and analyzed using comprehensive simulation codes that account for the coupled dynamics of the wind inflow, aerodynamics, elasticity, and controls of the turbine, along with the incident waves, sea current, hydrodynamics, and foundation dynamics of the support structure. This paper describes the latest findings of the code-to-code verification activities of the Offshore Code Comparison Collaboration, which operates under Subtask 2 of the International Energy Agency Wind Task 23. In the latest phase of the project, participants used an assortment of codes to model the coupled dynamic response of a 5-MW wind turbine installed on a floating spar buoy in 320 m of water. Code predictions were compared from load-case simulations selected to test different model features. The comparisons have resulted in a greater understanding of offshore floating wind turbine dynamics and modeling techniques, and better knowledge of the validity of various approximations. The lessons learned from this exercise have improved the participants' codes, thus improving the standard of offshore wind turbine modeling.

Comparison of beam deposition for three neutral beam injection codes

International Nuclear Information System (INIS)

Wieland, R.M.; Houlberg, W.A.; Mense, A.T.

1979-03-01

The three neutral beam injection codes BEAM (Houlberg, ORNL), HOFR (Howe, ORNL), and FREYA (Post, PPPL) are compared with respect to the calculation of the fast ion deposition profile H(r). Only plasmas of circular cross section are considered, with injection confined to the mid-plane of the torus. The approximations inherent in each code are pointed out, and a series of comparisons varying several parameters (beam energy and radius, machine size, and injection angle) shows excellent agreement among all the codes. A cost comparison (execution time and memory requirements) is made which points out the relative merits of each code within the context of incorporation into a plasma transport simulation code

Intracoin - International Nuclide Transport Code Intercomparison Study

International Nuclear Information System (INIS)

1984-09-01

The purpose of the project is to obtain improved knowledge of the influence of various strategies for radionuclide transport modelling for the safety assessment of final repositories for nuclear waste. This is a report of the first phase of the project which was devoted to a comparison of the numerical accuracy of the computer codes used in the study. The codes can be divided into five groups, namely advection-dispersion models, models including matrix diffusion and chemical effects and finally combined models. The results are presented as comparisons of calculations since the objective of level 1 was code verification. (G.B.)

Code comparison results for the loft LP-FP-2 experiment

International Nuclear Information System (INIS)

Merilo, M.; Mecham, D.C.

1991-01-01

Computer code calculations are compared with thermal hydraulic and fission product release, transport, and deposition data obtained from the OECD-LOFT LP-FP-2 experiment. Except for the MAAP code, which is a fully integrated severe accident code, the thermalhydraulic and fission product behavior were calculated with different codes. Six organizations participated in the thermal hydraulic portion of the code comparison exercise. These calculations were performed with RELAP 5, SCDAP/RELAP 5, and MAAP. The comparisons show generally well developed capabilities to determine the thermal-hydraulic conditions during the early stages of a severe core damage accident. Four participants submitted detailed fission product behavior calculations. Except for MAAP, as stated previously, the fission product inventory, core damage, fission product release, transport and deposition were calculated independently with different codes. Much larger differences than observed for the thermalhydraulic comparison were evident. The fission product inventory calculations were generally in good agreement with each other. Large differences were observed for release fractions and amounts of deposition. Net release calculations from the primary system were generally accurate within a factor of two or three for the more important fission products

Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data

International Nuclear Information System (INIS)

2012-08-01

Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and cooperative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled Intercomparison and Validation of Computer Codes for Thermalhydraulics Safety Analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. Two RD-14M small break loss of coolant accident (SBLOCA) tests, simulating HWR LOCA behaviour, conducted by Atomic Energy of Canada Ltd (AECL), were selected for this validation project. This report provides a comparison of the results obtained from eight participating organizations from six countries (Argentina, Canada, China, India, Republic of Korea, and Romania), utilizing four different computer codes (ATMIKA, CATHENA, MARS-KS, and RELAP5). General conclusions are reached and recommendations made.

A bar coding system for environmental projects

International Nuclear Information System (INIS)

Barber, R.B.; Hunt, B.J.; Burgess, G.M.

1988-01-01

This paper presents BeCode systems, a bar coding system which provides both nuclear and commercial clients with a data capture and custody management program that is accurate, timely, and beneficial to all levels of project operations. Using bar code identifiers is an essentially paperless and error-free method which provides more efficient delivery of data through its menu card-driven structure, which speeds collection of essential data for uploading to a compatible device. The effects of this sequence include real-time information for operator analysis, management review, audits, planning, scheduling, and cost control

Comparison of neutron spectrum unfolding codes

International Nuclear Information System (INIS)

Zijp, W.

1979-02-01

This final report contains a set of four ECN-reports. The first is dealing with the comparison of the neutron spectrum unfolding codes CRYSTAL BALL, RFSP-JUL, SAND II and STAY'SL. The other three present the results of calculations about the influence of statistical weights in CRYSTAL BALL, SAND II and RFSP-JUL

Comparison of computer code calculations with FEBA test data

International Nuclear Information System (INIS)

Zhu, Y.M.

1988-06-01

The FEBA forced feed reflood experiments included base line tests with unblocked geometry. The experiments consisted of separate effect tests on a full-length 5x5 rod bundle. Experimental cladding temperatures and heat transfer coefficients of FEBA test No. 216 are compared with the analytical data postcalculated utilizing the SSYST-3 computer code. The comparison indicates a satisfactory matching of the peak cladding temperatures, quench times and heat transfer coefficients for nearly all axial positions. This agreement was made possible by the use of an artificially adjusted value of the empirical code input parameter in the heat transfer for the dispersed flow regime. A limited comparison of test data and calculations using the RELAP4/MOD6 transient analysis code are also included. In this case the input data for the water entrainment fraction and the liquid weighting factor in the heat transfer for the dispersed flow regime were adjusted to match the experimental data. On the other hand, no fitting of the input parameters was made for the COBRA-TF calculations which are included in the data comparison. (orig.) [de

Project Everware - running other people's code doesn't have to be painful

CERN Multimedia

CERN. Geneva

2015-01-01

Everware is a project that allows you to edit and run someone else's code with one click, even if that code has complicated setup instructions. The main aim of the project is to encourage reuse of software between researchers by making it easy and risk free to try out someone else's code.

THE EXPERIENCE OF COMPARISON OF STATIC SECURITY CODE ANALYZERS

Directory of Open Access Journals (Sweden)

Alexey Markov

2015-09-01

Full Text Available This work presents a methodological approach to comparison of static security code analyzers. It substantiates the comparison of the static analyzers as to efficiency and functionality indicators, which are stipulated in the international regulatory documents. The test data for assessment of static analyzers efficiency is represented by synthetic sets of open-source software, which contain vulnerabilities. We substantiated certain criteria for quality assessment of the static security code analyzers subject to standards NIST SP 500-268 and SATEC. We carried out experiments that allowed us to assess a number of the Russian proprietary software tools and open-source tools. We came to the conclusion that it is of paramount importance to develop Russian regulatory framework for testing software security (firstly, for controlling undocumented features and evaluating the quality of static security code analyzers.

Reconciliation of international administrative coding systems for comparison of colorectal surgery outcome.

Science.gov (United States)

Munasinghe, A; Chang, D; Mamidanna, R; Middleton, S; Joy, M; Penninckx, F; Darzi, A; Livingston, E; Faiz, O

2014-07-01

Significant variation in colorectal surgery outcomes exists between different countries. Better understanding of the sources of variable outcomes using administrative data requires alignment of differing clinical coding systems. We aimed to map similar diagnoses and procedures across administrative coding systems used in different countries. Administrative data were collected in a central database as part of the Global Comparators (GC) Project. In order to unify these data, a systematic translation of diagnostic and procedural codes was undertaken. Codes for colorectal diagnoses, resections, operative complications and reoperative interventions were mapped across the respective national healthcare administrative coding systems. Discharge data from January 2006 to June 2011 for patients who had undergone colorectal surgical resections were analysed to generate risk-adjusted models for mortality, length of stay, readmissions and reoperations. In all, 52 544 case records were collated from 31 institutions in five countries. Mapping of all the coding systems was achieved so that diagnosis and procedures from the participant countries could be compared. Using the aligned coding systems to develop risk-adjusted models, the 30-day mortality rate for colorectal surgery was 3.95% (95% CI 0.86-7.54), the 30-day readmission rate was 11.05% (5.67-17.61), the 28-day reoperation rate was 6.13% (3.68-9.66) and the mean length of stay was 14 (7.65-46.76) days. The linkage of international hospital administrative data that we developed enabled comparison of documented surgical outcomes between countries. This methodology may facilitate international benchmarking. Colorectal Disease © 2014 The Association of Coloproctology of Great Britain and Ireland.

Comparison of Bit Error Rate of Line Codes in NG-PON2

Directory of Open Access Journals (Sweden)

Tomas Horvath

2016-05-01

Full Text Available This article focuses on simulation and comparison of line codes NRZ (Non Return to Zero, RZ (Return to Zero and Miller’s code for NG-PON2 (Next-Generation Passive Optical Network Stage 2 using. Our article provides solutions with Q-factor, BER (Bit Error Rate, and bandwidth comparison. Line codes are the most important part of communication over the optical fibre. The main role of these codes is digital signal representation. NG-PON2 networks use optical fibres for communication that is the reason why OptSim v5.2 is used for simulation.

Analysis of SCARABEE BE+3 experiment with ASTEC-Na and comparison with other SFR safety analysis codes

International Nuclear Information System (INIS)

Bandini, Giacomino; Ederli, Stefano; Perez-Martin, Sara; Pfrang, Werner; Girault, Nathalie; Cloarec, Laure

2017-01-01

The ASTEC-Na code was further developed and assessed in the frame of JASMIN project of the 7th EU Framework Program to extend the original capability of ASTEC, dealing with severe accident analysis in LWR to Sodium-cooled Fast Reactors (SFR). The in-pile BE+3 experiment from the SCARABEE-N program has been simulated with ASTEC-Na for thermal-hydraulic models validation purpose. The adequacy of ASTEC-Na thermal-hydraulic models has been also investigated through the comparison with other safety analysis codes. The analysis of SCARABEE BE+3 test confirms the good performance of ASTEC-Na code in the calculation of single-phase conditions and boiling onset, while larger deviations are encountered in the analysis of the two-phase conditions, mainly regarding the propagation of the boiling front. Furthermore, reasonable agreement was found with other code results. (author)

Code of Conduct for wind-power projects - Feasibility study

International Nuclear Information System (INIS)

Strub, P.; Ziegler, Ch.

2009-02-01

This final report deals with the results of a feasibility study concerning the development of a Code of Conduct for wind-power projects. The aim is to strengthen the acceptance of wind-power by the general public. The necessity of new, voluntary market instruments is discussed. The urgency of development in this area is quoted as being high, and the authors consider the feasibility of the definition of a code of conduct as being proven. The code of conduct can, according to the authors, be of use at various levels but primarily in project development. Further free-enterprise instruments are also suggested that should help support socially compatible and successful market development. It is noted that the predominant portion of those questioned are prepared to co-operate in further work on the subject

A comparison of LOCA analysis using SMOKIN and CERBERUS codes

Energy Technology Data Exchange (ETDEWEB)

Younis, M H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Gaboury, G [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

This paper presents the results of a comparison of the analyses of a postulated Loss of Coolant Accident (LOCA) in Pickering NGS A reactors using the two neutron kinetics codes SMOKIN and CERBERUS. Both codes have been used to simulate the space-time neutronic behaviour of CANDU-PHWR reactors. The main objective of the present study is to evaluate the accuracy with which SMOKIN can predict power transients compared to CERBERUS. The comparison shows that the two codes produce similar bulk power and reactivity transients. However, SMOKIN was found to overestimate the power transient (relative to CERBERUS) in some regions of the core, which is indicative of the spatial differences between the two codes. It was demonstrated that part of this overestimate is due to the use of reaction-rate averaged fuel properties in SMOKIN, compared to instantaneous fuel properties in CERBERUS. (author). 5 refs., 3 tabs., 6 figs.

Code of Conduct for wind-power projects - Feasibility study; Code of Conduct fuer windkraftprojekte. Machbarkeitsstudie - Schlussbericht

Energy Technology Data Exchange (ETDEWEB)

Strub, P. [Pierre Strub, freischaffender Berater, Binningen (Switzerland); Ziegler, Ch. [Inter Act, Basel (Switzerland)

2009-02-15

This final report deals with the results of a feasibility study concerning the development of a Code of Conduct for wind-power projects. The aim is to strengthen the acceptance of wind-power by the general public. The necessity of new, voluntary market instruments is discussed. The urgency of development in this area is quoted as being high, and the authors consider the feasibility of the definition of a code of conduct as being proven. The code of conduct can, according to the authors, be of use at various levels but primarily in project development. Further free-enterprise instruments are also suggested that should help support socially compatible and successful market development. It is noted that the predominant portion of those questioned are prepared to co-operate in further work on the subject

Data evaluation and code comparison activities

International Nuclear Information System (INIS)

Itikawa, Yukikazu; Takagi, Hidekazu; Nakamura, Yoshiharu; Imai, Makoto; Sasaki, Akira

2013-01-01

In atomic and molecular data base, intolerable numerical differences beyond error margin are found among some papers resulted from measurements or calculations even for the same collision processes. These differences spoil the reliability of the data base. This report describes the data evaluation for atomic and molecular data promoted by IAEA cooperated with other institutes, which Japanese researchers collaborate with. The reaction cross sections calculated numerically are evaluated for the collisions between electrons and molecular ions of H 2 + and HeH + . The application of an electron swarm parameter was shown for the evaluation and determination of the collision cross sections between electrons and molecules. In order to complete higher precision of atomic codes and a collisional-radiative model, IAEA held the workshop for the code comparison of the nonlocal thermodynamic equilibrium. (Y. Kazumata)

BER performance comparison of optical CDMA systems with/without turbo codes

Science.gov (United States)

Kulkarni, Muralidhar; Chauhan, Vijender S.; Dutta, Yashpal; Sinha, Ravindra K.

2002-08-01

In this paper, we have analyzed and simulated the BER performance of a turbo coded optical code-division multiple-access (TC-OCDMA) system. A performance comparison has been made between uncoded OCDMA and TC-OCDMA systems employing various OCDMA address codes (optical orthogonal codes (OOCs), Generalized Multiwavelength Prime codes (GMWPC's), and Generalized Multiwavelength Reed Solomon code (GMWRSC's)). The BER performance of TC-OCDMA systems has been analyzed and simulated by varying the code weight of address code employed by the system. From the simulation results, it is observed that lower weight address codes can be employed for TC-OCDMA systems that can have the equivalent BER performance of uncoded systems employing higher weight address codes for a fixed number of active users.

German nuclear codes revised: comparison with approaches used in other countries

International Nuclear Information System (INIS)

Raetzke, C.; Micklinghoff, M.

2005-01-01

The article deals with the plan of the German Federal Ministry for the Environment (BMU) to revise the German set of nuclear codes, and draws a comparison with approaches pursued in other countries in formulating and implementing new requirements imposed upon existing plants. A striking feature of the BMU project is the intention to have the codes reflect the state of the art in an entirely abstract way irrespective of existing plants. This implies new requirements imposed on plant design, among other things. However, the state authorities, which establish the licensing conditions for individual plants in concrete terms, will not be able to apply these new codes for legal reasons (protection of vested rights) to the extent in which they incorporate changes in safety philosophy. Also the procedure adopted has raised considerable concern. The processing time of two years is inordinately short, and participation of the public and of industry does not go beyond the strictly formal framework of general public participation. In the light of this absence of quality assurance, it would be surprising if this new set of codes did not suffer from considerable deficits in its contents. Other countries show that the BMU is embarking on an isolated approach in every respect. Elsewhere, backfitting requirements are developed carefully and over long periods of time; they are discussed in detail with the operators; costs and benefits are weighted, and the consequences are evaluated. These elements are in common to procedures in all countries, irrespective of very different steps in detail. (orig.)

Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyeong-Yeon, E-mail: hylee@kaeri.re.kr

2016-11-15

Highlights: • Comparison of elevated temperature design (ETD) codes was made. • Material properties and evaluation procedures were compared. • Two heat-resistant materials of Grade 91 steel and austenitic stainless steel 316 are the target materials in the present study. • Application of the ETD codes to Generation IV reactor components and a comparison of the conservatism was conducted. - Abstract: The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.

A directory of computer codes suitable for stress analysis of HLW containers - Compas project

International Nuclear Information System (INIS)

1989-01-01

This document reports the work carried out for the Compas project which looked at the capabilities of various computer codes for the stress analysis of high-level nuclear-waste containers and overpacks. The report concentrates on codes used by the project partners, but also includes a number of the major commercial finite element codes. The report falls into two parts. The first part of the report describes the capabilities of the codes. This includes details of the solution methods used in the codes, the types of analysis which they can carry out and the interfacing with pre - and post - processing packages. This is the more comprehensive section of the report. The second part of the report looks at the performance of a selection of the codes (those used by the project partners). This look at how the codes perform in a number of test problems which require calculations typical of those encountered in the design and analysis of high-level waste containers and overpacks

Simulation of single-phase rod bundle flow. Comparison between CFD-code ESTET, PWR core code THYC and experimental results

International Nuclear Information System (INIS)

Mur, J.; Larrauri, D.

1998-07-01

Computer simulation of flow in configurations close to pressurized water reactor (PWR) geometry is of great interest for Electricite de France (EDF). Although simulation of the flow through a whole PWR core with an all purpose CFD-code is not yet achievable, such a tool cna be quite useful to perform numerical experiments in order to try and improve the modeling introduced in computer codes devoted to reactor core thermal-hydraulic analysis. Further to simulation in small bare rod bundle configurations, the present study is focused on the simulation, with CFD-code ESTET and PWR core code THYC, of the flow in the experimental configuration VATICAN-1. ESTET simulation results are compared on the one hand to local velocity and concentration measurements, on the other hand with subchannel averaged values calculated by THYC. As far as the comparison with measurements is concerned, ESTET results are quite satisfactory relatively to available experimental data and their uncertainties. The effect of spacer grids and the prediction of the evolution of an unbalanced velocity profile seem to be correctly treated. As far as the comparison with THYC subchannel averaged values is concerned, the difficulty of a direct comparison between subchannel averaged and local values is pointed out. ESTET calculated local values are close to experimental local values. ESTET subchannel averaged values are also close to THYC calculation results. Thus, THYC results are satisfactory whereas their direct comparison to local measurements could show some disagreement. (author)

Comparison of ANL containment codes with SNR-300 simulation experiments

International Nuclear Information System (INIS)

Marchertas, A.H.; Wang, C.Y.; Fistedis, S.H.

1976-01-01

A comparison of REXCO and ICECO code predictions is made with data obtained from experiments of LMFBR excursion models. The comparisons are based on published results of tests conducted for the safety analysis of the SNR-300 fast breeder. The test configurations consist of a centrally located spherical source immersed in a pool of water which is encased in a cylindrical container. The cylinical walls of the container are prestressed by holddown bolts which span the two rigid ends. The space above the surface of the water within the container is occupied by air. Although certain aspects of the tests could not be simulated by the analytical models exactly, the comparison of results shows quite close agreement. The fact that the REXCO and ICECO codes involve different analytical formulations, their own close correspondence of results lends added credence to the value of analytical predictions

Code comparison for accelerator design and analysis

International Nuclear Information System (INIS)

Parsa, Z.

1988-01-01

We present a comparison between results obtained from standard accelerator physics codes used for the design and analysis of synchrotrons and storage rings, with programs SYNCH, MAD, HARMON, PATRICIA, PATPET, BETA, DIMAD, MARYLIE and RACE-TRACK. In our analysis we have considered 5 (various size) lattices with large and small angles including AGS Booster (10/degree/ bend), RHIC (2.24/degree/), SXLS, XLS (XUV ring with 45/degree/ bend) and X-RAY rings. The differences in the integration methods used and the treatment of the fringe fields in these codes could lead to different results. The inclusion of nonlinear (e.g., dipole) terms may be necessary in these calculations specially for a small ring. 12 refs., 6 figs., 10 tabs

Code of Conduct for wind-power projects - Phase 3

International Nuclear Information System (INIS)

Strub, P.; Ziegler, Ch.

2008-11-01

This paper discusses the results of phase three of a project concerning wind-power projects. Feasibility and strategy aspects are examined and discussed. The current state of the wind power market is discussed on the basis of the results of a survey made on the subject. The social acceptance of wind power installations is discussed, whereby the rejection of particular projects is compared with a general lack of acceptance. Requirements placed on such projects and possible solutions are discussed. Finally, the feasibility of setting up a code of conduct in the area of wind-power projects is discussed and the definition of further instruments is examined

Decoding using back-project algorithm from coded image in ICF

International Nuclear Information System (INIS)

Jiang shaoen; Liu Zhongli; Zheng Zhijian; Tang Daoyuan

1999-01-01

The principle of the coded imaging and its decoding in inertial confinement fusion is described simply. The authors take ring aperture microscope for example and use back-project (BP) algorithm to decode the coded image. The decoding program has been performed for numerical simulation. Simulations of two models are made, and the results show that the accuracy of BP algorithm is high and effect of reconstruction is good. Thus, it indicates that BP algorithm is applicable to decoding for coded image in ICF experiments

Towards a heavy-ion transport capability in the MARS15 Code

International Nuclear Information System (INIS)

Mokhov, N.V.; Gudima, K.K.; Mashnik, S.G.; Rakhno, I.L.; Striganov, S.

2004-01-01

In order to meet the challenges of new accelerator and space projects and further improve modelling of radiation effects in microscopic objects, heavy-ion interaction and transport physics have been recently incorporated into the MARS15 Monte Carlo code. A brief description of new modules is given in comparison with experimental data. The MARS Monte Carlo code is widely used in numerous accelerator, detector, shielding and cosmic ray applications. The needs of the Relativistic Heavy-Ion Collider, Large Hadron Collider, Rare Isotope Accelerator and NASA projects have recently induced adding heavy-ion interaction and transport physics to the MARS15 code. The key modules of the new implementation are described below along with their comparisons to experimental data.

Severe accident analysis code Sampson for impact project

International Nuclear Information System (INIS)

Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

2001-01-01

Four years of the IMPACT project Phase 1 (1994-1997) had been completed with financial sponsorship from the Japanese government's Ministry of Economy, Trade and Industry. At the end of the phase, demonstration simulations by combinations of up to 11 analysis modules developed for severe accident analysis in the SAMPSON Code were performed and physical models in the code were verified. The SAMPSON prototype was validated by TMI-2 and Phebus-FP test analyses. Many of empirical correlation and conventional models have been replaced by mechanistic models during Phase 2 (1998-2000). New models for Accident Management evaluation have been also developed. (author)

A CFD code comparison of wind turbine wakes

International Nuclear Information System (INIS)

Van der Laan, M P; Sørensen, N N; Storey, R C; Cater, J E; Norris, S E

2014-01-01

A comparison is made between the EllipSys3D and SnS CFD codes. Both codes are used to perform Large-Eddy Simulations (LES) of single wind turbine wakes, using the actuator disk method. The comparison shows that both LES models predict similar velocity deficits and stream-wise Reynolds-stresses for four test cases. A grid resolution study, performed in EllipSys3D and SnS, shows that a minimal uniform cell spacing of 1/30 of the rotor diameter is necessary to resolve the wind turbine wake. In addition, the LES-predicted velocity deficits are also compared with Reynolds-Averaged Navier Stokes simulations using EllipSys3D for a test case that is based on field measurements. In these simulations, two eddy viscosity turbulence models are employed: the k-ε model and the k-ε-f p model. Where the k-ε model fails to predict the velocity deficit, the results of the k-ε-f P model show good agreement with both LES models and measurements

The Second Workshop on Lineshape Code Comparison: Isolated Lines

Directory of Open Access Journals (Sweden)

Spiros Alexiou

2014-05-01

Full Text Available In this work, we briefly summarize the theoretical aspects of isolated line broadening. We present and discuss test run comparisons from different participating lineshape codes for the 2s-2p transition for LiI, B III and NV.

The APS SASE FEL: modeling and code comparison

International Nuclear Information System (INIS)

Biedron, S. G.

1999-01-01

A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL

Comparison of DT neutron production codes MCUNED, ENEA-JSI source subroutine and DDT

Energy Technology Data Exchange (ETDEWEB)

Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Kodeli, Ivan [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Milocco, Alberto [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Sauvan, Patrick [Departamento de Ingeniería Energética, E.T.S. Ingenieros Industriales, UNED, C/Juan del Rosal 12, 28040 Madrid (Spain); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

2016-11-01

Highlights: • Results of three codes capable of simulating the accelerator based DT neutron generators were compared on a simple model where only a thin target made of mixture of titanium and tritium is present. Two typical deuteron beam energies, 100 keV and 250 keV, were used in the comparison. • Comparisons of the angular dependence of the total neutron flux and spectrum as well as the neutron spectrum of all the neutrons emitted from the target show general agreement of the results but also some noticeable differences. • A comparison of figures of merit of the calculations using different codes showed that the computational time necessary to achieve the same statistical uncertainty can vary for more than 30× when different codes for the simulation of the DT neutron generator are used. - Abstract: As the DT fusion reaction produces neutrons with energies significantly higher than in fission reactors, special fusion-relevant benchmark experiments are often performed using DT neutron generators. However, commonly used Monte Carlo particle transport codes such as MCNP or TRIPOLI cannot be directly used to analyze these experiments since they do not have the capabilities to model the production of DT neutrons. Three of the available approaches to model the DT neutron generator source are the MCUNED code, the ENEA-JSI DT source subroutine and the DDT code. The MCUNED code is an extension of the well-established and validated MCNPX Monte Carlo code. The ENEA-JSI source subroutine was originally prepared for the modelling of the FNG experiments using different versions of the MCNP code (−4, −5, −X) and was later extended to allow the modelling of both DT and DD neutron sources. The DDT code prepares the DT source definition file (SDEF card in MCNP) which can then be used in different versions of the MCNP code. In the paper the methods for the simulation of the DT neutron production used in the codes are briefly described and compared for the case of a

Parallel Scaling Characteristics of Selected NERSC User ProjectCodes

Energy Technology Data Exchange (ETDEWEB)

Skinner, David; Verdier, Francesca; Anand, Harsh; Carter,Jonathan; Durst, Mark; Gerber, Richard

2005-03-05

This report documents parallel scaling characteristics of NERSC user project codes between Fiscal Year 2003 and the first half of Fiscal Year 2004 (Oct 2002-March 2004). The codes analyzed cover 60% of all the CPU hours delivered during that time frame on seaborg, a 6080 CPU IBM SP and the largest parallel computer at NERSC. The scale in terms of concurrency and problem size of the workload is analyzed. Drawing on batch queue logs, performance data and feedback from researchers we detail the motivations, benefits, and challenges of implementing highly parallel scientific codes on current NERSC High Performance Computing systems. An evaluation and outlook of the NERSC workload for Allocation Year 2005 is presented.

Wind-induced transmission tower foundation loads. A field study-design code comparison

Energy Technology Data Exchange (ETDEWEB)

Savory, E. [Department of Mechanical and Materials Engineering, University of Western Ontario, London, Ont. (Canada); Parke, G.A.R.; Disney, P.; Toy, N. [School of Engineering, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom)

2008-06-15

This paper presents a comparison between the wind-induced foundation loads measured on a type L6 transmission line tower during a field study in the UK and those computed using the UK Code of Practice for lattice tower and transmission line design (BS8100). In this work, the Code provisions have been generalised to give the wind-induced strain in each of the tower legs immediately above the foundation as a function of wind direction and wind speed at the top of the tower. The complete data set from the field monitoring has been decomposed to provide a similar formulation for comparison purposes. The analysis shows excellent agreement between the Code calculations and the measured results, within the overall accuracy of the field data. This indicates that, at least for the tower type examined here, the existing design Code provides a reliable transformation of the local wind speed at the top of the tower into tension and compression loads on the foundations. (author)

Pool swell sub-scale testing and code comparison

International Nuclear Information System (INIS)

Elisson, K.

1981-01-01

The main objective of the experiment was to investigate the pool swell dynamics in general and the forces on the lowered central part of the diaphragm between drywell and wetwell in particular. Apart from the high speed camera pressure transducers and strain gauges were used to monitor the transient. Data was recorded on a 14 channel FM recorder and then digitalised and plotted. In total more than one hundred tests were performed including parametric variations of for example geometry, break flow, initial drywell pressure and initial water level. In parallel to this experiment pool swell calculations have been performed with the computer codes COPTA and STEALTH. COPTA which is a lumped mass code for pressure suppression containment analysis has a slug pool swell mode. STEALTH which is a general purpose lagrangian hydrodynamics code has been used in a 2-D axisymmetric version. The STEALTH code has been used to calculate the radial variations in the vertical displacement and velocity of the pool surface and to predict the load on the lowered central part of the diaphragm. A comparison between the calculations and the experimental data indicates that both codes are sufficiently correct in their description of the pool swell transient. (orig.)

Comparison of design margin for core shroud in between design and construction code and fitness-for-service code

International Nuclear Information System (INIS)

Dozaki, Koji

2007-01-01

Structural design methods for core shroud of BWR are specified in JSME Design and Construction Code, like ASME Boiler and Pressure Vessel Code Sec. III, as a part of core support structure. Design margins are defined according to combination of the structural design method selected and service limit considered. Basically, those margins in JSME Code were determined after ASME Sec. III. Designers can select so-called twice-slope method for core shroud design among those design methods. On the other hand, flaw evaluation rules have been established for core shroud in JSME Fitness-for-Service Code. Twice-slope method is also adopted for fracture evaluation in that code even when the core shroud contains a flaw. Design margin was determined as structural factors separately from Design and Construction Code. As a natural consequence, there is a difference in those design margins between the two codes. In this paper, it is shown that the design margin in Fitness-for-Service Code is conservative by experimental evidences. Comparison of design margins between the two codes is discussed. (author)

Enhancing co-operation between AVN, IRSN and GRS: the junior staff pilot project on the comparative testing of IPA codes

International Nuclear Information System (INIS)

Hoyos, A. de; Keesmann, S.; Smidts, O.

2006-01-01

- Objectives: The project takes place within the framework of the Junior Staff Program of AVN, GRS and IRSN which aims at creating a junior staff network among European TSOs. The objective of this project is to apply integrated performance assessment (IPA) tools used by AVN, IRSN and GRS to two generic and simplified models (Bure site in France and Mol site in Belgium) for disposal systems in argillaceous formations. The comparison of the results from different codes applied to the disposal systems of the two mentioned sites aims at a better understanding of the confinement capabilities of the considered geological formations and of the IPA methodology in general. The incentive is a common understanding of approaches developed by each partner and the improvement of this expertise. More specifically, this pilot project aims at enhancing exchanges of views and mutual experiences in the field of understanding major safety functions. - Tools and Methods: A new code for the assessment of barrier systems in argillaceous formations has only recently been developed at GRS, as in the past such formations played a minor role as a possible hosting environment for a repository in Germany. The project also serves as a test case for this code. The considered disposal systems are defined on the basis of the concepts and data available for Mol and Bure. The program packages used for the performance assessment calculations are: HYDRUS-1D with source term module (AVN), GoldSim (IRSN) and EMOS-modules CLAYPOS and CHET (GRS). While the coupling of HYDRUS-1D with a source term module and the EMOS-modules are FORTRAN77- coded programs specifically developed for the simulation of parts of a barrier system of a final repository, GoldSim is a general purpose simulation environment with an integrated graphical user interface for modelling and data output. Models realized in GoldSim are flexible and can be easily adapted to new requirements. The software also offers an intrinsic

CFD code comparison for 2D airfoil flows

DEFF Research Database (Denmark)

Sørensen, Niels N.; Méndez, B.; Muñoz, A.

2016-01-01

The current paper presents the effort, in the EU AVATAR project, to establish the necessary requirements to obtain consistent lift over drag ratios among seven CFD codes. The flow around a 2D airfoil case is studied, for both transitional and fully turbulent conditions at Reynolds numbers of 3...

EG and G and NASA face seal codes comparison

Science.gov (United States)

Basu, Prit

1994-01-01

This viewgraph presentation presents the following results for the example comparison: EG&G code with face deformations suppressed and SPIRALG agree well with each other as well as with the experimental data; 0 rpm stiffness data calculated by EG&G code are about 70-100 percent lower than that by SPIRALG; there is no appreciable difference between 0 rpm and 16,000 rpm stiffness and damping coefficients calculated by SPIRALG; and the film damping above 500 psig calculated by SPIRALG is much higher than the O-Ring secondary seal damping (e.g. 50 lbf.s/in).

Comparisons of 'Identical' Simulations by the Eulerian Gyrokinetic Codes GS2 and GYRO

Science.gov (United States)

Bravenec, R. V.; Ross, D. W.; Candy, J.; Dorland, W.; McKee, G. R.

2003-10-01

A major goal of the fusion program is to be able to predict tokamak transport from first-principles theory. To this end, the Eulerian gyrokinetic code GS2 was developed years ago and continues to be improved [1]. Recently, the Eulerian code GYRO was developed [2]. These codes are not subject to the statistical noise inherent to particle-in-cell (PIC) codes, and have been very successful in treating electromagnetic fluctuations. GS2 is fully spectral in the radial coordinate while GYRO uses finite-differences and ``banded" spectral schemes. To gain confidence in nonlinear simulations of experiment with these codes, ``apples-to-apples" comparisons (identical profile inputs, flux-tube geometry, two species, etc.) are first performed. We report on a series of linear and nonlinear comparisons (with overall agreement) including kinetic electrons, collisions, and shaped flux surfaces. We also compare nonlinear simulations of a DIII-D discharge to measurements of not only the fluxes but also the turbulence parameters. [1] F. Jenko, et al., Phys. Plasmas 7, 1904 (2000) and refs. therein. [2] J. Candy, J. Comput. Phys. 186, 545 (2003).

Comparison of Two Commercial FE-Codes for Sheet Metal Forming

International Nuclear Information System (INIS)

Revuelta, A.; Larkiola, J.; Kanervo, K.; Korhonen, A. S.; Myllykoski, P.

2007-01-01

There is urgent need to develop new advanced fast and cost-effective mass-production methods for small sheet metal components. Traditionally progressive dies have been designed by using various CAD techniques. Recent results in mass production of small sheet metal parts using progressive dies and a transfer press showed that the tool design time may be cut in up to a half by using 3D finite element simulation of forming. In numerical simulation of sheet metal forming better constitutive models are required to obtain more accurate results, reduce the time for tool design and cut the production costs further. Accurate models are needed to describe the initial yielding, subsequent work hardening and to predict the formability. In this work two commercially available finite element simulation codes, PAM-STAMP and LS-DYNA, were compared in forming of small austenitic stainless steel sheet part for electronic industry. Several constitutive models were used in both codes and the results were compared. Comparisons were made between the same models in each of the codes and also between different models in the same code. Material models ranged from very simple to advanced ones, which took into account anisotropy and both isotropic and kinematic hardening behavior. In order to make a valid comparison we employed similar finite element meshes. The effects of the material models parameters were studied and the results were compared with experiments. The effects of the computational time were also studied

Coded aperture optimization using Monte Carlo simulations

International Nuclear Information System (INIS)

Martineau, A.; Rocchisani, J.M.; Moretti, J.L.

2010-01-01

Coded apertures using Uniformly Redundant Arrays (URA) have been unsuccessfully evaluated for two-dimensional and three-dimensional imaging in Nuclear Medicine. The images reconstructed from coded projections contain artifacts and suffer from poor spatial resolution in the longitudinal direction. We introduce a Maximum-Likelihood Expectation-Maximization (MLEM) algorithm for three-dimensional coded aperture imaging which uses a projection matrix calculated by Monte Carlo simulations. The aim of the algorithm is to reduce artifacts and improve the three-dimensional spatial resolution in the reconstructed images. Firstly, we present the validation of GATE (Geant4 Application for Emission Tomography) for Monte Carlo simulations of a coded mask installed on a clinical gamma camera. The coded mask modelling was validated by comparison between experimental and simulated data in terms of energy spectra, sensitivity and spatial resolution. In the second part of the study, we use the validated model to calculate the projection matrix with Monte Carlo simulations. A three-dimensional thyroid phantom study was performed to compare the performance of the three-dimensional MLEM reconstruction with conventional correlation method. The results indicate that the artifacts are reduced and three-dimensional spatial resolution is improved with the Monte Carlo-based MLEM reconstruction.

Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

Energy Technology Data Exchange (ETDEWEB)

Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Ha, Sang Jun; Choi, Hoon [KHNP-CENTERAL RESEARCH INSTITUTE, Daejeon (Korea, Republic of)

2013-10-15

CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic

Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

International Nuclear Information System (INIS)

Choo, Yeon Jun; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul; Ha, Sang Jun; Choi, Hoon

2013-01-01

CAP, in the form that is linked with SPACE, computed the containment back-pressure during LOCA accident. In previous SAR (safety analysis report) report of Shin-Kori Units 3 and 4, the CONTEMPT series of codes(hereby referred to as just 'CONTEMPT') is used to evaluate the containment safety during the postulated loss-of-coolant accident (LOCA). In more detail, CONTEMPT-LT/028 was used to calculate the containment maximum PT, while CONTEMPT4/MOD5 to calculate the minimum PT. Actually, in minimum PT analysis, CONTEMPT4/MOD5, which provide back pressure condition of containment, was linked with RELAP5/MOD3.3 which calculate the amount of blowdown into containment. In this analysis, CONTEMPT4/MOD5 was modified based on KREM. CONTEMPT code was developed to predict the long term behavior of water-cooled nuclear reactor containment systems subjected to LOCA conditions. It calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP has the similar code features and it therefore is expected to show the similar analysis performance with CAP. In this study, the differences between CAP and two CONTEMPT code versions (CONTEMPT-LT/028 for maximum PT and CONTEMPT4/MOD5 for minimum PT) are, in detail, identified and the code performances were compared for the same problem. Code by code comparison was carried out to identify the difference of LOCA analysis between a series of COMTEMPT and CAP code. With regard to important factors that affect the transient behavior of compartment thermodynamic state in

Codification of LMFBR rules and comparison of codes

International Nuclear Information System (INIS)

Faure, O.; Debaene, J.P.

1993-01-01

The first part of this report presents the basic RCC-MR (regles de conception et de construction des materiels mecaniques des ilots nucleaires, reacteurs a neutrons rapides) design rules and their purpose. The second part is a qualitative comparison between RCC-MR, Code case N47 (ASME) and ETSDG Guide (MONJU Guide), made on the following topics: negligible creep test, ratcheting, creep fatigue, buckling, piping rules. An outline is given on improvements to RCC-MR rules now in progress

The FRAM code: Description and some comparisons with MGA

International Nuclear Information System (INIS)

Sampson, T.E.; Kelley, T.A.

1994-01-01

The authors describe the initial development of the FRAM gamma-ray spectrometry code for analyzing plutonium isotopics, discuss its methodology, and present some comparisons with MGA on identical items. They also present some of the features of a new Windows 3.1-based version (PC/FRAM) and describe some current measurement problems. Development of the FRAM code began in about 1985, growing out of the need at the Los Alamos TA-55 Plutonium Facility for an isotopic analysis code to give accurate results for the effective specific power of heterogeneous (Am/Pu) pyrochemical residues. These residues present a difficult challenge because the americium is present mostly in a low-Z salt matrix (AmCl 3 ) with fines and small pieces of plutonium metal dispersed throughout the salt. Plutonium gamma rays suffer different attenuation than americium gamma rays of the same energy; this makes conventional analysis with a single relative efficiency function inaccurate for Am/Pu ratios and affects the analysis in other subtle ways

THE AGORA HIGH-RESOLUTION GALAXY SIMULATIONS COMPARISON PROJECT

International Nuclear Information System (INIS)

Kim, Ji-hoon; Conroy, Charlie; Goldbaum, Nathan J.; Krumholz, Mark R.; Abel, Tom; Agertz, Oscar; Gnedin, Nickolay Y.; Kravtsov, Andrey V.; Bryan, Greg L.; Ceverino, Daniel; Christensen, Charlotte; Hummels, Cameron B.; Dekel, Avishai; Guedes, Javiera; Hahn, Oliver; Hobbs, Alexander; Hopkins, Philip F.; Iannuzzi, Francesca; Keres, Dusan; Klypin, Anatoly

2014-01-01

We introduce the Assembling Galaxies Of Resolved Anatomy (AGORA) project, a comprehensive numerical study of well-resolved galaxies within the ΛCDM cosmology. Cosmological hydrodynamic simulations with force resolutions of ∼100 proper pc or better will be run with a variety of code platforms to follow the hierarchical growth, star formation history, morphological transformation, and the cycle of baryons in and out of eight galaxies with halo masses M vir ≅ 10 10 , 10 11 , 10 12 , and 10 13 M ☉ at z = 0 and two different ('violent' and 'quiescent') assembly histories. The numerical techniques and implementations used in this project include the smoothed particle hydrodynamics codes GADGET and GASOLINE, and the adaptive mesh refinement codes ART, ENZO, and RAMSES. The codes share common initial conditions and common astrophysics packages including UV background, metal-dependent radiative cooling, metal and energy yields of supernovae, and stellar initial mass function. These are described in detail in the present paper. Subgrid star formation and feedback prescriptions will be tuned to provide a realistic interstellar and circumgalactic medium using a non-cosmological disk galaxy simulation. Cosmological runs will be systematically compared with each other using a common analysis toolkit and validated against observations to verify that the solutions are robust—i.e., that the astrophysical assumptions are responsible for any success, rather than artifacts of particular implementations. The goals of the AGORA project are, broadly speaking, to raise the realism and predictive power of galaxy simulations and the understanding of the feedback processes that regulate galaxy 'metabolism'. The initial conditions for the AGORA galaxies as well as simulation outputs at various epochs will be made publicly available to the community. The proof-of-concept dark-matter-only test of the formation of a galactic halo with a z = 0 mass of M

A comparison of oxide thickness predictability from the perspective of codes

Energy Technology Data Exchange (ETDEWEB)

Park, Joo-Young; Shin, Hye-In; Kim, Kyung-Tae; Han, Hee-Tak; Kim, Hong-Jin; Kim, Yong-Hwan [KEPCO Nuclear Fuel Co. Ltd., Daejeon (Korea, Republic of)

2016-10-15

In Korea, OPR1000 and Westinghouse type nuclear power plant reactor fuel rods oxide thickness has been evaluated by imported code A. Because of this, there have been multiple constraints in operation and maintenance of fuel rod design system. For this reason, there has been a growing demand to establish an independent fuel rod design system. To meet this goal, KNF has recently developed its own code B for fuel rod design. The objective of this study is to compare oxide thickness prediction performance between code A and code B and to check the validity of predicting corrosion behaviors of newly developed code B. This study is based on Pool Side Examination (PSE) data for the performance confirmation. For the examination procedures, the oxide thickness measurement methods and equipment of PSE are described in detail. In this study, code B is confirmed conservatism and validity on evaluating cladding oxide thickness through the comparison with code A. Code prediction values show higher value than measured data from PSE. Throughout this study, the values by code B are evaluated and proved to be valid in a view point of the oxide thickness evaluation. However, the code B input for prediction has been made by designer's judgment with complex handwork that might be lead to excessive conservative result and ineffective design process with some possibility of errors.

Comparisons of coded aperture imaging using various apertures and decoding methods

International Nuclear Information System (INIS)

Chang, L.T.; Macdonald, B.; Perez-Mendez, V.

1976-07-01

The utility of coded aperture γ camera imaging of radioisotope distributions in Nuclear Medicine is in its ability to give depth information about a three dimensional source. We have calculated imaging with Fresnel zone plate and multiple pinhole apertures to produce coded shadows and reconstruction of these shadows using correlation, Fresnel diffraction, and Fourier transform deconvolution. Comparisons of the coded apertures and decoding methods are made by evaluating their point response functions both for in-focus and out-of-focus image planes. Background averages and standard deviations were calculated. In some cases, background subtraction was made using combinations of two complementary apertures. Results using deconvolution reconstruction for finite numbers of events are also given

Inter-comparison of Computer Codes for TRISO-based Fuel Micro-Modeling and Performance Assessment

International Nuclear Information System (INIS)

Boer, Brian; Keun Jo, Chang; Wu, Wen; Ougouag, Abderrafi M.; McEachren, Donald; Venneri, Francesco

2010-01-01

The Next Generation Nuclear Plant (NGNP), the Deep Burn Pebble Bed Reactor (DB-PBR) and the Deep Burn Prismatic Block Reactor (DB-PMR) are all based on fuels that use TRISO particles as their fundamental constituent. The TRISO particle properties include very high durability in radiation environments, hence the designs reliance on the TRISO to form the principal barrier to radioactive materials release. This durability forms the basis for the selection of this fuel type for applications such as Deep Bun (DB), which require exposures up to four times those expected for light water reactors. It follows that the study and prediction of the durability of TRISO particles must be carried as part of the safety and overall performance characterization of all the designs mentioned above. Such evaluations have been carried out independently by the performers of the DB project using independently developed codes. These codes, PASTA, PISA and COPA, incorporate models for stress analysis on the various layers of the TRISO particle (and of the intervening matrix material for some of them), model for fission products release and migration then accumulation within the SiC layer of the TRISO particle, just next to the layer, models for free oxygen and CO formation and migration to the same location, models for temperature field modeling within the various layers of the TRISO particle and models for the prediction of failure rates. All these models may be either internal to the code or external. This large number of models and the possibility of different constitutive data and model formulations and the possibility of a variety of solution techniques makes it highly unlikely that the model would give identical results in the modeling of identical situations. The purpose of this paper is to present the results of an inter-comparison between the codes and to identify areas of agreement and areas that need reconciliation. The inter-comparison has been carried out by the cooperating

Comparison of accelerator codes for a RHIC [Relativistic Heavy Ion Collider] lattice

International Nuclear Information System (INIS)

Milutinovic, J.; Ruggiero, A.G.

1989-01-01

We present the results of comparison of performances of several tracking or/and analysis codes. The basic purpose of this program was to assess reliability and accuracy of these codes, i.e., to determine the so-called ''error bars'' for the predicted values of tunes and other lattice functions as a minimum and, if possible, to discover potential difficulties with underlying physical models in these codes, inadequate algorithms, residual bugs and the like. Not only have we been able to determine the error bars, which for instance for the tunes at dp/p = +1% are Δν/sub ξ/ = 0.0027, Δν/sub y/ = 0.0010, but also our program has brought about improvements of several codes. 8 refs., 3 figs., 2 tabs

Overview of ACTYS project on development of indigenous state-of-the-art code suites for nuclear activation analysis

International Nuclear Information System (INIS)

Subhash, P.V.; Tadepalli, Sai Chaitanya; Deshpande, Shishir P.; Kanth, Priti; Srinivasan, R.

2017-01-01

Rigorous activation calculations are warranted for safer and efficient design of future fusion machines. Suitable activation codes, which yield accurate results with faster performance yet include all fusion relevant reactions are a prerequisite. To meet these, an indigenous project called ACTYS-Project is initiated and as a result, four state-of-art codes are developed so far. The goal of this project is to develop indigenous state-of-the-art code suites for nuclear activation analysis

Computerized Dental Comparison: A Critical Review of Dental Coding and Ranking Algorithms Used in Victim Identification.

Science.gov (United States)

Adams, Bradley J; Aschheim, Kenneth W

2016-01-01

Comparison of antemortem and postmortem dental records is a leading method of victim identification, especially for incidents involving a large number of decedents. This process may be expedited with computer software that provides a ranked list of best possible matches. This study provides a comparison of the most commonly used conventional coding and sorting algorithms used in the United States (WinID3) with a simplified coding format that utilizes an optimized sorting algorithm. The simplified system consists of seven basic codes and utilizes an optimized algorithm based largely on the percentage of matches. To perform this research, a large reference database of approximately 50,000 antemortem and postmortem records was created. For most disaster scenarios, the proposed simplified codes, paired with the optimized algorithm, performed better than WinID3 which uses more complex codes. The detailed coding system does show better performance with extremely large numbers of records and/or significant body fragmentation. © 2015 American Academy of Forensic Sciences.

Code of Conduct for wind-power projects - Phase 3; Code of Conduct fuer windkraftprojekte. Phase 3 Machbarkeit und Strategie

Energy Technology Data Exchange (ETDEWEB)

Strub, P. [Pierre Strub, freischaffender Berater, Binningen (Switzerland); Ziegler, Ch. [Inter Act, Basel (Switzerland)

2008-11-15

This paper discusses the results of phase three of a project concerning wind-power projects. Feasibility and strategy aspects are examined and discussed. The current state of the wind power market is discussed on the basis of the results of a survey made on the subject. The social acceptance of wind power installations is discussed, whereby the rejection of particular projects is compared with a general lack of acceptance. Requirements placed on such projects and possible solutions are discussed. Finally, the feasibility of setting up a code of conduct in the area of wind-power projects is discussed and the definition of further instruments is examined

Automated Testing Infrastructure and Result Comparison for Geodynamics Codes

Science.gov (United States)

Heien, E. M.; Kellogg, L. H.

2013-12-01

The geodynamics community uses a wide variety of codes on a wide variety of both software and hardware platforms to simulate geophysical phenomenon. These codes are generally variants of finite difference or finite element calculations involving Stokes flow or wave propagation. A significant problem is that codes of even low complexity will return different results depending on the platform due to slight differences in hardware, software, compiler, and libraries. Furthermore, changes to the codes during development may affect solutions in unexpected ways such that previously validated results are altered. The Computational Infrastructure for Geodynamics (CIG) is funded by the NSF to enhance the capabilities of the geodynamics community through software development. CIG has recently done extensive work in setting up an automated testing and result validation system based on the BaTLab system developed at the University of Wisconsin, Madison. This system uses 16 variants of Linux and Mac platforms on both 32 and 64-bit processors to test several CIG codes, and has also recently been extended to support testing on the XSEDE TACC (Texas Advanced Computing Center) Stampede cluster. In this work we overview the system design and demonstrate how automated testing and validation occurs and results are reported. We also examine several results from the system from different codes and discuss how changes in compilers and libraries affect the results. Finally we detail some result comparison tools for different types of output (scalar fields, velocity fields, seismogram data), and discuss within what margins different results can be considered equivalent.

Multi-dimensional free-electron laser simulation codes: a comparison study

CERN Document Server

Biedron, S G; Dejus, Roger J; Faatz, B; Freund, H P; Milton, S V; Nuhn, H D; Reiche, S

2000-01-01

A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL.

Multi-dimensional free-electron laser simulation codes: a comparison study

International Nuclear Information System (INIS)

Biedron, S. G.; Chae, Y. C.; Dejus, R. J.; Faatz, B.; Freund, H. P.; Milton, S. V.; Nuhn, H.-D.; Reiche, S.

1999-01-01

A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL

Comparison of community managed projects and conventional ...

African Journals Online (AJOL)

Comparison of community managed projects and conventional approaches in rural water supply of Ethiopia. ... African Journal of Environmental Science and Technology ... This study aimed to compare Community Managed Projects (CMP) approach with the conventional approaches (Non-CMP) in the case of Ethiopia.

OC3—Benchmark Exercise of Aero-elastic Offshore Wind Turbine Codes

Science.gov (United States)

Passon, P.; Kühn, M.; Butterfield, S.; Jonkman, J.; Camp, T.; Larsen, T. J.

2007-07-01

This paper introduces the work content and status of the first international investigation and verification of aero-elastic codes for offshore wind turbines as performed by the "Offshore Code Comparison Collaboration"(OC3) within the "IEA Wind Annex XXIII - Subtask 2". An overview is given on the state-of-the-art of the concerned offshore wind turbine simulation codes. Exemplary results of benchmark simulations from the first phase of the project are presented and discussed while subsequent phases are introduced. Furthermore, the paper discusses areas where differences between the codes have been identified and the sources of those differences, such as the differing theories implemented into the individual codes. Finally, further research and code development needs are presented based on the latest findings from the current state of the project.

COSA II Further benchmark exercises to compare geomechanical computer codes for salt

International Nuclear Information System (INIS)

Lowe, M.J.S.; Knowles, N.C.

1989-01-01

Project COSA (COmputer COdes COmparison for SAlt) was a benchmarking exercise involving the numerical modelling of the geomechanical behaviour of heated rock salt. Its main objective was to assess the current European capability to predict the geomechanical behaviour of salt, in the context of the disposal of heat-producing radioactive waste in salt formations. Twelve organisations participated in the exercise in which their solutions to a number of benchmark problems were compared. The project was organised in two distinct phases: The first, from 1984-1986, concentrated on the verification of the computer codes. The second, from 1986-1988 progressed to validation, using three in-situ experiments at the Asse research facility in West Germany as a basis for comparison. This document reports the activities of the second phase of the project and presents the results, assessments and conclusions

A comparison of two three-dimensional shell-element transient electromagnetics codes

International Nuclear Information System (INIS)

Yugo, J.J.; Williamson, D.E.

1992-01-01

Electromagnetic forces due to eddy currents strongly influence the design of components for the next generation of fusion devices. An effort has been made to benchmark two computer programs used to generate transient electromagnetic loads: SPARK and EddyCuFF. Two simple transient field problems were analyzed, both of which had been previously analyzed by the SPARK code with results recorded in the literature. A third problem that uses an ITER inboard blanket benchmark model was analyzed as well. This problem was driven with a self-consistent, distributed multifilament plasma model generated by an axisymmetric physics code. The benchmark problems showed good agreement between the two shell-element codes. Variations in calculated eddy currents of 1--3% have been found for similar, finely meshed models. A difference of 8% was found in induced current and 20% in force for a coarse mesh and complex, multifilament field driver. Because comparisons were made to results obtained from literature, model preparation and code execution times were not evaluated

Comparison of secondary flows predicted by a viscous code and an inviscid code with experimental data for a turning duct

Science.gov (United States)

Schwab, J. R.; Povinelli, L. A.

1984-01-01

A comparison of the secondary flows computed by the viscous Kreskovsky-Briley-McDonald code and the inviscid Denton code with benchmark experimental data for turning duct is presented. The viscous code is a fully parabolized space-marching Navier-Stokes solver while the inviscid code is a time-marching Euler solver. The experimental data were collected by Taylor, Whitelaw, and Yianneskis with a laser Doppler velocimeter system in a 90 deg turning duct of square cross-section. The agreement between the viscous and inviscid computations was generally very good for the streamwise primary velocity and the radial secondary velocity, except at the walls, where slip conditions were specified for the inviscid code. The agreement between both the computations and the experimental data was not as close, especially at the 60.0 deg and 77.5 deg angular positions within the duct. This disagreement was attributed to incomplete modelling of the vortex development near the suction surface.

Comparison of Zgoubi and S-Code regarding the FFAG muon acceleration

International Nuclear Information System (INIS)

Fourrier, J.; Machida, S.

2006-06-01

The high flux accelerator based neutrino source is foreseen as one of the next generation facilities of particle physics. Called Neutrino Factory (NuFact), it will be based on a muon storage ring where muons will decay, creating high flux neutrino beams. Muons are supposed to be accelerated from 5 to 20 GeV before being injected into the storage ring. In that purpose, Fixed Field Alternating Gradient accelerators (FFAG) are one of the possibilities. Cell designs have been done and tracking studies are on their way using codes such as MAD, S-Code or Zgoubi. In order to cross-check results so obtained, we have performed comparisons between S-Code and Zgoubi at Rutherford Appleton Laboratory. The present report will explain the different simulations done and the results. The paper has the following contents: 1. Introduction; 2. Time of Flight comparisons; 3. Particle acceleration with different emittances; 4. Bunch acceleration. In conclusion it is shown that the difference of time of flight between the particle 1 and the others is comparable from Zgoubi to S- Code. Nevertheless, further investigation is necessary to make sure that the same revolution time will be found with the same initial conditions. Concerning the acceleration it was found that that the larger the emittance the less efficient appears to be the acceleration. Particles on ellipses 10 to 40 π mm rad are accelerated beyond 9 GeV while particles on ellipse 50 π mm rad are only accelerated up to 7.5 GeV and those on 60 π mm rad ellipse just reach 6 GeV. Thus, a beam whom emittance would be larger than 30 π mm rad would not be accelerated enough to reach 10 GeV and to be injected into the second FFAG. In the same way as for the first FFAG the larger the emittance the less efficient the acceleration. Particles on ellipses 10 to 40 π mm rad are accelerated beyond 18 GeV while particles on ellipses 50 and 60 π mm rad are only accelerated up to 12 GeV. Thus, beams whom emittance is lower than 30 π mm

Impact of dynamic rate coding aspects of mobile phone networks on forensic voice comparison.

Science.gov (United States)

Alzqhoul, Esam A S; Nair, Balamurali B T; Guillemin, Bernard J

2015-09-01

Previous studies have shown that landline and mobile phone networks are different in their ways of handling the speech signal, and therefore in their impact on it. But the same is also true of the different networks within the mobile phone arena. There are two major mobile phone technologies currently in use today, namely the global system for mobile communications (GSM) and code division multiple access (CDMA) and these are fundamentally different in their design. For example, the quality of the coded speech in the GSM network is a function of channel quality, whereas in the CDMA network it is determined by channel capacity (i.e., the number of users sharing a cell site). This paper examines the impact on the speech signal of a key feature of these networks, namely dynamic rate coding, and its subsequent impact on the task of likelihood-ratio-based forensic voice comparison (FVC). Surprisingly, both FVC accuracy and precision are found to be better for both GSM- and CDMA-coded speech than for uncoded. Intuitively one expects FVC accuracy to increase with increasing coded speech quality. This trend is shown to occur for the CDMA network, but, surprisingly, not for the GSM network. Further, in respect to comparisons between these two networks, FVC accuracy for CDMA-coded speech is shown to be slightly better than for GSM-coded speech, particularly when the coded-speech quality is high, but in terms of FVC precision the two networks are shown to be very similar. Copyright © 2015 The Chartered Society of Forensic Sciences. Published by Elsevier Ireland Ltd. All rights reserved.

Combinatorial neural codes from a mathematical coding theory perspective.

Science.gov (United States)

Curto, Carina; Itskov, Vladimir; Morrison, Katherine; Roth, Zachary; Walker, Judy L

2013-07-01

Shannon's seminal 1948 work gave rise to two distinct areas of research: information theory and mathematical coding theory. While information theory has had a strong influence on theoretical neuroscience, ideas from mathematical coding theory have received considerably less attention. Here we take a new look at combinatorial neural codes from a mathematical coding theory perspective, examining the error correction capabilities of familiar receptive field codes (RF codes). We find, perhaps surprisingly, that the high levels of redundancy present in these codes do not support accurate error correction, although the error-correcting performance of receptive field codes catches up to that of random comparison codes when a small tolerance to error is introduced. However, receptive field codes are good at reflecting distances between represented stimuli, while the random comparison codes are not. We suggest that a compromise in error-correcting capability may be a necessary price to pay for a neural code whose structure serves not only error correction, but must also reflect relationships between stimuli.

Recommendations for computer modeling codes to support the UMTRA groundwater restoration project

Energy Technology Data Exchange (ETDEWEB)

Tucker, M.D. [Sandia National Labs., Albuquerque, NM (United States); Khan, M.A. [IT Corp., Albuquerque, NM (United States)

1996-04-01

The Uranium Mill Tailings Remediation Action (UMTRA) Project is responsible for the assessment and remedial action at the 24 former uranium mill tailings sites located in the US. The surface restoration phase, which includes containment and stabilization of the abandoned uranium mill tailings piles, has a specific termination date and is nearing completion. Therefore, attention has now turned to the groundwater restoration phase, which began in 1991. Regulated constituents in groundwater whose concentrations or activities exceed maximum contaminant levels (MCLs) or background levels at one or more sites include, but are not limited to, uranium, selenium, arsenic, molybdenum, nitrate, gross alpha, radium-226 and radium-228. The purpose of this report is to recommend computer codes that can be used to assist the UMTRA groundwater restoration effort. The report includes a survey of applicable codes in each of the following areas: (1) groundwater flow and contaminant transport modeling codes, (2) hydrogeochemical modeling codes, (3) pump and treat optimization codes, and (4) decision support tools. Following the survey of the applicable codes, specific codes that can best meet the needs of the UMTRA groundwater restoration program in each of the four areas are recommended.

Recommendations for computer modeling codes to support the UMTRA groundwater restoration project

International Nuclear Information System (INIS)

Tucker, M.D.; Khan, M.A.

1996-04-01

The Uranium Mill Tailings Remediation Action (UMTRA) Project is responsible for the assessment and remedial action at the 24 former uranium mill tailings sites located in the US. The surface restoration phase, which includes containment and stabilization of the abandoned uranium mill tailings piles, has a specific termination date and is nearing completion. Therefore, attention has now turned to the groundwater restoration phase, which began in 1991. Regulated constituents in groundwater whose concentrations or activities exceed maximum contaminant levels (MCLs) or background levels at one or more sites include, but are not limited to, uranium, selenium, arsenic, molybdenum, nitrate, gross alpha, radium-226 and radium-228. The purpose of this report is to recommend computer codes that can be used to assist the UMTRA groundwater restoration effort. The report includes a survey of applicable codes in each of the following areas: (1) groundwater flow and contaminant transport modeling codes, (2) hydrogeochemical modeling codes, (3) pump and treat optimization codes, and (4) decision support tools. Following the survey of the applicable codes, specific codes that can best meet the needs of the UMTRA groundwater restoration program in each of the four areas are recommended

Comparison of computer codes related to the sodium oxide aerosol behavior in a containment building

International Nuclear Information System (INIS)

Fermandjian, J.

1984-09-01

In order to ensure that the problems of describing the physical behavior of sodium aerosols, during hypothetical fast reactor accidents, were adequately understood, a comparison of the computer codes (ABC/INTG, PNC, Japan; AEROSIM, UKAEA/SRD, United Kingdom; PARDISEKO IIIb, KfK, Germany; AEROSOLS/A2 and AEROSOLS/B1, CEA France) was undertaken in the frame of the CEC: exercise in which code users have run their own codes with a prearranged input

Comparison of the calculations of the stability properties of a specific stellarator equilibrium with different MHD stability codes

International Nuclear Information System (INIS)

Nakamura, Y.; Matsumoto, T.; Wakatani, M.; Ichiguchi, K.; Garcia, L.; Carreras, B.A.

1995-04-01

A particular configuration of the LHD stellarator with an unusually flat pressure profile has been chosen to be a test case for comparison of the MHD stability property predictions of different three-dimensional and averaged codes for the purpose of code comparison and validation. In particular, two relatively localized instabilities, the fastest growing modes with toroidal mode number n = 2 and n = 3 were studied using several different codes, with the good agreement that has been found providing justification for the use of any of them for equilibria of the type considered

Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code

International Nuclear Information System (INIS)

Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

2017-02-01

Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

Code-code comparisons of DIVIMP's 'onion-skin model' and the EDGE2D fluid code

International Nuclear Information System (INIS)

Stangeby, P.C.; Elder, J.D.; Horton, L.D.; Simonini, R.; Taroni, A.; Matthews, O.F.; Monk, R.D.

1997-01-01

In onion-skin modelling, O-SM, of the edge plasma, the cross-field power and particle flows are treated very simply e.g. as spatially uniform. The validity of O-S modelling requires demonstration that such approximations can still result in reasonable solutions for the edge plasma. This is demonstrated here by comparison of O-SM with full 2D fluid edge solutions generated by the EDGE2D code. The target boundary conditions for the O-SM are taken from the EDGE2D output and the complete O-SM solutions are then compared with the EDGE2D ones. Agreement is generally within 20% for n e , T e , T i and parallel particle flux density Γ for the medium and high recycling JET cases examined and somewhat less good for a strongly detached CMOD example. (orig.)

Comparison of SISEC code simulations with earthquake data of ordinary and base-isolated buildings

International Nuclear Information System (INIS)

Wang, C.Y.; Gvildys, J.

1991-01-01

At Argonne National Laboratory (ANL), a 3-D computer program SISEC (Seismic Isolation System Evaluation Code) is being developed for simulating the system response of isolated and ordinary structures (Wang et al. 1991). This paper describes comparison of SISEC code simulations with building response data of actual earthquakes. To ensure the accuracy of analytical simulations, recorded data of full-size reinforced concrete structures located in Sendai, Japan are used in this benchmark comparison. The test structures consist of two three-story buildings, one base-isolated and the other one ordinary founded. They were constructed side by side to investigate the effect of base isolation on the acceleration response. Among 20 earthquakes observed since April 1989, complete records of three representative earthquakes, no.2, no.6, and no.17, are used for the code validation presented in this paper. Correlations of observed and calculated accelerations at all instrument locations are made. Also, relative response characteristics of ordinary and isolated building structures are investigated. (J.P.N.)

Learning binary code via PCA of angle projection for image retrieval

Science.gov (United States)

Yang, Fumeng; Ye, Zhiqiang; Wei, Xueqi; Wu, Congzhong

2018-01-01

With benefits of low storage costs and high query speeds, binary code representation methods are widely researched for efficiently retrieving large-scale data. In image hashing method, learning hashing function to embed highdimensions feature to Hamming space is a key step for accuracy retrieval. Principal component analysis (PCA) technical is widely used in compact hashing methods, and most these hashing methods adopt PCA projection functions to project the original data into several dimensions of real values, and then each of these projected dimensions is quantized into one bit by thresholding. The variances of different projected dimensions are different, and with real-valued projection produced more quantization error. To avoid the real-valued projection with large quantization error, in this paper we proposed to use Cosine similarity projection for each dimensions, the angle projection can keep the original structure and more compact with the Cosine-valued. We used our method combined the ITQ hashing algorithm, and the extensive experiments on the public CIFAR-10 and Caltech-256 datasets validate the effectiveness of the proposed method.

Yucca Mountain Project thermal and mechanical codes first benchmark exercise: Part 3, Jointed rock mass analysis

International Nuclear Information System (INIS)

Costin, L.S.; Bauer, S.J.

1991-10-01

Thermal and mechanical models for intact and jointed rock mass behavior are being developed, verified, and validated at Sandia National Laboratories for the Yucca Mountain Site Characterization Project. Benchmarking is an essential part of this effort and is one of the tools used to demonstrate verification of engineering software used to solve thermomechanical problems. This report presents the results of the third (and final) phase of the first thermomechanical benchmark exercise. In the first phase of this exercise, nonlinear heat conduction code were used to solve the thermal portion of the benchmark problem. The results from the thermal analysis were then used as input to the second and third phases of the exercise, which consisted of solving the structural portion of the benchmark problem. In the second phase of the exercise, a linear elastic rock mass model was used. In the third phase of the exercise, two different nonlinear jointed rock mass models were used to solve the thermostructural problem. Both models, the Sandia compliant joint model and the RE/SPEC joint empirical model, explicitly incorporate the effect of the joints on the response of the continuum. Three different structural codes, JAC, SANCHO, and SPECTROM-31, were used with the above models in the third phase of the study. Each model was implemented in two different codes so that direct comparisons of results from each model could be made. The results submitted by the participants showed that the finite element solutions using each model were in reasonable agreement. Some consistent differences between the solutions using the two different models were noted but are not considered important to verification of the codes. 9 refs., 18 figs., 8 tabs

THE AGORA HIGH-RESOLUTION GALAXY SIMULATIONS COMPARISON PROJECT. II. ISOLATED DISK TEST

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji-hoon [Kavli Institute for Particle Astrophysics and Cosmology, SLAC National Accelerator Laboratory, Menlo Park, CA 94025 (United States); Agertz, Oscar [Department of Physics, University of Surrey, Guildford, Surrey, GU2 7XH (United Kingdom); Teyssier, Romain; Feldmann, Robert [Centre for Theoretical Astrophysics and Cosmology, Institute for Computational Science, University of Zurich, Zurich, 8057 (Switzerland); Butler, Michael J. [Max-Planck-Institut für Astronomie, D-69117 Heidelberg (Germany); Ceverino, Daniel [Zentrum für Astronomie der Universität Heidelberg, Institut für Theoretische Astrophysik, D-69120 Heidelberg (Germany); Choi, Jun-Hwan [Department of Astronomy, University of Texas, Austin, TX 78712 (United States); Keller, Ben W. [Department of Physics and Astronomy, McMaster University, Hamilton, ON L8S 4M1 (Canada); Lupi, Alessandro [Institut d’Astrophysique de Paris, Sorbonne Universites, UPMC Univ Paris 6 et CNRS, F-75014 Paris (France); Quinn, Thomas; Wallace, Spencer [Department of Astronomy, University of Washington, Seattle, WA 98195 (United States); Revaz, Yves [Institute of Physics, Laboratoire d’Astrophysique, École Polytechnique Fédérale de Lausanne, CH-1015 Lausanne (Switzerland); Gnedin, Nickolay Y. [Particle Astrophysics Center, Fermi National Accelerator Laboratory, Batavia, IL 60510 (United States); Leitner, Samuel N. [Department of Astronomy, University of Maryland, College Park, MD 20742 (United States); Shen, Sijing [Kavli Institute for Cosmology, University of Cambridge, Cambridge, CB3 0HA (United Kingdom); Smith, Britton D., E-mail: me@jihoonkim.org [Institute for Astronomy, University of Edinburgh, Royal Observatory, Edinburgh EH9 3HJ (United Kingdom); Collaboration: AGORA Collaboration; and others

2016-12-20

Using an isolated Milky Way-mass galaxy simulation, we compare results from nine state-of-the-art gravito-hydrodynamics codes widely used in the numerical community. We utilize the infrastructure we have built for the AGORA High-resolution Galaxy Simulations Comparison Project. This includes the common disk initial conditions, common physics models (e.g., radiative cooling and UV background by the standardized package Grackle) and common analysis toolkit yt, all of which are publicly available. Subgrid physics models such as Jeans pressure floor, star formation, supernova feedback energy, and metal production are carefully constrained across code platforms. With numerical accuracy that resolves the disk scale height, we find that the codes overall agree well with one another in many dimensions including: gas and stellar surface densities, rotation curves, velocity dispersions, density and temperature distribution functions, disk vertical heights, stellar clumps, star formation rates, and Kennicutt–Schmidt relations. Quantities such as velocity dispersions are very robust (agreement within a few tens of percent at all radii) while measures like newly formed stellar clump mass functions show more significant variation (difference by up to a factor of ∼3). Systematic differences exist, for example, between mesh-based and particle-based codes in the low-density region, and between more diffusive and less diffusive schemes in the high-density tail of the density distribution. Yet intrinsic code differences are generally small compared to the variations in numerical implementations of the common subgrid physics such as supernova feedback. Our experiment reassures that, if adequately designed in accordance with our proposed common parameters, results of a modern high-resolution galaxy formation simulation are more sensitive to input physics than to intrinsic differences in numerical schemes.

A comparison of the steady-state facility in the RELAP-UK code with the CUSH code and with collant flow in the Winfrith SGHWR

International Nuclear Information System (INIS)

Roberts, H.A.; Smith, C.P.

1976-02-01

Provision of capability for performing steady-state calculations in RELAP-UK has led to the possibility of the wider use of this code for steady-state assessments of the behaviour of commercial systems with complicated circuit arrangements. In the studies discussed in this report, the first objective is to demonstrate the performance of the RELAP-UK code in a steady-state role, and to make comparisons with the CUSH code, and with measurements obtained on the Winfrith Steam-Generating Heavy Water Reactor. (U.K.)

Verification of three dimensional triangular prismatic discrete ordinates transport code ENSEMBLE-TRIZ by comparison with Monte Carlo code GMVP

International Nuclear Information System (INIS)

Homma, Y.; Moriwaki, H.; Ikeda, K.; Ohdi, S.

2013-01-01

This paper deals with the verification of the 3 dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with the multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at the beginning of cycle of an initial core and at the beginning and the end of cycle of an equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multiplication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity. (authors)

Utilization of Relap 5 computer code for analyzing thermohydraulic projects

International Nuclear Information System (INIS)

Silva Filho, E.

1987-01-01

This work deals with the design of a scaled test facility of a typical pressurized water reactor plant of the 1300 MW (electric) class. A station blackout has been choosen to investigate the thermohydraulic behaviour of the the test facility in comparison to the reactor plant. The computer code RELAPS/MOD1 has been utilized to simulate the blackout and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermohydraulic behaviours of the two systems. (author) [pt

Comparison of results between different precision MAFIA codes

International Nuclear Information System (INIS)

Farkas, D.; Tice, B.

1990-01-01

In order to satisfy the inquiries of the Mafia Code users at SLAC, an evaluation of these codes was done. This consisted of running a cavity with known solutions. This study considered only the time independent solutions. No wake-field calculations were tried. The two machines involved were the NMFECC Cray (e-machine) at LLNL and the IBM/3081 at SLAC. The primary difference between the implementation of the codes on these machines is that the Cray has 64-bit accuracy while the IBM version has 32-bit accuracy. Unfortunately this study is incomplete as the Post-processor (P3) could not be made to work properly on the SLAC machine. This meant that no q's were calculated and no field patterns were generated. A certain amount of guessing had to be done when constructing the comparison tables. This problem aside, the probable conclusions that may be drawn are: (1) thirty-two bit precision is adequate for frequency determination; (2) sixty-four bit precision is desirable for field determination. This conclusion is deduced from the accuracy statistics. The cavity selected for study was a rectangular one with the dimensions (4,3,5) in centimeters. Only half of this cavity was used (2,3,5) with the x dimension being the one that was halved. The boundary conditions (B.C.) on the plane of symmetry were varied between Neumann and Dirichlet so as to cover all possible modes. Ten (10) modes were ran for each boundary condition

Coding in pigeons: Multiple-coding versus single-code/default strategies.

Science.gov (United States)

Pinto, Carlos; Machado, Armando

2015-05-01

To investigate the coding strategies that pigeons may use in a temporal discrimination tasks, pigeons were trained on a matching-to-sample procedure with three sample durations (2s, 6s and 18s) and two comparisons (red and green hues). One comparison was correct following 2-s samples and the other was correct following both 6-s and 18-s samples. Tests were then run to contrast the predictions of two hypotheses concerning the pigeons' coding strategies, the multiple-coding and the single-code/default. According to the multiple-coding hypothesis, three response rules are acquired, one for each sample. According to the single-code/default hypothesis, only two response rules are acquired, one for the 2-s sample and a "default" rule for any other duration. In retention interval tests, pigeons preferred the "default" key, a result predicted by the single-code/default hypothesis. In no-sample tests, pigeons preferred the key associated with the 2-s sample, a result predicted by multiple-coding. Finally, in generalization tests, when the sample duration equaled 3.5s, the geometric mean of 2s and 6s, pigeons preferred the key associated with the 6-s and 18-s samples, a result predicted by the single-code/default hypothesis. The pattern of results suggests the need for models that take into account multiple sources of stimulus control. © Society for the Experimental Analysis of Behavior.

Comparison of LIFE-4 and TEMECH code predictions with TREAT transient test data

International Nuclear Information System (INIS)

Gneiting, B.C.; Bard, F.E.; Hunter, C.W.

1984-09-01

Transient tests in the TREAT reactor were performed on FFTF Reference design mixed-oxide fuel pins, most of which had received prior steady-state irradiation in the EBR-II reactor. These transient test results provide a data base for calibration and verification of fuel performance codes and for evaluation of processes that affect pin damage during transient events. This paper presents a comparison of the LIFE-4 and TEMECH fuel pin thermal/mechanical analysis codes with the results from 20 HEDL TREAT experiments, ten of which resulted in pin failure. Both the LIFE-4 and TEMECH codes provided an adequate representation of the thermal and mechanical data from the TREAT experiments. Also, a criterion for 50% probability of pin failure was developed for each code using an average cumulative damage fraction value calculated for the pins that failed. Both codes employ the two major cladding loading mechanisms of differential thermal expansion and central cavity pressurization which were demonstrated by the test results. However, a detailed evaluation of the code predictions shows that the two code systems weigh the loading mechanism differently to reach the same end points of the TREAT transient results

Improvement of Computer Codes Used for Fuel Behaviour Simulation (FUMEX-III). Report of a Coordinated Research Project 2008-2012

International Nuclear Information System (INIS)

2013-03-01

It is fundamental to the future of nuclear power that reactors can be run safely and economically to compete with other forms of power generation. As a consequence, it is essential to develop the understanding of fuel performance and to embody that knowledge in codes to provide best estimate predictions of fuel behaviour. This in turn leads to a better understanding of fuel performance, a reduction in operating margins, flexibility in fuel management and improved operating economics. The IAEA has therefore embarked on a series of programmes addressing different aspects of fuel behaviour modelling with the following objectives: - To assess the maturity and prediction capabilities of fuel performance codes, and to support interaction and information exchange between countries with code development and application needs (FUMEX series); - To build a database of well defined experiments suitable for code validation in association with the OECD Nuclear Energy Agency (OECD/NEA); - To transfer a mature fuel modelling code to developing countries, to support teams in these countries in their efforts to adapt the code to the requirements of particular reactors, and to provide guidance on applying the code to reactor operation and safety assessments; - To provide guidelines for code quality assurance, code licensing and code application to fuel licensing. This report describes the results of the coordinated research project on the ''Improvement of computer codes used for fuel behaviour simulation (FUMEX-III)''. This programme was initiated in 2008 and completed in 2012. It followed previous programmes on fuel modelling: D-COM 1982-1984, FUMEX 1993-1996 and FUMEX-II 2002-2006. The participants used a mixture of data derived from commercial and experimental irradiation histories, in particular data designed to investigate the mechanical interactions occurring in fuel during normal, transient and severe transient operation. All participants carried out calculations on priority

Application of ADINAT/ADINA and MINIMEF codes to the resolution of the different exercises of COSA project

International Nuclear Information System (INIS)

1989-01-01

This report deals with the activities developed by the working group of the Escuela Tecnica Superior de Ingenieros de Minas de Madrid (E.T.S.I.M.M.) in collaboration with ENRESA within the framework of COSA project organised by the CEC. The ENRESA/ETSIMM working group became involved in the project under contract with the CEC the beginning of the phase named COSA II. For this reason it has been necessary to intensify the efforts in order to carry out, whithin the time available, the benchmarks corresponding to the previous phase COSA I. In this sense, the experience and results obtained by other european groups participating in the project, have proved to be very useful. The benchmarks cases for the different exercises were solved using the commercially available code ADINAT/ADINA and also the code developed by ETSIMM from the code named MINIMEF. In this report, in first place, a brief description of MINIMEF code is presented. A description of ADINA/ADINAT codes has not been included as these codes are well known. Next, the different benchmarks used, shall be described. The first two belong to the COSA I exercise while the latter is the one solved in the COSA II phase. All the results showed satisfactory as compared with those obtained by other working groups

Verification of the MOTIF code version 3.0

International Nuclear Information System (INIS)

Chan, T.; Guvanasen, V.; Nakka, B.W.; Reid, J.A.K.; Scheier, N.W.; Stanchell, F.W.

1996-12-01

As part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), AECL has developed a three-dimensional finite-element code, MOTIF (Model Of Transport In Fractured/ porous media), for detailed modelling of groundwater flow, heat transport and solute transport in a fractured rock mass. The code solves the transient and steady-state equations of groundwater flow, solute (including one-species radionuclide) transport, and heat transport in variably saturated fractured/porous media. The initial development was completed in 1985 (Guvanasen 1985) and version 3.0 was completed in 1986. This version is documented in detail in Guvanasen and Chan (in preparation). This report describes a series of fourteen verification cases which has been used to test the numerical solution techniques and coding of MOTIF, as well as demonstrate some of the MOTIF analysis capabilities. For each case the MOTIF solution has been compared with a corresponding analytical or independently developed alternate numerical solution. Several of the verification cases were included in Level 1 of the International Hydrologic Code Intercomparison Project (HYDROCOIN). The MOTIF results for these cases were also described in the HYDROCOIN Secretariat's compilation and comparison of results submitted by the various project teams (Swedish Nuclear Power Inspectorate 1988). It is evident from the graphical comparisons presented that the MOTIF solutions for the fourteen verification cases are generally in excellent agreement with known analytical or numerical solutions obtained from independent sources. This series of verification studies has established the ability of the MOTIF finite-element code to accurately model the groundwater flow and solute and heat transport phenomena for which it is intended. (author). 20 refs., 14 tabs., 32 figs

Code Sharing and Collaboration: Experiences From the Scientist's Expert Assistant Project and Their Relevance to the Virtual Observatory

Science.gov (United States)

Korathkar, Anuradha; Grosvenor, Sandy; Jones, Jeremy; Li, Connie; Mackey, Jennifer; Neher, Ken; Obenschain, Arthur F. (Technical Monitor)

2001-01-01

In the Virtual Observatory (VO), software tools will perform the functions that have traditionally been performed by physical observatories and their instruments. These tools will not be adjuncts to VO functionality but will make up the very core of the VO. Consequently, the tradition of observatory and system independent tools serving a small user base is not valid for the VO. For the VO to succeed, we must improve software collaboration and code sharing between projects and groups. A significant goal of the Scientist's Expert Assistant (SEA) project has been promoting effective collaboration and code sharing among groups. During the past three years, the SEA project has been developing prototypes for new observation planning software tools and strategies. Initially funded by the Next Generation Space Telescope, parts of the SEA code have since been adopted by the Space Telescope Science Institute. SEA has also supplied code for the SIRTF (Space Infrared Telescope Facility) planning tools, and the JSky Open Source Java library. The potential benefits of sharing code are clear. The recipient gains functionality for considerably less cost. The provider gains additional developers working with their code. If enough users groups adopt a set of common code and tools, de facto standards can emerge (as demonstrated by the success of the FITS standard). Code sharing also raises a number of challenges related to the management of the code. In this talk, we will review our experiences with SEA--both successes and failures, and offer some lessons learned that might promote further successes in collaboration and re-use.

A comparison of two nodal codes : Advanced nodal code (ANC) and analytic function expansion nodal (AFEN) code

International Nuclear Information System (INIS)

Chung, S.K.; Hah, C.J.; Lee, H.C.; Kim, Y.H.; Cho, N.Z.

1996-01-01

Modern nodal methods usually employs the transverse integration technique in order to reduce a multi-dimensional diffusion equation to one-dimensional diffusion equations. The use of the transverse integration technique requires two major approximations such as a transverse leakage approximation and a one-dimensional flux approximation. Both the transverse leakage and the one-dimensional flux are approximated by polynomials. ANC (Advanced Nodal Code) developed by Westinghouse employs a modern nodal expansion method for the flux calculation, the equivalence theory for the homogenization error reduction and a group theory for pin power recovery. Unlike the conventional modern nodal methods, AFEN (Analytic Function Expansion Nodal) method expands homogeneous flux distributions within a node into non-separable analytic basis functions, which eliminate two major approximations of the modern nodal methods. A comparison study of AFEN with ANC has been performed to see the applicability of AFEN to commercial PWR and different types of reactors such as MOX fueled reactor. The qualification comparison results demonstrate that AFEN methodology is accurate enough to apply for commercial PWR analysis. The results show that AFEN provides very accurate results (core multiplication factor and assembly power distribution) for cores that exhibit strong flux gradients as in a MOX loaded core. (author)

Application of the integral code MELCOR for German NPPs and use within accident management and PSA projects

International Nuclear Information System (INIS)

Sonnenkalb, Martin

2006-01-01

The paper summarizes the application of MELCOR to German NPPS with PWR and BWR. A development of different code systems like ATHLET/ATHLET-CD, COCOSYS and ASTEC is done as well at GRS but it is not discussed in this paper. GRS has been using MELCOR since 1990 for real plant calculations. The results of MELCOR analyses are used mainly in PSA level 2 studies and in Accident Management projects for both types of NPPs. MELCOR has been a very useful and robust tool for these analyses. The calculations performed within the PSA level 2 studies for both types of German NPPs have shown that typical severe accident scenarios are characterized by several phases and that the consideration of plant specifics are important not only for realistic source term calculations. An overview of typically severe accident phases together with main accident management measures installed in German NPPs is presented in the paper. Several severe accident sequences have been calculated for both reactor types and some detailed nodalisation studies and code to code comparisons have been prepared in the past, to prove the developed core, reactor circuit and containment/building nodalisation schemes. Together with the compilation of the MELCOR data set, the qualification of the nodalisation schemes has been pursued with comparative calculations with detailed GRS codes for selected phases of severe accidents. The results of these comparative analyses showed in most of the areas a good agreement of essential parameters and of the general description of the plant behaviour during the accident progression. The in general detail of the German plant nodalisation schemes developed for MELCOR contributes significantly to this good agreement between integral and detailed code results. The implementation of MELCOR into the GRS simulator ATLAS was very important for the assessment of the results, not only due to the great detail of the nodalisation schemes used. It is used for training of severe accident

Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

2017-02-15

Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

The CEDAR Project

CERN Document Server

Butterworth, J M; Waugh, B M; Stirling, W J; Whalley, M R

2005-01-01

We describe the plans and objectives of the CEDAR project (Combined e-Science Data Analysis Resource for High Energy Physics) newly funded by the PPARC e-Science programme in the UK. CEDAR will combine the strengths of the well established and widely used HEPDATA database of HEP data and the innovative JetWeb data/Monte Carlo comparison facility, built on the HZTOOL package, and will exploit developing grid technology. The current status and future plans of both of these individual sub-projects within the CEDAR framework are described, showing how they will cohesively provide (a) an extensive archive of Reaction Data, (b) validation and tuning of Monte Carlo programs against these reaction data sets, and (c) a validated code repository for a wide range of HEP code such as parton distribution functions and other calculation codes used by particle physicists. Once established it is envisaged CEDAR will become an important Grid tool used by LHC experimentalists in their analyses and may well serve as a model in ...

Comparison of PWR-IMF and FR fuel cycles

International Nuclear Information System (INIS)

Darilek, Petr; Zajac, Radoslav; Breza, Juraj; Necas, Vladimir

2007-01-01

The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

Assessment of the system code DRUFAN/ATHLET using results of LOBI tests

International Nuclear Information System (INIS)

Burwell, J.M.; Kirmse, R.E.; Kyncl, M.; Malhotra, P.K.

1989-09-01

Four post-test analyses have been performed by GRS within the Shared Cost Action Programme (SCAP) sponsored by the Commission of the European Communities (contract 3015-86-07 EL ISP D) and by the Bundesminister fuer Forschung und Technologie of the Federal Republic of Germany (Research project RS 739). The four tests were mutually selected by the contractors (CEA, GRS, IKE, Univ. Pisa) of activity No. 3 and by the project organizer. Some of the tests were selected to be analyzed by more than one participant in order to allow comparison between analytical results obtained with different codes or obtained by different code-users. DRUFAN/ATHLET verification analyses were performed by IKE too. The four tests selected for the GRS activity are: - A2-77A (Natural Circulation Test), Analysis with ATHLET - A1-76 (Steam Generator Performance Test), Analysis with DRUFAN - BL-01 (Intermediate Leak), Analysis with ATHLET - A2-81 (Small Leak), Analysis with ATHLET. This final report contains the results of the four post test analysis including the comparison between measured and calculated quantities and the description of the applied codes, the selected model of the LOBI facility and the conclusions drawn for the improvement of the codes models

MELCOR Accident Consequence Code System (MACCS)

International Nuclear Information System (INIS)

Chanin, D.I.; Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems

MELCOR Accident Consequence Code System (MACCS)

Energy Technology Data Exchange (ETDEWEB)

Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

Comparison of neutron diffusion theory codes in two and three space dimensions using a sodium cooled fast reactor benchmark

International Nuclear Information System (INIS)

Butland, A.T.D.; Putney, J.; Sweet, D.W.

1980-04-01

This report describes work performed to compare two UK neutron diffusion theory codes, TIGAR and SNAP, with published results for eight other codes available abroad. Both mesh edge and mesh centred finite difference diffusion theory codes as well as one axial synthesis code are included in the comparison and a range of iteration procedures are used by them. Comparison is made of calculations for a model of the sodium cooled fast reactor SNR-300 in both triangular and rectangular geometry and for a range of spatial meshes, enabling extrapolations to infinite mesh to be made. Calculated values of the effective multiplication constant, keff, for all the codes, agree very well when extrapolated to infinite mesh, indicating that no significant errors arising from the finite difference approximation but independent of mesh spacing are present in the calculations. The variation of keff with mesh area is found to be linear for the small meshes considered here, with the gradients for the mesh centred and mesh edged codes being of opposite sign. The results obtained using the mesh centred codes TIGAR, SNAP and CITATION agree closely with one another for all the meshes considered; the mesh edge codes agree less closely. (author)

Results of aerosol code comparisons with releases from ACE MCCI tests

International Nuclear Information System (INIS)

Fink, J.K.; Corradini, M.; Hidaka, A.; Hontanon, E.; Mignanelli, M.A.; Schroedl, E.; Strizhov, V.

1992-01-01

Results of aerosol release calculations by six groups from six countries are compared with the releases from ACE MCCI Test L6. The codes used for these calculations included: SOLGASMIX-PV, SOLGASMIX Reactor 1986, CORCON.UW, VANESA 1.01, and CORCON mod2.04/VANESA 1.01. Calculations were performed with the standard VANESA 1.01 code and with modifications to the VANESA code such as the inclusion of various zirconium-silica chemical reactions. Comparisons of results from these calculations were made with Test L6 release fractions for U, Zr, Si, the fission-product elements Te, Ba, Sr, Ce, La, Mo and control materials Ag, In, and Ru. Reasonable agreement was obtained between calculations and Test L6 results for the volatile elements Ag, In and Te. Calculated releases of the low volatility fission products ranged from within an order of magnitude to five orders of magnitude of Test L6 values. Releases were over and underestimated by calculations. Poorest agreements were obtained for Mo and Si

A comparison of two fully coupled codes for integrated dynamic analysis of floating vertical axis wind turbines

NARCIS (Netherlands)

Koppenol, Boy; Cheng, Zhengshun; Gao, Zhen; Simao Ferreira, C.; Moan, T; Tande, John Olav Giæver; Kvamsdal, Trond; Muskulus, Michael

2017-01-01

This paper presents a comparison of two state-of-the-art codes that are capable of modelling floating vertical axis wind turbines (VAWTs) in fully coupled time-domain simulations, being the HAWC2 by DTU and the SIMO-RIFLEX-AC code by NTNU/MARINTEK. The comparative study focusses on the way

Fuel model studies. Comparison of our present version of GAPCON-THERMAL-2 with results from the EPRI code comparison study. Partial report

International Nuclear Information System (INIS)

Malen, K.; Jansson, L.

1978-08-01

Runs with our present version of GAPCON-THERMAL-2 have been compared to results from the EPRI code comparison study. Usually also our version of GAPCON predicts high temperatures, 100-300 K or 10-15% higher than average code predictions and experimental results. The well-known temperaturegas release instablility is found also with GAPCON. In this case one identifies the gas release limits 1400 deg C and 1700 deg C as instablility points. (author)

Comparisons of the simulation results using different codes for ADS spallation target

International Nuclear Information System (INIS)

Yu Hongwei; Fan Sheng; Shen Qingbiao; Zhao Zhixiang; Wan Junsheng

2002-01-01

The calculations to the standard thick target were made by using different codes. The simulation of the thick Pb target with length of 60 cm, diameter of 20 cm bombarded with 800, 1000, 1500 and 2000 MeV energetic proton beam was carried out. The yields and the spectra of emitted neutron were studied. The spallation target was simulated by SNSP, SHIELD, DCM/CEM (Dubna Cascade Model /Cascade Evaporation Mode) and LAHET codes. The Simulation Results were compared with experiments. The comparisons show good agreement between the experiments and the SNSP simulated leakage neutron yield. The SHIELD simulated leakage neutron spectra are in good agreement with the LAHET and the DCM/CEM simulated leakage neutron spectra

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

Comparison Results on the Basic Phenomena between CAP and CONTEMPT-LT

Energy Technology Data Exchange (ETDEWEB)

Choo, Yeon Joon; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul [FNC Tech., SNU, Seoul (Korea, Republic of); Ha, Sang Jun; Choi, Hoon [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2011-10-15

A development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. As a part of this project, CAP (Containment Analysis Package) code has been developing for the containment safety and performance analysis side by side with SPACE. CAP Beta version has been released lately and validation processes are under way currently. Code by code comparison activity is scheduled in the validation processes and the first comparable code is CONTEMPT-LT. CONTEMPT-LT was developed to predict the longterm behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-ofcoolant accident (LOCA) conditions. CONTEMPT-LT calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP-LT has the similar code features and it, therefore, is expected to show the similar analysis performance with CAP. In this study, the code performances were compared for the same phenomena between CAP and CONTEMPT-LT. Code comparison is carried out through two stages; separate and integral effect comparison

Comparison Results on the Basic Phenomena between CAP and CONTEMPT-LT

International Nuclear Information System (INIS)

Choo, Yeon Joon; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul; Ha, Sang Jun; Choi, Hoon

2011-01-01

A development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. As a part of this project, CAP (Containment Analysis Package) code has been developing for the containment safety and performance analysis side by side with SPACE. CAP Beta version has been released lately and validation processes are under way currently. Code by code comparison activity is scheduled in the validation processes and the first comparable code is CONTEMPT-LT. CONTEMPT-LT was developed to predict the longterm behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-ofcoolant accident (LOCA) conditions. CONTEMPT-LT calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments, leakage on containment response. Models are provided for fan cooler and cooling spray as engineered safety systems. Any compartment may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. As mentioned above, CONTEMP-LT has the similar code features and it, therefore, is expected to show the similar analysis performance with CAP. In this study, the code performances were compared for the same phenomena between CAP and CONTEMPT-LT. Code comparison is carried out through two stages; separate and integral effect comparison

Dual Coding, Reasoning and Fallacies.

Science.gov (United States)

Hample, Dale

1982-01-01

Develops the theory that a fallacy is not a comparison of a rhetorical text to a set of definitions but a comparison of one person's cognition with another's. Reviews Paivio's dual coding theory, relates nonverbal coding to reasoning processes, and generates a limited fallacy theory based on dual coding theory. (PD)

Fracture flow code

International Nuclear Information System (INIS)

Dershowitz, W; Herbert, A.; Long, J.

1989-03-01

The hydrology of the SCV site will be modelled utilizing discrete fracture flow models. These models are complex, and can not be fully cerified by comparison to analytical solutions. The best approach for verification of these codes is therefore cross-verification between different codes. This is complicated by the variation in assumptions and solution techniques utilized in different codes. Cross-verification procedures are defined which allow comparison of the codes developed by Harwell Laboratory, Lawrence Berkeley Laboratory, and Golder Associates Inc. Six cross-verification datasets are defined for deterministic and stochastic verification of geometric and flow features of the codes. Additional datasets for verification of transport features will be documented in a future report. (13 figs., 7 tabs., 10 refs.) (authors)

MELCOR Accident Consequence Code System (MACCS)

International Nuclear Information System (INIS)

Rollstin, J.A.; Chanin, D.I.; Jow, H.N.

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projections, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management

MELCOR Accident Consequence Code System (MACCS)

Energy Technology Data Exchange (ETDEWEB)

Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

MELCOR Accident Consequence Code System (MACCS)

International Nuclear Information System (INIS)

Jow, H.N.; Sprung, J.L.; Ritchie, L.T.; Rollstin, J.A.; Chanin, D.I.

1990-02-01

This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs

The FORTRAN NALAP code adapted to a microcomputer compiler

International Nuclear Information System (INIS)

Lobo, Paulo David de Castro; Borges, Eduardo Madeira; Braz Filho, Francisco Antonio; Guimaraes, Lamartine Nogueira Frutuoso

2010-01-01

The Nuclear Energy Division of the Institute for Advanced Studies (IEAv) is conducting the TERRA project (TEcnologia de Reatores Rapidos Avancados), Technology for Advanced Fast Reactors project, aimed at a space reactor application. In this work, to attend the TERRA project, the NALAP code adapted to a microcomputer compiler called Compaq Visual Fortran (Version 6.6) is presented. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic responses for sodium cooled fast reactors. The strategy to run the code in a PC was divided in some steps mainly to remove unnecessary routines, to eliminate old statements, to introduce new ones and also to include extension precision mode. The source program was able to solve three sample cases under conditions of protected transients suggested in literature: the normal reactor shutdown, with a delay of 200 ms to start the control rod movement and a delay of 500 ms to stop the pumps; reactor scram after transient of loss of flow; and transients protected from overpower. Comparisons were made with results from the time when the NALAP code was acquired by the IEAv, back in the 80's. All the responses for these three simulations reproduced the calculations performed with the CDC compiler in 1985. Further modifications will include the usage of gas as coolant for the nuclear reactor to allow a Closed Brayton Cycle Loop - CBCL - to be used as a heat/electric converter. (author)

The FORTRAN NALAP code adapted to a microcomputer compiler

Energy Technology Data Exchange (ETDEWEB)

Lobo, Paulo David de Castro; Borges, Eduardo Madeira; Braz Filho, Francisco Antonio; Guimaraes, Lamartine Nogueira Frutuoso, E-mail: plobo.a@uol.com.b, E-mail: eduardo@ieav.cta.b, E-mail: fbraz@ieav.cta.b, E-mail: guimarae@ieav.cta.b [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)

2010-07-01

The Nuclear Energy Division of the Institute for Advanced Studies (IEAv) is conducting the TERRA project (TEcnologia de Reatores Rapidos Avancados), Technology for Advanced Fast Reactors project, aimed at a space reactor application. In this work, to attend the TERRA project, the NALAP code adapted to a microcomputer compiler called Compaq Visual Fortran (Version 6.6) is presented. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic responses for sodium cooled fast reactors. The strategy to run the code in a PC was divided in some steps mainly to remove unnecessary routines, to eliminate old statements, to introduce new ones and also to include extension precision mode. The source program was able to solve three sample cases under conditions of protected transients suggested in literature: the normal reactor shutdown, with a delay of 200 ms to start the control rod movement and a delay of 500 ms to stop the pumps; reactor scram after transient of loss of flow; and transients protected from overpower. Comparisons were made with results from the time when the NALAP code was acquired by the IEAv, back in the 80's. All the responses for these three simulations reproduced the calculations performed with the CDC compiler in 1985. Further modifications will include the usage of gas as coolant for the nuclear reactor to allow a Closed Brayton Cycle Loop - CBCL - to be used as a heat/electric converter. (author)

Comparison of ATF and TJ-II stellarator equilibria as computed by the 3-D VMEC and PIES codes

International Nuclear Information System (INIS)

Johnson, J.L.; Monticello, D.A.; Reiman, A.H.; Hirshman, S.P.

1992-01-01

A comparison is made of results from the PIES code, which determines the equilibrium properties of three-dimensional toroidal configurations by direct integration along the magnetic field lines, with those from the VMEC code, which uses an energy minimization in a flux representation to determine the equilibrium configuration, for two devices: the ATF stellarator at Oak Ridge and the TJ-11 heliac which is being built in Madrid. The results obtained from the two codes are in good agreement, providing additional validation for the codes

Ready, steady… Code!

CERN Multimedia

Anaïs Schaeffer

2013-01-01

This summer, CERN took part in the Google Summer of Code programme for the third year in succession. Open to students from all over the world, this programme leads to very successful collaborations for open source software projects.   Image: GSoC 2013. Google Summer of Code (GSoC) is a global programme that offers student developers grants to write code for open-source software projects. Since its creation in 2005, the programme has brought together some 6,000 students from over 100 countries worldwide. The students selected by Google are paired with a mentor from one of the participating projects, which can be led by institutes, organisations, companies, etc. This year, CERN PH Department’s SFT (Software Development for Experiments) Group took part in the GSoC programme for the third time, submitting 15 open-source projects. “Once published on the Google Summer for Code website (in April), the projects are open to applications,” says Jakob Blomer, one of the o...

Japanese national project for establishment of codes and standards for stationary PEFC system

International Nuclear Information System (INIS)

Sumi, S.; Ohmura, T.; Yamaguchi, R.; Kikuzawa, H.

2003-01-01

For the purpose of practical utilization of the PEFC cogeneration system, we are promoting the national projects of the 'Establishment of Codes and Standards for Stationary PEFC System'. The objective is to prepare the software platforms for wide spreading use, which are required in the introduction stage of the PEFC cogeneration systems, such as code and standards for safety, reliability, performance and so on. For this objective, using test samples of the systems and the stacks, developments of test and evaluation devices, collection of various kinds of data and establishment of test and evaluation methods are under way. (author)

A Comparison of Source Code Plagiarism Detection Engines

Science.gov (United States)

Lancaster, Thomas; Culwin, Fintan

2004-06-01

Automated techniques for finding plagiarism in student source code submissions have been in use for over 20 years and there are many available engines and services. This paper reviews the literature on the major modern detection engines, providing a comparison of them based upon the metrics and techniques they deploy. Generally the most common and effective techniques are seen to involve tokenising student submissions then searching pairs of submissions for long common substrings, an example of what is defined to be a paired structural metric. Computing academics are recommended to use one of the two Web-based detection engines, MOSS and JPlag. It is shown that whilst detection is well established there are still places where further research would be useful, particularly where visual support of the investigation process is possible.

Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 2

International Nuclear Information System (INIS)

Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

1993-08-01

The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

Accident and safety analyses for the HTR-modul. Partial project 1: Computer codes for system behaviour calculation. Final report. Pt. 1

International Nuclear Information System (INIS)

Lohnert, G.; Becker, D.; Dilcher, L.; Doerner, G.; Feltes, W.; Gysler, G.; Haque, H.; Kindt, T.; Kohtz, N.; Lange, L.; Ragoss, H.

1993-08-01

The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP) [de

Results from the First Validation Phase of CAP code

International Nuclear Information System (INIS)

Choo, Yeon Joon; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul; Ha, Sang Jun; Choi, Hoon

2010-01-01

The second stage of Safety Analysis Code Development for Nuclear Power Plants was lunched on Apirl, 2010 and is scheduled to be through 2012, of which the scope of work shall cover from code validation to licensing preparation. As a part of this project, CAP(Containment Analysis Package) will follow the same procedures. CAP's validation works are organized hieratically into four validation steps using; 1) Fundamental phenomena. 2) Principal phenomena (mixing and transport) and components in containment. 3) Demonstration test by small, middle, large facilities and International Standard Problems. 4) Comparison with other containment codes such as GOTHIC or COMTEMPT. In addition, collecting the experimental data related to containment phenomena and then constructing the database is one of the major works during the second stage as a part of this project. From the validation process of fundamental phenomenon, it could be expected that the current capability and the future improvements of CAP code will be revealed. For this purpose, simple but significant problems, which have the exact analytical solution, were selected and calculated for validation of fundamental phenomena. In this paper, some results of validation problems for the selected fundamental phenomena will be summarized and discussed briefly

Comparison of aerosol behavior codes with experimental results from a sodium fire in a containment

International Nuclear Information System (INIS)

Lhiaubet, G.; Kissane, M.P.; Seino, H.; Miyake, O.; Himeno, Y.

1990-01-01

The containment expert group (CONT), a subgroup of the CEC fast reactor Safety Working Group (SWG), has carried out several studies on the behavior of sodium aerosols which might form in a severe fast reactor accident during which primary sodium leaks into the secondary containment. These studies comprise an intercalibration of measurement devices used to determine the aerosol particle size spectrum, and the analysis and comparison of codes applied to the determination of aerosol behavior in a reactor containment. The paper outlines the results of measurements of typical data made for aerosols produced in a sodium fire and their comparison with results from different codes (PARDISEKO, AEROSIM, CONTAIN, AEROSOLS/B2). The sodium fire experiment took place at CEN-Cadarache (France) in a 400 m 3 vessel. The fire lasted 90 minutes and the aerosol measurements were made over 10 hours at different locations inside the vessel. The results showed that the suspended mass calculated along the time with different codes was in good agreement with the experiment. However, the calculated aerosol deposition on the walls was diverging and always significantly lower than the measured values

Light-water reactor safety analysis codes

International Nuclear Information System (INIS)

Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

1980-01-01

A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

Experimental research and comparison of LDPC and RS channel coding in ultraviolet communication systems.

Science.gov (United States)

Wu, Menglong; Han, Dahai; Zhang, Xiang; Zhang, Feng; Zhang, Min; Yue, Guangxin

2014-03-10

We have implemented a modified Low-Density Parity-Check (LDPC) codec algorithm in ultraviolet (UV) communication system. Simulations are conducted with measured parameters to evaluate the LDPC-based UV system performance. Moreover, LDPC (960, 480) and RS (18, 10) are implemented and experimented via a non-line-of-sight (NLOS) UV test bed. The experimental results are in agreement with the simulation and suggest that based on the given power and 10(-3)bit error rate (BER), in comparison with an uncoded system, average communication distance increases 32% with RS code, while 78% with LDPC code.

Comparison study of inelastic analysis codes for high temperature structure

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Bum; Lee, H. Y.; Park, C. K.; Geon, G. P.; Lee, J. H

2004-02-01

LMR high temperature structures subjected to operating and transient loadings may exhibit very complex deformation behaviors due to the use of ductile material such as 316SS and the systematic analysis technology of high temperature structure for reliable safety assessment is essential. In this project, comparative study with developed inelastic analysis program NONSTA and the existing analysis codes was performed applying various types of loading including non-proportional loading. The performance of NONSTA was confirmed and the effect of inelastic constants on the analysis result was analyzed. Also, the applicability of the inelastic analysis was enlarged as a result of applying both the developed program and the existing codes to the analyses of the enhanced creep behavior and the elastic follow-up behavior of high temperature structures and the necessary items for improvements were deduced. Further studies on the improvement of NONSTA program and the decision of the proper values of inelastic constants are necessary.

The use of the codes from MCU family for calculations of WWER type reactors

International Nuclear Information System (INIS)

Abagijan, L.P.; Alexeyev, N.I.; Bryzgalov, V.I.; Gomin, E.A.; Glushkov, A.E.; Gorodkov, S.S.; Gurevich, M.I.; Kalugin, M.A.; Marin, S.V.; Shkarovsky, D.A.; Yudkevich, M.S.

2000-01-01

The MCU-RFFI/A and MCU-REA codes developed within the framework of the long term MCU project are widely used for calculations of neutron physic characteristics of WWER type reactors. Complete descriptions of the codes are available in both Russian and English. The codes are verified and validated by means of the comparison of calculated results with experimental data and mathematical benchmarks. The codes are licensed by Russian Nuclear and Criticality Safety Regulatory Body (Gosatomnadzor RF) (Code Passports: N 61 of 17.10.1966 and N 115 of 02.03.2000 accordingly)). The report gives examples of WWER reactor physic tasks important for practice solved using the codes from the MCU family. Some calculational results are given too. (Authors)

SPACE CHARGE SIMULATION METHODS INCORPORATED IN SOME MULTI - PARTICLE TRACKING CODES AND THEIR RESULTS COMPARISON

International Nuclear Information System (INIS)

BEEBE - WANG, J.; LUCCIO, A.U.; D IMPERIO, N.; MACHIDA, S.

2002-01-01

Space charge in high intensity beams is an important issue in accelerator physics. Due to the complicity of the problems, the most effective way of investigating its effect is by computer simulations. In the resent years, many space charge simulation methods have been developed and incorporated in various 2D or 3D multi-particle-tracking codes. It has becoming necessary to benchmark these methods against each other, and against experimental results. As a part of global effort, we present our initial comparison of the space charge methods incorporated in simulation codes ORBIT++, ORBIT and SIMPSONS. In this paper, the methods included in these codes are overviewed. The simulation results are presented and compared. Finally, from this study, the advantages and disadvantages of each method are discussed

SPACE CHARGE SIMULATION METHODS INCORPORATED IN SOME MULTI - PARTICLE TRACKING CODES AND THEIR RESULTS COMPARISON.

Energy Technology Data Exchange (ETDEWEB)

BEEBE - WANG,J.; LUCCIO,A.U.; D IMPERIO,N.; MACHIDA,S.

2002-06-03

Space charge in high intensity beams is an important issue in accelerator physics. Due to the complicity of the problems, the most effective way of investigating its effect is by computer simulations. In the resent years, many space charge simulation methods have been developed and incorporated in various 2D or 3D multi-particle-tracking codes. It has becoming necessary to benchmark these methods against each other, and against experimental results. As a part of global effort, we present our initial comparison of the space charge methods incorporated in simulation codes ORBIT++, ORBIT and SIMPSONS. In this paper, the methods included in these codes are overviewed. The simulation results are presented and compared. Finally, from this study, the advantages and disadvantages of each method are discussed.

Comparisons for ESTA-Task3: ASTEC, CESAM and CLÉS

Science.gov (United States)

Christensen-Dalsgaard, J.

The ESTA activity under the CoRoT project aims at testing the tools for computing stellar models and oscillation frequencies that will be used in the analysis of asteroseismic data from CoRoT and other large-scale upcoming asteroseismic projects. Here I report results of comparisons between calculations using the Aarhus code (ASTEC) and two other codes, for models that include diffusion and settling. It is found that there are likely deficiencies, requiring further study, in the ASTEC computation of models including convective cores.

Comparison of two LES codes for wind turbine wake studies

International Nuclear Information System (INIS)

Sarlak, H; Mikkelsen, R; Sørensen, J N; Pierella, F

2014-01-01

For the third time a blind test comparison in Norway 2013, was conducted comparing numerical simulations for the rotor C p and C t and wake profiles with the experimental results. As the only large eddy simulation study among participants, results of the Technical University of Denmark (DTU) using their in-house CFD solver, EllipSys3D, proved to be more reliable among the other models for capturing the wake profiles and the turbulence intensities downstream the turbine. It was therefore remarked in the workshop to investigate other LES codes to compare their performance with EllipSys3D. The aim of this paper is to investigate on two CFD solvers, the DTU's in-house code, EllipSys3D and the open-sourse toolbox, OpenFoam, for a set of actuator line based LES computations. Two types of simulations are performed: the wake behind a signle rotor and the wake behind a cluster of three inline rotors. Results are compared in terms of velocity deficit, turbulence kinetic energy and eddy viscosity. It is seen that both codes predict similar near-wake flow structures with the exception of OpenFoam's simulations without the subgrid-scale model. The differences begin to increase with increasing the distance from the upstream rotor. From the single rotor simulations, EllipSys3D is found to predict a slower wake recovery in the case of uniform laminar flow. From the 3-rotor computations, it is seen that the difference between the codes is smaller as the disturbance created by the downstream rotors causes break down of the wake structures and more homogenuous flow structures. It is finally observed that OpenFoam computations are more sensitive to the SGS models

The fast code

Energy Technology Data Exchange (ETDEWEB)

Freeman, L.N.; Wilson, R.E. [Oregon State Univ., Dept. of Mechanical Engineering, Corvallis, OR (United States)

1996-09-01

The FAST Code which is capable of determining structural loads on a flexible, teetering, horizontal axis wind turbine is described and comparisons of calculated loads with test data are given at two wind speeds for the ESI-80. The FAST Code models a two-bladed HAWT with degrees of freedom for blade bending, teeter, drive train flexibility, yaw, and windwise and crosswind tower motion. The code allows blade dimensions, stiffnesses, and weights to differ and models tower shadow, wind shear, and turbulence. Additionally, dynamic stall is included as are delta-3 and an underslung rotor. Load comparisons are made with ESI-80 test data in the form of power spectral density, rainflow counting, occurrence histograms, and azimuth averaged bin plots. It is concluded that agreement between the FAST Code and test results is good. (au)

Code Glosses in Academic Writing: The Comparison of Iranian and Native Authors

Directory of Open Access Journals (Sweden)

Mahdi Dehghan

2016-01-01

Full Text Available One of the challenges of academic writing is the creation of a structurally and communicatively well-organized and coherent text. Metadiscourse enables authors of journal articles to achieve this goal by raising the writers‘ awareness about discourse features which can contribute to a better academic content production. Also, L1 background has always been a hot topic in applied linguistics and native versus non-native comparisons have been of particular interest in this field. The purpose of this study was to see whether native English speakers and Iranians use code glosses as a sub-category of metadiscourse similarly in their academic writings. To this end, the introduction section of 30 journal articles written by native and Iranian non-native English authors were investigated and the number of code-glosses in each group was counted and analyzed. Both descriptive and inferential statistics were used to answer the research question of this study. The result of data analysis revealed that there was no significant difference between the frequency count of code glosses used by Iranian and native English authors. This study can have pedagogical implications for EAP course designers as well as academic writing instructors and students.

The AGORA High-resolution Galaxy Simulations Comparison Project

OpenAIRE

Kim Ji-hoon; Abel Tom; Agertz Oscar; Bryan Greg L.; Ceverino Daniel; Christensen Charlotte; Conroy Charlie; Dekel Avishai; Gnedin Nickolay Y.; Goldbaum Nathan J.; Guedes Javiera; Hahn Oliver; Hobbs Alexander; Hopkins Philip F.; Hummels Cameron B.

2014-01-01

The Astrophysical Journal Supplement Series 210.1 (2014): 14 reproduced by permission of the AAS We introduce the Assembling Galaxies Of Resolved Anatomy (AGORA) project, a comprehensive numerical study of well-resolved galaxies within the ΛCDM cosmology. Cosmological hydrodynamic simulations with force resolutions of ∼100 proper pc or better will be run with a variety of code platforms to follow the hierarchical growth, star formation history, morphological transformation, and the cycle o...

Modelling and Simulation of National Electronic Product Code Network Demonstrator Project

Science.gov (United States)

Mo, John P. T.

The National Electronic Product Code (EPC) Network Demonstrator Project (NDP) was the first large scale consumer goods track and trace investigation in the world using full EPC protocol system for applying RFID technology in supply chains. The NDP demonstrated the methods of sharing information securely using EPC Network, providing authentication to interacting parties, and enhancing the ability to track and trace movement of goods within the entire supply chain involving transactions among multiple enterprise. Due to project constraints, the actual run of the NDP was 3 months only and was unable to consolidate with quantitative results. This paper discusses the modelling and simulation of activities in the NDP in a discrete event simulation environment and provides an estimation of the potential benefits that can be derived from the NDP if it was continued for one whole year.

Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project

International Nuclear Information System (INIS)

Campolina, Daniel de Almeida Magalhaes

2009-01-01

In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)

Instructional Coding System for Mathematics Program of Studies. MET, A Title IV-C Project.

Science.gov (United States)

Fairfax County Public Schools, VA. Dept. of Instructional Services.

This document is part of the Management for Effective Teaching (MET) support kit, a pilot project designed by the Fairfax County (Virginia) Public Schools to assist elementary school teachers in planning, managing, and implementing the county's Program of Studies (POS). This document provides an alpha-numeric coding system to be used in…

Development and application of methods and computer codes of fuel management and nuclear design of reload cycles in PWR

International Nuclear Information System (INIS)

Ahnert, C.; Aragones, J.M.; Corella, M.R.; Esteban, A.; Martinez-Val, J.M.; Minguez, E.; Perlado, J.M.; Pena, J.; Matias, E. de; Llorente, A.; Navascues, J.; Serrano, J.

1976-01-01

Description of methods and computer codes for Fuel Management and Nuclear Design of Reload Cycles in PWR, developed at JEN by adaptation of previous codes (LEOPARD, NUTRIX, CITATION, FUELCOST) and implementation of original codes (TEMP, SOTHIS, CICLON, NUDO, MELON, ROLLO, LIBRA, PENELOPE) and their application to the project of Management and Design of Reload Cycles of a 510 Mwt PWR, including comparison with results of experimental operation and other calculations for validation of methods. (author) [es

Improved lossless intra coding for H.264/MPEG-4 AVC.

Science.gov (United States)

Lee, Yung-Lyul; Han, Ki-Hun; Sullivan, Gary J

2006-09-01

A new lossless intra coding method based on sample-by-sample differential pulse code modulation (DPCM) is presented as an enhancement of the H.264/MPEG-4 AVC standard. The H.264/AVC design includes a multidirectional spatial prediction method to reduce spatial redundancy by using neighboring samples as a prediction for the samples in a block of data to be encoded. In the new lossless intra coding method, the spatial prediction is performed based on samplewise DPCM instead of in the block-based manner used in the current H.264/AVC standard, while the block structure is retained for the residual difference entropy coding process. We show that the new method, based on samplewise DPCM, does not have a major complexity penalty, despite its apparent pipeline dependencies. Experiments show that the new lossless intra coding method reduces the bit rate by approximately 12% in comparison with the lossless intra coding method previously included in the H.264/AVC standard. As a result, the new method is currently being adopted into the H.264/AVC standard in a new enhancement project.

The research on multi-projection correction based on color coding grid array

Science.gov (United States)

Yang, Fan; Han, Cheng; Bai, Baoxing; Zhang, Chao; Zhao, Yunxiu

2017-10-01

There are many disadvantages such as lower timeliness, greater manual intervention in multi-channel projection system, in order to solve the above problems, this paper proposes a multi-projector correction technology based on color coding grid array. Firstly, a color structured light stripe is generated by using the De Bruijn sequences, then meshing the feature information of the color structured light stripe image. We put the meshing colored grid intersection as the center of the circle, and build a white solid circle as the feature sample set of projected images. It makes the constructed feature sample set not only has the perceptual localization, but also has good noise immunity. Secondly, we establish the subpixel geometric mapping relationship between the projection screen and the individual projectors by using the structure of light encoding and decoding based on the color array, and the geometrical mapping relation is used to solve the homography matrix of each projector. Lastly the brightness inconsistency of the multi-channel projection overlap area is seriously interfered, it leads to the corrected image doesn't fit well with the observer's visual needs, and we obtain the projection display image of visual consistency by using the luminance fusion correction algorithm. The experimental results show that this method not only effectively solved the problem of distortion of multi-projection screen and the issue of luminance interference in overlapping region, but also improved the calibration efficient of multi-channel projective system and reduced the maintenance cost of intelligent multi-projection system.

Software requirements, design, and verification and validation for the FEHM application - a finite-element heat- and mass-transfer code

International Nuclear Information System (INIS)

Dash, Z.V.; Robinson, B.A.; Zyvoloski, G.A.

1997-07-01

The requirements, design, and verification and validation of the software used in the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multicomponent flow in porous media, are described. The test of the DOE Code Comparison Project, Problem Five, Case A, which verifies that FEHM has correctly implemented heat and mass transfer and phase partitioning, is also covered

APPC - A new standardised coding system for trans-organisational PACS retrieval

International Nuclear Information System (INIS)

Fruehwald, F.; Lindner, A.; Mostbeck, G.; Hruby, W.; Fruehwald-Pallamar, J.

2010-01-01

As part of a general strategy to integrate the health care enterprise, Austria plans to connect the Picture Archiving and Communication Systems (PACS) of all radiological institutions into a nationwide network. To facilitate the search for relevant correlative imaging data in the PACS of different organisations, a coding system was compiled for all radiological procedures and necessary anatomical details. This code, called the Austrian PACS Procedure Code (APPC), was granted the status of a standard under HL7. Examples are provided of effective coding and filtering when searching for relevant imaging material using the APPC, as well as the planned process for future adjustments of the APPC. The implementation and how the APPC will fit into the future electronic environment, which will include an electronic health act for all citizens in Austria, are discussed. A comparison to other nationwide electronic health record projects and coding systems is given. Limitations and possible use in physical storage media are contemplated. (orig.)

The Visual Code Navigator : An Interactive Toolset for Source Code Investigation

NARCIS (Netherlands)

Lommerse, Gerard; Nossin, Freek; Voinea, Lucian; Telea, Alexandru

2005-01-01

We present the Visual Code Navigator, a set of three interrelated visual tools that we developed for exploring large source code software projects from three different perspectives, or views: The syntactic view shows the syntactic constructs in the source code. The symbol view shows the objects a

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

International Nuclear Information System (INIS)

Benuzzi, A.

1986-01-01

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

Best practice in the management of clinical coding services: Insights from a project in the Republic of Ireland, Part 2.

Science.gov (United States)

Reid, Beth A; Ridoutt, Lee; O'Connor, Paul; Murphy, Deirdre

2017-09-01

This is the second of two articles about best practice in the management of coding services. The best practice project was part of a year-long project conducted in the Republic of Ireland to review the quality of the Hospital Inpatient Enquiry data for its use in activity-based funding. The four methods used to address the best practice aspect of the project were described in detail in Part 1. The results included in this article are those relating to the coding manager's background, preparation and style, clinical coder (CC) workforce adequacy, the CC workforce structure and career pathway, and the physical and psychological work environment for the clinical coding service. Examples of best practice were found in the study hospitals but there were also areas for improvement. Coding managers would benefit from greater support in the form of increased opportunities for management training and a better method for calculating CC workforce numbers. A career pathway is needed for CCs to progress from entry to expert CC, mentor, manager and quality controller. Most hospitals could benefit from investment in infrastructure that places CCs in a physical environment that tells them they are an important part of the hospital and their work is valued.

Modelling of LOCA Tests with the BISON Fuel Performance Code

Energy Technology Data Exchange (ETDEWEB)

Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

2016-05-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

Evaluation of Computational Fluids Dynamics (CFD) code Open FOAM in the study of the pressurized thermal stress of PWR reactors. Comparison with the commercial code Ansys-CFX

International Nuclear Information System (INIS)

Martinez, M.; Barrachina, T.; Miro, R.; Verdu Martin, G.; Chiva, S.

2012-01-01

In this work is proposed to evaluate the potential of the OpenFOAM code for the simulation of typical fluid flows in reactors PWR, in particular for the study of pressurized thermal stress. Test T1-1 has been simulated , within the OECD ROSA project, with the objective of evaluating the performance of the code OpenFOAM and models of turbulence that has implemented to capture the effect of the thrust forces in the case study.

Comparison of the thermal neutron scattering treatment in MCNP6 and GEANT4 codes

Science.gov (United States)

Tran, H. N.; Marchix, A.; Letourneau, A.; Darpentigny, J.; Menelle, A.; Ott, F.; Schwindling, J.; Chauvin, N.

2018-06-01

To ensure the reliability of simulation tools, verification and comparison should be made regularly. This paper describes the work performed in order to compare the neutron transport treatment in MCNP6.1 and GEANT4-10.3 in the thermal energy range. This work focuses on the thermal neutron scattering processes for several potential materials which would be involved in the neutron source designs of Compact Accelerator-based Neutrons Sources (CANS), such as beryllium metal, beryllium oxide, polyethylene, graphite, para-hydrogen, light water, heavy water, aluminium and iron. Both thermal scattering law and free gas model, coming from the evaluated data library ENDF/B-VII, were considered. It was observed that the GEANT4.10.03-patch2 version was not able to account properly the coherent elastic process occurring in crystal lattice. This bug is treated in this work and it should be included in the next release of the code. Cross section sampling and integral tests have been performed for both simulation codes showing a fair agreement between the two codes for most of the materials except for iron and aluminium.

Remote-Handled Low-Level Waste Disposal Project Code of Record

Energy Technology Data Exchange (ETDEWEB)

S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

2012-06-01

The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

Remote-Handled Low-Level Waste Disposal Project Code of Record

Energy Technology Data Exchange (ETDEWEB)

S.L. Austad, P.E.; L.E. Guillen, P.E.; C. W. McKnight, P.E.; D. S. Ferguson, P.E.

2014-06-01

The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

Remote-Handled Low-Level Waste Disposal Project Code of Record

Energy Technology Data Exchange (ETDEWEB)

Austad, S. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Guillen, L. E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); McKnight, C. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ferguson, D. S. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-04-01

The Remote-Handled Low-Level Waste (LLW) Disposal Project addresses an anticipated shortfall in remote-handled LLW disposal capability following cessation of operations at the existing facility, which will continue until it is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). Development of a new onsite disposal facility will provide necessary remote-handled LLW disposal capability and will ensure continuity of operations that generate remote-handled LLW. This report documents the Code of Record for design of a new LLW disposal capability. The report is owned by the Design Authority, who can authorize revisions and exceptions. This report will be retained for the lifetime of the facility.

OC5 Project Phase II: Validation of Global Loads of the DeepCwind Floating Semisubmersible Wind Turbine

DEFF Research Database (Denmark)

Robertson, Amy N.; Wendt, Fabian; Jonkman, Jason M.

2017-01-01

This paper summarizes the findings from Phase II of the Offshore Code Comparison, Collaboration, Continued, with Correlation project. The project is run under the International Energy Agency Wind Research Task 30, and is focused on validating the tools used for modeling offshore wind systems thro...

Comparison calculations of WWER-1000 fuel assemblies by using the MCNP 4.2 a KASSETA codes

International Nuclear Information System (INIS)

Trgina, M.

1993-12-01

The power multiplication and distribution factors are compared for various geometries and material configurations of WWER-1000 fuel assemblies. The calculations were performed in 2 ways: (i) using nuclear data, employing older and current data collections, and (ii) using the author's own model based on the KASSETA code. The comparison code MCNP 4.2 is described, intended for computerized simulation of the transport of neutrons, photons and electrons. This code uses its own cross section library. The methodology is outlined and a specification of the Monte Carlo method employed is given. The use of the refined data library gave rise to appreciable deviations of the multiplication factors in all variants. The use of the older data library led to identical criticality results for the variant with water holes. For inserted absorbers the discrepancies in criticality and in power distribution data are appreciable. The marked disagreement between the results of application of the MCNP 4.2 and KASSETA codes for the variants with inserted control elements is indicative of inappropriateness of the approximation procedure in the latter code. (J.B.). 2 tabs., 11 figs., 11 refs

LiveCode mobile development

CERN Document Server

Lavieri, Edward D

2013-01-01

A practical guide written in a tutorial-style, ""LiveCode Mobile Development Hotshot"" walks you step-by-step through 10 individual projects. Every project is divided into sub tasks to make learning more organized and easy to follow along with explanations, diagrams, screenshots, and downloadable material.This book is great for anyone who wants to develop mobile applications using LiveCode. You should be familiar with LiveCode and have access to a smartphone. You are not expected to know how to create graphics or audio clips.

A Comparison of Creativity in Project Groups in Science and Engineering Education in Denmark and China

DEFF Research Database (Denmark)

Zhou, Chunfang; Valero, Paola

2015-01-01

Different pedagogical strategies influence the development of creativity in project groups in science and engineering education. This study is a comparison between two cases: Problem-Based Learning (PBL) in Denmark and Project-Organized Learning (POL) in China.......Different pedagogical strategies influence the development of creativity in project groups in science and engineering education. This study is a comparison between two cases: Problem-Based Learning (PBL) in Denmark and Project-Organized Learning (POL) in China....

Comparison of the THYC and FLICA-3M codes by the pseudo-cubic thin-plate method

International Nuclear Information System (INIS)

Banner, D.; Crecy, F. de.

1993-06-01

The pseudo cubic Spline method (PCSM) is a statistical tool developed by the CEA. It is designed to analyse experimental points and in particular thermalhydraulic data. Predictors of the occurrence of critical heat flux are obtained by using Spline functions. In this paper, predictors have been computed from the same CHF databases by using two different flow analyses to derive local thermal-hydraulic variables at the CHF location. In fact, CEA's FLICA-3M represents rod bundles by interconnected subchannels whereas EDF's THYC code uses a porous 3D approach. In a first step, the PCSM is briefly presented as well as the two codes studied here. Then, the comparison methodology is explained in order to prove that advanced analysis of thermalhydraulic codes can be achieved with the PCSM. (authors). 6 figs., 2 tabs., 5 refs

Design project of fast subcritical system 'Mala Lasta'

International Nuclear Information System (INIS)

Milosevic, M.; Stefanovic, D.; Popovic, D.; Pesic, M.; Zavaljevski, N.; Nikolic, D.; Arsenovic, M.

1988-10-01

This report contains two parts. Part one covers the objective and fundamental elements for the choice of fast subcritical system 'Mala Lasta', review of the existing fast subcritical assemblies, and a description of the available domestic computer codes applied for calculating neutron reactor parameters. Comparison of results obtained by these codes for a number of existing subcritical assemblies was used for the choice of the design project described in part two of this report. It contains detailed description of the operating parameters of the chosen subcritical system based on the obtained calculated parameters

Safeguarding public values in gas infrastructure expansion. A comparison of two investment projects

International Nuclear Information System (INIS)

De Joode, J.

2007-02-01

The realisation of new gas infrastructure projects affects overall gas market performance with respect to the public values of affordability and security of supply. However, the actual contribution of a gas infrastructure expansion project to system affordability and security of supply depends upon the institutional design of the market (legislation, regulatory codes and arrangements, market rules, etc.). In this paper we link the institutional design applicable to two specific gas infrastructure projects with the safeguarding of the aforementioned public values. We conclude that path dependencies can cause large differences in the contribution of the projects to the safeguarding of public values

Alternative Fuels Data Center: Codes and Standards Resources

Science.gov (United States)

resources linked below help project developers and code officials prepare and review code-compliant projects , storage, and infrastructure. The following charts show the SDOs responsible for these alternative fuel codes and standards. Biodiesel Vehicle and Infrastructure Codes and Standards Chart Electric Vehicle and

Pre-Test Analysis of the MEGAPIE Spallation Source Target Cooling Loop Using the TRAC/AAA Code

International Nuclear Information System (INIS)

Bubelis, Evaldas; Coddington, Paul; Leung, Waihung

2006-01-01

A pilot project is being undertaken at the Paul Scherrer Institute in Switzerland to test the feasibility of installing a Lead-Bismuth Eutectic (LBE) spallation target in the SINQ facility. Efforts are coordinated under the MEGAPIE project, the main objectives of which are to design, build, operate and decommission a 1 MW spallation neutron source. The technology and experience of building and operating a high power spallation target are of general interest in the design of an Accelerator Driven System (ADS) and in this context MEGAPIE is one of the key experiments. The target cooling is one of the important aspects of the target system design that needs to be studied in detail. Calculations were performed previously using the RELAP5/Mod 3.2.2 and ATHLET codes, but in order to verify the previous code results and to provide another capability to model LBE systems, a similar study of the MEGAPIE target cooling system has been conducted with the TRAC/AAA code. In this paper a comparison is presented for the steady-state results obtained using the above codes. Analysis of transients, such as unregulated cooling of the target, loss of heat sink, the main electro-magnetic pump trip of the LBE loop and unprotected proton beam trip, were studied with TRAC/AAA and compared to those obtained earlier using RELAP5/Mod 3.2.2. This work extends the existing validation data-base of TRAC/AAA to heavy liquid metal systems and comprises the first part of the TRAC/AAA code validation study for LBE systems based on data from the MEGAPIE test facility and corresponding inter-code comparisons. (authors)

Benchmark Problems of the Geothermal Technologies Office Code Comparison Study

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-02

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

High-dynamic range compressive spectral imaging by grayscale coded aperture adaptive filtering

Directory of Open Access Journals (Sweden)

Nelson Eduardo Diaz

2015-09-01

Full Text Available The coded aperture snapshot spectral imaging system (CASSI is an imaging architecture which senses the three dimensional informa-tion of a scene with two dimensional (2D focal plane array (FPA coded projection measurements. A reconstruction algorithm takes advantage of the compressive measurements sparsity to recover the underlying 3D data cube. Traditionally, CASSI uses block-un-block coded apertures (BCA to spatially modulate the light. In CASSI the quality of the reconstructed images depends on the design of these coded apertures and the FPA dynamic range. This work presents a new CASSI architecture based on grayscaled coded apertu-res (GCA which reduce the FPA saturation and increase the dynamic range of the reconstructed images. The set of GCA is calculated in a real-time adaptive manner exploiting the information from the FPA compressive measurements. Extensive simulations show the attained improvement in the quality of the reconstructed images when GCA are employed.  In addition, a comparison between traditional coded apertures and GCA is realized with respect to noise tolerance.

A Comparison Among Plastic Deformation Capacities of RC Members According to International Codes

International Nuclear Information System (INIS)

Tripepi, C.; Failla, G.; Santini, A.; Nucera, F.

2008-01-01

The aim is to compare plastic deformation capacities of flexure-controlled reinforced concrete members, as predicted by the Italian Seismic Code, Eurocode 8 and FEMA356. For completeness, recent studies in the literature are also referred to. The comparison is pursued in context with a nonlinear static analysis run on 2D frame structures. This allows to assess whether and to which extent plastic deformation capacities may be affected by variations in those quantities, such as shear span and/or axial load, depending on which plastic deformation capacities are generally given

Multichannel Filtered-X Error Coded Affine Projection-Like Algorithm with Evolving Order

Directory of Open Access Journals (Sweden)

J. G. Avalos

2017-01-01

Full Text Available Affine projection (AP algorithms are commonly used to implement active noise control (ANC systems because they provide fast convergence. However, their high computational complexity can restrict their use in certain practical applications. The Error Coded Affine Projection-Like (ECAP-L algorithm has been proposed to reduce the computational burden while maintaining the speed of AP, but no version of this algorithm has been derived for active noise control, for which the adaptive structures are very different from those of other configurations. In this paper, we introduce a version of the ECAP-L for single-channel and multichannel ANC systems. The proposed algorithm is implemented using the conventional filtered-x scheme, which incurs a lower computational cost than the modified filtered-x structure, especially for multichannel systems. Furthermore, we present an evolutionary method that dynamically decreases the projection order in order to reduce the dimensions of the matrix used in the algorithm’s computations. Experimental results demonstrate that the proposed algorithm yields a convergence speed and a final residual error similar to those of AP algorithms. Moreover, it achieves meaningful computational savings, leading to simpler hardware implementation of real-time ANC applications.

Integrated Validation System for a Thermal-hydraulic System Code, TASS/SMR-S

Energy Technology Data Exchange (ETDEWEB)

Kim, Hee-Kyung; Kim, Hyungjun; Kim, Soo Hyoung; Hwang, Young-Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Hyeon-Soo [Chungnam National University, Daejeon (Korea, Republic of)

2015-10-15

Development including enhancement and modification of thermal-hydraulic system computer code is indispensable to a new reactor, SMART. Usually, a thermal-hydraulic system code validation is achieved by a comparison with the results of corresponding physical effect tests. In the reactor safety field, a similar concept, referred to as separate effect tests has been used for a long time. But there are so many test data for comparison because a lot of separate effect tests and integral effect tests are required for a code validation. It is not easy to a code developer to validate a computer code whenever a code modification is occurred. IVS produces graphs which shown the comparison the code calculation results with the corresponding test results automatically. IVS was developed for a validation of TASS/SMR-S code. The code validation could be achieved by a comparison code calculation results with corresponding test results. This comparison was represented as a graph for convenience. IVS is useful before release a new code version. The code developer can validate code result easily using IVS. Even during code development, IVS could be used for validation of code modification. The code developer could gain a confidence about his code modification easily and fast and could be free from tedious and long validation work. The popular software introduced in IVS supplies better usability and portability.

Verification of the depletion capabilities of the MCNPX code on a LWR MOX fuel assembly

International Nuclear Information System (INIS)

Cerba, S.; Hrncir, M.; Necas, V.

2012-01-01

The study deals with the verification of the depletion capabilities of the MCNPX code, which is a linked Monte-Carlo depletion code. For such a purpose the IV-B phase of the OECD NEA Burnup credit benchmark has been chosen. The mentioned benchmark is a code to code comparison of the multiplication coefficient k eff and the isotopic composition of a LWR MOX fuel assembly at three given burnup levels and after five years of cooling. The benchmark consists of 6 cases, 2 different Pu vectors and 3 geometry models, however in this study only the fuel assembly calculations with two Pu vectors were performed. The aim of this study was to compare the obtained result with data from the participants of the OECD NEA Burnup Credit project and confirm the burnup capability of the MCNPX code. (Authors)

Comparison of Engine Cycle Codes for Rocket-Based Combined Cycle Engines

Science.gov (United States)

Waltrup, Paul J.; Auslender, Aaron H.; Bradford, John E.; Carreiro, Louis R.; Gettinger, Christopher; Komar, D. R.; McDonald, J.; Snyder, Christopher A.

2002-01-01

This paper summarizes the results from a one day workshop on Rocket-Based Combined Cycle (RBCC) Engine Cycle Codes held in Monterey CA in November of 2000 at the 2000 JANNAF JPM with the authors as primary participants. The objectives of the workshop were to discuss and compare the merits of existing Rocket-Based Combined Cycle (RBCC) engine cycle codes being used by government and industry to predict RBCC engine performance and interpret experimental results. These merits included physical and chemical modeling, accuracy and user friendliness. The ultimate purpose of the workshop was to identify the best codes for analyzing RBCC engines and to document any potential shortcomings, not to demonstrate the merits or deficiencies of any particular engine design. Five cases representative of the operating regimes of typical RBCC engines were used as the basis of these comparisons. These included Mach 0 sea level static and Mach 1.0 and Mach 2.5 Air-Augmented-Rocket (AAR), Mach 4 subsonic combustion ramjet or dual-mode scramjet, and Mach 8 scramjet operating modes. Specification of a generic RBCC engine geometry and concomitant component operating efficiencies, bypass ratios, fuel/oxidizer/air equivalence ratios and flight dynamic pressures were provided. The engine included an air inlet, isolator duct, axial rocket motor/injector, axial wall fuel injectors, diverging combustor, and exit nozzle. Gaseous hydrogen was used as the fuel with the rocket portion of the system using a gaseous H2/O2 propellant system to avoid cryogenic issues. The results of the workshop, even after post-workshop adjudication of differences, were surprising. They showed that the codes predicted essentially the same performance at the Mach 0 and I conditions, but progressively diverged from a common value (for example, for fuel specific impulse, Isp) as the flight Mach number increased, with the largest differences at Mach 8. The example cases and results are compared and discussed in this paper.

Cross-verification of the GENE and XGC codes in preparation for their coupling

Science.gov (United States)

Jenko, Frank; Merlo, Gabriele; Bhattacharjee, Amitava; Chang, Cs; Dominski, Julien; Ku, Seunghoe; Parker, Scott; Lanti, Emmanuel

2017-10-01

A high-fidelity Whole Device Model (WDM) of a magnetically confined plasma is a crucial tool for planning and optimizing the design of future fusion reactors, including ITER. Aiming at building such a tool, in the framework of the Exascale Computing Project (ECP) the two existing gyrokinetic codes GENE (Eulerian delta-f) and XGC (PIC full-f) will be coupled, thus enabling to carry out first principle kinetic WDM simulations. In preparation for this ultimate goal, a benchmark between the two codes is carried out looking at ITG modes in the adiabatic electron limit. This verification exercise is also joined by the global Lagrangian PIC code ORB5. Linear and nonlinear comparisons have been carried out, neglecting for simplicity collisions and sources. A very good agreement is recovered on frequency, growth rate and mode structure of linear modes. A similarly excellent agreement is also observed comparing the evolution of the heat flux and of the background temperature profile during nonlinear simulations. Work supported by the US DOE under the Exascale Computing Project (17-SC-20-SC).

Comparison of computer code calculations with experimental results obtained in the NSPP series of experiments

International Nuclear Information System (INIS)

Tobias, M.L.

1987-01-01

Experiments were done on several aerosols in air atmospheres at varying temperatures and humidity conditions of interest in forming a data base for testing aerosol behavior models used as part of the process of evaluating the ''source term'' in light water reactor accidents. This paper deals with the problems of predicting the observed experimental data for suspended aerosol concentration with aerosol calculational codes. Comparisons of measured versus predicted data are provided

Development of an aeroelastic code based on three-dimensional viscous–inviscid method for wind turbine computations

DEFF Research Database (Denmark)

Sessarego, Matias; Ramos García, Néstor; Sørensen, Jens Nørkær

2017-01-01

Aerodynamic and structural dynamic performance analysis of modern wind turbines are routinely estimated in the wind energy field using computational tools known as aeroelastic codes. Most aeroelastic codes use the blade element momentum (BEM) technique to model the rotor aerodynamics and a modal......, multi-body or the finite-element approach to model the turbine structural dynamics. The present work describes the development of a novel aeroelastic code that combines a three-dimensional viscous–inviscid interactive method, method for interactive rotor aerodynamic simulations (MIRAS...... Code Comparison Collaboration Project. Simulation tests consist of steady wind inflow conditions with different combinations of yaw error, wind shear, tower shadow and turbine-elastic modeling. Turbulent inflow created by using a Mann box is also considered. MIRAS-FLEX results, such as blade tip...

UEDGE code comparisons with DIII-D bolometer data

Energy Technology Data Exchange (ETDEWEB)

Daniel, J.M.

1994-12-01

This paper describes the work done to develop a bolometer post processor that converts volumetric radiated power values taken from a UEDGE solution, to a line integrated radiated power along chords of the bolometers in the DIII-D tokamak. The UEDGE code calculates plasma physics quantities, such as plasma density, radiated power, or electron temperature, and compares them to actual diagnostic measurements taken from the scrape off layer (SOL) and divertor regions of the DIII-D tokamak. Bolometers are devices measuring radiated power within the tokamak. The bolometer interceptors are made up of two complete arrays, an upper array with a vertical view and a lower array with a horizontal view, so that a two dimensional profile of the radiated power may be obtained. The bolometer post processor stores line integrated values taken from UEDGE solutions into a file in tabular format. Experimental data is then put into tabular form and placed in another file. Comparisons can be made between the UEDGE solutions and actual bolometer data. Analysis has been done to determine the accuracy of the plasma physics involved in producing UEDGE simulations.

Intercomparison and validation of computer codes for thermalhydraulic safety analysis of heavy water reactors

International Nuclear Information System (INIS)

2004-08-01

Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and co-operative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled: Intercomparison and validation of computer codes for thermalhydraulics safety analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. The RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for this validation project. This report provides a comparison of the results obtained from six participating countries, utilizing four different computer codes. General conclusions are reached and recommendations made

Comparison of MACCS users calculations for the international comparison exercise on probabilistic accident consequence assessment code, October 1989--June 1993

International Nuclear Information System (INIS)

Neymotin, L.

1994-04-01

Over the past several years, the OECD/NEA and CEC sponsored an international program intercomparing a group of six probabilistic consequence assessment (PCA) codes designed to simulate health and economic consequences of radioactive releases into atmosphere of radioactive materials following severe accidents at nuclear power plants (NPPs): ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this effort, two separate groups performed similar calculations using the MACCS and COSYMA codes. Results produced in the MACCS Users Group (Greece, Italy, Spain, and USA) calculations and their comparison are contained in the present report. Version 1.5.11.1 of the MACCS code was used for the calculations. Good agreement between the results produced in the four participating calculations has been reached, with the exception of the results related to the ingestion pathway dose predictions. The main reason for the scatter in those particular results is attributed to the lack of a straightforward implementation of the specifications for agricultural production and counter-measures criteria provided for the exercise. A significantly smaller scatter in predictions of other consequences was successfully explained by differences in meteorological files and weather sampling, grids, rain distance intervals, dispersion model options, and population distributions

Parameter calculation tool for the application of radiological dose projection codes

International Nuclear Information System (INIS)

Galindo G, I. F.; Vergara del C, J. A.; Galvan A, S. J.; Tijerina S, F.

2016-09-01

The use of specialized codes to estimate the radiation dose projection to an emergency postulated event at a nuclear power plant requires that certain plant data be available according to the event being simulated. The calculation of the possible radiological release is the critical activity to carry out the emergency actions. However, not all of the plant data required are obtained directly from the plant but need to be calculated. In this paper we present a computational tool that calculates the plant data required to use the radiological dose estimation codes. The tool provides the required information when there is a gas emergency venting event in the primary containment atmosphere, whether well or dry well and also calculates the time in which the spent fuel pool would be discovered in the event of a leak of water on some of the walls or floor of the pool. The tool developed has mathematical models for the processes involved such as: compressible flow in pipes considering area change and for constant area, taking into account the effects of friction and for the case of the spent fuel pool hydraulic models to calculate the time in which a container is emptied. The models implemented in the tool are validated with data from the literature for simulated cases. The results with the tool are very similar to those of reference. This tool will also be very supportive so that in postulated emergency cases can use the radiological dose estimation codes to adequately and efficiently determine the actions to be taken in a way that affects as little as possible. (Author)

Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

International Nuclear Information System (INIS)

Bujan, A.; Adamik, V.; Misak, J.

1986-01-01

A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

Simulation of guided-wave ultrasound propagation in composite laminates: Benchmark comparisons of numerical codes and experiment.

Science.gov (United States)

Leckey, Cara A C; Wheeler, Kevin R; Hafiychuk, Vasyl N; Hafiychuk, Halyna; Timuçin, Doğan A

2018-03-01

Ultrasonic wave methods constitute the leading physical mechanism for nondestructive evaluation (NDE) and structural health monitoring (SHM) of solid composite materials, such as carbon fiber reinforced polymer (CFRP) laminates. Computational models of ultrasonic wave excitation, propagation, and scattering in CFRP composites can be extremely valuable in designing practicable NDE and SHM hardware, software, and methodologies that accomplish the desired accuracy, reliability, efficiency, and coverage. The development and application of ultrasonic simulation approaches for composite materials is an active area of research in the field of NDE. This paper presents comparisons of guided wave simulations for CFRP composites implemented using four different simulation codes: the commercial finite element modeling (FEM) packages ABAQUS, ANSYS, and COMSOL, and a custom code executing the Elastodynamic Finite Integration Technique (EFIT). Benchmark comparisons are made between the simulation tools and both experimental laser Doppler vibrometry data and theoretical dispersion curves. A pristine and a delamination type case (Teflon insert in the experimental specimen) is studied. A summary is given of the accuracy of simulation results and the respective computational performance of the four different simulation tools. Published by Elsevier B.V.

Use of Lump Parameter Codes at SNSA

International Nuclear Information System (INIS)

Muehleisen, A.

2006-01-01

The lump parameter codes are due to the specifics of Slovenian regulation used only in a very limited scope by the SNSA itself. The law requires that most of the analysis needed for regulatory decision making have to be performed by technical support organisations (TSOs). The use of lump parameter codes is therefore limited to the amount needed to maintain necessary technical competence and to support, to a degree, the reasoning for raising new issues and methodologies. SNSA has available its own NPP MELCOR model and uses for its own purposes NPP Krsko RELAP model. RELAP model is also part of the SNSA NPA analyser. Here presented recent uses at SNSA include use of NPA in support of a project, aimed at estimating maturity and uses of CFD codes for regulatory purposes, transition from MELCOR 1.8.3 to 1.8.5 model and its validation, developing MELCOR PAR model and use of NPA for training purposes. NPA use in support of investigation of CFD usability has been in performing lump parameter code calculation against which the CFD results could be compared. The case of SI injection and the following boron distribution in the reactor vessel has been used for this purpose. The comparison showed that for the particular case there is no urgent need for CFD code calculations, nevertheless the project clearly demonstrated wealth of additional information that can be gained by the use of CFD code. As far as MELCOR model is concerned, only transition of the model to the newer code version has been performed and PAR input prepared and tested. Even though there is a feeling at SNSA that some preliminary analysis with it (such as analysis of typical accidents with PARs present and analysis in support of wet cavity modification) would be useful as a support for decision making as well as for simple training purposes we have not been able to perform them due to other priorities and lack of human resources. SNSA is additionally tasked with support to TSOs in their efforts to maintain and

A class of Sudan-decodable codes

DEFF Research Database (Denmark)

Nielsen, Rasmus Refslund

2000-01-01

In this article, Sudan's algorithm is modified into an efficient method to list-decode a class of codes which can be seen as a generalization of Reed-Solomon codes. The algorithm is specialized into a very efficient method for unique decoding. The code construction can be generalized based...... on algebraic-geometry codes and the decoding algorithms are generalized accordingly. Comparisons with Reed-Solomon and Hermitian codes are made....

Validation of containment thermal hydraulic computer codes for VVER reactor

Energy Technology Data Exchange (ETDEWEB)

Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

2005-07-01

Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

First step of the project for implementation of two non-symmetric cooling loops modeled by the ALMOD3 code

International Nuclear Information System (INIS)

Dominguez, L.; Camargo, C.T.M.

1984-09-01

The first step of the project for implementation of two non-symmetric cooling loops modeled by the ALMOD3 computer code is presented. This step consists of the introduction of a simplified model for simulating the steam generator. This model is the GEVAP computer code, integrant part of LOOP code, which simulates the primary coolant circuit of PWR nuclear power plants during transients. The ALMOD3 computer code has a model for the steam generator, called UTSG, which is very detailed. This model has spatial dependence, correlations for 2-phase flow, distinguished correlations for different heat transfer process. The GEVAP model has thermal equilibrium between phases (gaseous and liquid homogeneous mixture), no spatial dependence and uses only one generalized correlation to treat several heat transfer processes. (Author) [pt

Update on comparison of the particle production using Mars simulation code

CERN Document Server

Prior, G; Kirk, H G; Souchlas, N; Ding, X

2011-01-01

In the International Design Study for the Neutrino Factory (IDS-NF), a 5-15 GeV (kinetic energy) proton beam impinges a Hg jet target, in order to produce pions that will decay into muons. The muons are captured and transformed into a beam, then passed to the downstream acceleration system. The target sits in a solenoid eld tapering from 20 T down to below 2 T, over several meters, permitting an optimized capture of the pions that will produce useful muons for the machine. The target and pion capture systems have been simulated using MARS. This paper presents an updated comparison of the particles production using the MARS code versions m1507 and m1510 on different machines located at the European Organization for Nuclear Research (CERN) and Brookhaven National Laboratory (BNL).

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

International Nuclear Information System (INIS)

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

1992-03-01

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

Validation of one-dimensional module of MARS 2.1 computer code by comparison with the RELAP5/MOD3.3 developmental assessment results

International Nuclear Information System (INIS)

Lee, Y. J.; Bae, S. W.; Chung, B. D.

2003-02-01

This report records the results of the code validation for the one-dimensional module of the MARS 2.1 thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 code development assessment problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS 2.1 code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The results suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

NEACRP comparison of source term codes for the radiation protection assessment of transportation packages

International Nuclear Information System (INIS)

Broadhead, B.L.; Locke, H.F.; Avery, A.F.

1994-01-01

The results for Problems 5 and 6 of the NEACRP code comparison as submitted by six participating countries are presented in summary. These problems concentrate on the prediction of the neutron and gamma-ray sources arising in fuel after a specified irradiation, the fuel being uranium oxide for problem 5 and a mixture of uranium and plutonium oxides for problem 6. In both problems the predicted neutron sources are in good agreement for all participants. For gamma rays, however, there are differences, largely due to the omission of bremsstrahlung in some calculations

Project planning and project management of Baseball II-T

International Nuclear Information System (INIS)

Kozman, T.A.; Chargin, A.K.

1975-01-01

The details of the project planning and project management work done on the Baseball II-T experiment are reviewed. The LLL Baseball program is a plasma confinement experiment accomplished with a superconducting magnet in the shape of a baseball seam. Both project planning and project management made use of the Critical Path Management (CPM) computer code. The computer code, input, and results from the project planning and project management runs, and the cost and effectiveness of this method of systems planning are discussed

What to do with a Dead Research Code

Science.gov (United States)

Nemiroff, Robert J.

2016-01-01

The project has ended -- should all of the computer codes that enabled the project be deleted? No. Like research papers, research codes typically carry valuable information past project end dates. Several possible end states to the life of research codes are reviewed. Historically, codes are typically left dormant on an increasingly obscure local disk directory until forgotten. These codes will likely become any or all of: lost, impossible to compile and run, difficult to decipher, and likely deleted when the code's proprietor moves on or dies. It is argued here, though, that it would be better for both code authors and astronomy generally if project codes were archived after use in some way. Archiving is advantageous for code authors because archived codes might increase the author's ADS citable publications, while astronomy as a science gains transparency and reproducibility. Paper-specific codes should be included in the publication of the journal papers they support, just like figures and tables. General codes that support multiple papers, possibly written by multiple authors, including their supporting websites, should be registered with a code registry such as the Astrophysics Source Code Library (ASCL). Codes developed on GitHub can be archived with a third party service such as, currently, BackHub. An important code version might be uploaded to a web archiving service like, currently, Zenodo or Figshare, so that this version receives a Digital Object Identifier (DOI), enabling it to found at a stable address into the future. Similar archiving services that are not DOI-dependent include perma.cc and the Internet Archive Wayback Machine at archive.org. Perhaps most simply, copies of important codes with lasting value might be kept on a cloud service like, for example, Google Drive, while activating Google's Inactive Account Manager.

BeerDeCoded: the open beer metagenome project.

Science.gov (United States)

Sobel, Jonathan; Henry, Luc; Rotman, Nicolas; Rando, Gianpaolo

2017-01-01

Next generation sequencing has radically changed research in the life sciences, in both academic and corporate laboratories. The potential impact is tremendous, yet a majority of citizens have little or no understanding of the technological and ethical aspects of this widespread adoption. We designed BeerDeCoded as a pretext to discuss the societal issues related to genomic and metagenomic data with fellow citizens, while advancing scientific knowledge of the most popular beverage of all. In the spirit of citizen science, sample collection and DNA extraction were carried out with the participation of non-scientists in the community laboratory of Hackuarium, a not-for-profit organisation that supports unconventional research and promotes the public understanding of science. The dataset presented herein contains the targeted metagenomic profile of 39 bottled beers from 5 countries, based on internal transcribed spacer (ITS) sequencing of fungal species. A preliminary analysis reveals the presence of a large diversity of wild yeast species in commercial brews. With this project, we demonstrate that coupling simple laboratory procedures that can be carried out in a non-professional environment with state-of-the-art sequencing technologies and targeted metagenomic analyses, can lead to the detection and identification of the microbial content in bottled beer.

The aeroelastic code FLEXLAST

Energy Technology Data Exchange (ETDEWEB)

Visser, B. [Stork Product Eng., Amsterdam (Netherlands)

1996-09-01

To support the discussion on aeroelastic codes, a description of the code FLEXLAST was given and experiences within benchmarks and measurement programmes were summarized. The code FLEXLAST has been developed since 1982 at Stork Product Engineering (SPE). Since 1992 FLEXLAST has been used by Dutch industries for wind turbine and rotor design. Based on the comparison with measurements, it can be concluded that the main shortcomings of wind turbine modelling lie in the field of aerodynamics, wind field and wake modelling. (au)

International outage coding system for nuclear power plants. Results of a co-ordinated research project

International Nuclear Information System (INIS)

2004-05-01

The experience obtained in each individual plant constitutes the most relevant source of information for improving its performance. However, experience of the level of the utility, country and worldwide is also extremely valuable, because there are limitations to what can be learned from in-house experience. But learning from the experience of others is admittedly difficult, if the information is not harmonized. Therefore, such systems should be standardized and applicable to all types of reactors satisfying the needs of the broad set of nuclear power plant operators worldwide and allowing experience to be shared internationally. To cope with the considerable amount of information gathered from nuclear power plants worldwide, it is necessary to codify the information facilitating the identification of causes of outages, systems or component failures. Therefore, the IAEA established a sponsored Co-ordinated Research Project (CRP) on the International Outage Coding System to develop a general, internationally applicable system of coding nuclear power plant outages, providing worldwide nuclear utilities with a standardized tool for reporting outage information. This TECDOC summarizes the results of this CRP and provides information for transformation of the historical outage data into the new coding system, taking into consideration the existing systems for coding nuclear power plant events (WANO, IAEA-IRS and IAEA PRIS) but avoiding duplication of efforts to the maximum possible extent

Validation of thermalhydraulic codes

International Nuclear Information System (INIS)

Wilkie, D.

1992-01-01

Thermalhydraulic codes require to be validated against experimental data collected over a wide range of situations if they are to be relied upon. A good example is provided by the nuclear industry where codes are used for safety studies and for determining operating conditions. Errors in the codes could lead to financial penalties, to the incorrect estimation of the consequences of accidents and even to the accidents themselves. Comparison between prediction and experiment is often described qualitatively or in approximate terms, e.g. ''agreement is within 10%''. A quantitative method is preferable, especially when several competing codes are available. The codes can then be ranked in order of merit. Such a method is described. (Author)

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

International Nuclear Information System (INIS)

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

2016-09-01

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

`95 computer system operation project

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Taek; Lee, Hae Cho; Park, Soo Jin; Kim, Hee Kyung; Lee, Ho Yeun; Lee, Sung Kyu; Choi, Mi Kyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-12-01

This report describes overall project works related to the operation of mainframe computers, the management of nuclear computer codes and the project of nuclear computer code conversion. The results of the project are as follows ; 1. The operation and maintenance of the three mainframe computers and other utilities. 2. The management of the nuclear computer codes. 3. The finishing of the computer codes conversion project. 26 tabs., 5 figs., 17 refs. (Author) .new.

'95 computer system operation project

International Nuclear Information System (INIS)

Kim, Young Taek; Lee, Hae Cho; Park, Soo Jin; Kim, Hee Kyung; Lee, Ho Yeun; Lee, Sung Kyu; Choi, Mi Kyung

1995-12-01

This report describes overall project works related to the operation of mainframe computers, the management of nuclear computer codes and the project of nuclear computer code conversion. The results of the project are as follows ; 1. The operation and maintenance of the three mainframe computers and other utilities. 2. The management of the nuclear computer codes. 3. The finishing of the computer codes conversion project. 26 tabs., 5 figs., 17 refs. (Author) .new

Summary of computational support and general documentation for computer code (GENTREE) used in Office of Nuclear Waste Isolation Pilot Salt Site Selection Project

International Nuclear Information System (INIS)

Beatty, J.A.; Younker, J.L.; Rousseau, W.F.; Elayat, H.A.

1983-01-01

A Decision Tree Computer Model was adapted for the purposes of a Pilot Salt Site Selection Project conducted by the Office of Nuclear Waste Isolation (ONWI). A deterministic computer model was developed to structure the site selection problem with submodels reflecting the five major outcome categories (Cost, Safety, Delay, Environment, Community Impact) to be evaluated in the decision process. Time-saving modifications were made in the tree code as part of the effort. In addition, format changes allowed retention of information items which are valuable in directing future research and in isolation of key variabilities in the Site Selection Decision Model. The deterministic code was linked to the modified tree code and the entire program was transferred to the ONWI-VAX computer for future use by the ONWI project

The Coding Process and Its Challenges

Directory of Open Access Journals (Sweden)

Judith A. Holton, Ph.D.

2010-02-01

Full Text Available Coding is the core process in classic grounded theory methodology. It is through coding that the conceptual abstraction of data and its reintegration as theory takes place. There are two types of coding in a classic grounded theory study: substantive coding, which includes both open and selective coding procedures, and theoretical coding. In substantive coding, the researcher works with the data directly, fracturing and analysing it, initially through open coding for the emergence of a core category and related concepts and then subsequently through theoretical sampling and selective coding of data to theoretically saturate the core and related concepts. Theoretical saturation is achieved through constant comparison of incidents (indicators in the data to elicit the properties and dimensions of each category (code. This constant comparing of incidents continues until the process yields the interchangeability of indicators, meaning that no new properties or dimensions are emerging from continued coding and comparison. At this point, the concepts have achieved theoretical saturation and the theorist shifts attention to exploring the emergent fit of potential theoretical codes that enable the conceptual integration of the core and related concepts to produce hypotheses that account for relationships between the concepts thereby explaining the latent pattern of social behaviour that forms the basis of the emergent theory. The coding of data in grounded theory occurs in conjunction with analysis through a process of conceptual memoing, capturing the theorist’s ideation of the emerging theory. Memoing occurs initially at the substantive coding level and proceeds to higher levels of conceptual abstraction as coding proceeds to theoretical saturation and the theorist begins to explore conceptual reintegration through theoretical coding.

Code-experiment comparison on wall condensation tests in the presence of non-condensable gases-Numerical calculations for containment studies

Energy Technology Data Exchange (ETDEWEB)

Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), PSN-RES, SCA, BP 68, 91192 Gif-sur-Yvette (France); Porcheron, E.; Dumay, F.; Vendel, J. [Institut de Radioprotection et de Surete Nucleaire (IRSN), PSN-RES, SCA, BP 68, 91192 Gif-sur-Yvette (France)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Steam condensation on walls has been investigated in the TOSQAN vessel. Black-Right-Pointing-Pointer Experiments on 7 different tests have been performed. Black-Right-Pointing-Pointer Different steam injections and wall temperatures are used. Black-Right-Pointing-Pointer Simulations are performed in 2D using the TONUS code. Black-Right-Pointing-Pointer Code-experiments comparisons at many different locations show a good agreement. - Abstract: During the course of a severe Pressurized Water Reactor accident, pressurization of the containment occurs and hydrogen can be produced by the reactor core oxidation and distributed in the containment according to convection flows and wall condensation. Filmwise wall condensation in the presence of non-condensable gases is a subject of many interests and extensive studies have been performed in the past. Some empirical correlations have demonstrated their limit for extrapolation under different thermal-hydraulic conditions and at different geometries/scales. The French Institute for Radiological Protection and Nuclear Safety (IRSN) has developed a numerical tool and an experimental facility in order to investigate free convection flows in the presence of condensation. The objective of this paper is to present numerical results obtained on different wall condensation tests in 7 m{sup 3} volume vessel (TOSQAN facility), and to compare them with the experimental ones. Over eight tests are considered here, and code-experiment comparison is performed on many different locations, giving an extensive insight of the code assessment for air-steam mixture flows involving wall condensation in the presence of non-condensable gases.

An Efficient Method for Verifying Gyrokinetic Microstability Codes

Science.gov (United States)

Bravenec, R.; Candy, J.; Dorland, W.; Holland, C.

2009-11-01

Benchmarks for gyrokinetic microstability codes can be developed through successful ``apples-to-apples'' comparisons among them. Unlike previous efforts, we perform the comparisons for actual discharges, rendering the verification efforts relevant to existing experiments and future devices (ITER). The process requires i) assembling the experimental analyses at multiple times, radii, discharges, and devices, ii) creating the input files ensuring that the input parameters are faithfully translated code-to-code, iii) running the codes, and iv) comparing the results, all in an organized fashion. The purpose of this work is to automate this process as much as possible: At present, a python routine is used to generate and organize GYRO input files from TRANSP or ONETWO analyses. Another routine translates the GYRO input files into GS2 input files. (Translation software for other codes has not yet been written.) Other python codes submit the multiple GYRO and GS2 jobs, organize the results, and collect them into a table suitable for plotting. (These separate python routines could easily be consolidated.) An example of the process -- a linear comparison between GYRO and GS2 for a DIII-D discharge at multiple radii -- will be presented.

Verification and benchmarking of PORFLO: an equivalent porous continuum code for repository scale analysis

International Nuclear Information System (INIS)

Eyler, L.L.; Budden, M.J.

1984-11-01

The objective of this work was to perform an assessment of prediction capabilities and features of the PORFLO code in relation to its intended use in the Basalt Waste Isolation Project. This objective was to be accomplished through a code verification and benchmarking task. Results were to be documented which either support correctness of prediction capabilities or identify areas of intended application in which the code exhibits weaknesses. A test problem set consisting of 10 problems was developed. Results of PORFLO simulations of these problems were provided for use in this work. The 10 problems were designed to test the three basic computational capabilities or categories of the code. Broken down by physical process, these are heat transfer, fluid flow, and radionuclide transport. Two verification problems were included within each of these categories. They were problems designed to test basic features of PORFLO for which analytical solutions are available for use as a known comparison basis. Hence they are referred to as verification problems. Of the remaining four problems, one repository scale problem representative of intended PORFLO use within BWIP was included in each of the three basic capabilities categories. The remaining problem was a case specifically designed to test features of decay and retardation in radionuclide transport. These four problems are referred to as benchmarking problems, because results computed with an additional computer code were used as a basis for comparison. 38 figures

Case Study of the NENE Code Project

National Research Council Canada - National Science Library

Kendall, Richard; Post, Douglass; Mark, Andrew

2007-01-01

...) Program is sponsoring a series of case studies to identify the life cycles, workflows, and technical challenges of computational science and engineering code development that are representative...

Performance evaluation based on data from code reviews

OpenAIRE

Andrej, Sekáč

2016-01-01

Context. Modern code review tools such as Gerrit have made available great amounts of code review data from different open source projects as well as other commercial projects. Code reviews are used to keep the quality of produced source code under control but the stored data could also be used for evaluation of the software development process. Objectives. This thesis uses machine learning methods for an approximation of review expert’s performance evaluation function. Due to limitations in ...

Optix: A Monte Carlo scintillation light transport code

Energy Technology Data Exchange (ETDEWEB)

Safari, M.J., E-mail: mjsafari@aut.ac.ir [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Ghal-Eh, N. [School of Physics, Damghan University, PO Box 36716-41167, Damghan (Iran, Islamic Republic of); Davani, F. Abbasi [Nuclear Engineering Department, Shahid Beheshti University, PO Box 1983963113, Tehran (Iran, Islamic Republic of)

2014-02-11

The paper reports on the capabilities of Monte Carlo scintillation light transport code Optix, which is an extended version of previously introduced code Optics. Optix provides the user a variety of both numerical and graphical outputs with a very simple and user-friendly input structure. A benchmarking strategy has been adopted based on the comparison with experimental results, semi-analytical solutions, and other Monte Carlo simulation codes to verify various aspects of the developed code. Besides, some extensive comparisons have been made against the tracking abilities of general-purpose MCNPX and FLUKA codes. The presented benchmark results for the Optix code exhibit promising agreements. -- Highlights: • Monte Carlo simulation of scintillation light transport in 3D geometry. • Evaluation of angular distribution of detected photons. • Benchmark studies to check the accuracy of Monte Carlo simulations.

GENII [Generation II]: The Hanford Environmental Radiation Dosimetry Software System: Volume 3, Code maintenance manual: Hanford Environmental Dosimetry Upgrade Project

International Nuclear Information System (INIS)

Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.

1988-09-01

The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). This coupled system of computer codes is intended for analysis of environmental contamination resulting from acute or chronic releases to, or initial contamination of, air, water, or soil, on through the calculation of radiation doses to individuals or populations. GENII is described in three volumes of documentation. This volume is a Code Maintenance Manual for the serious user, including code logic diagrams, global dictionary, worksheets to assist with hand calculations, and listings of the code and its associated data libraries. The first volume describes the theoretical considerations of the system. The second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. 7 figs., 5 tabs

GENII (Generation II): The Hanford Environmental Radiation Dosimetry Software System: Volume 3, Code maintenance manual: Hanford Environmental Dosimetry Upgrade Project

Energy Technology Data Exchange (ETDEWEB)

Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.

1988-09-01

The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). This coupled system of computer codes is intended for analysis of environmental contamination resulting from acute or chronic releases to, or initial contamination of, air, water, or soil, on through the calculation of radiation doses to individuals or populations. GENII is described in three volumes of documentation. This volume is a Code Maintenance Manual for the serious user, including code logic diagrams, global dictionary, worksheets to assist with hand calculations, and listings of the code and its associated data libraries. The first volume describes the theoretical considerations of the system. The second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. 7 figs., 5 tabs.

Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)

Energy Technology Data Exchange (ETDEWEB)

Agrawal, A.K.

1978-02-01

Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation.

Advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)

International Nuclear Information System (INIS)

Agrawal, A.K.

1978-02-01

Physical models for various processes that are encountered in preaccident and transient simulation of thermohydraulic transients in the entire liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-L, was written as a part of the Super System Code (SSC) development project for the ''loop''-type designs of LMFBRs. This code has the self-starting capability, i.e., preaccident or steady-state calculations are performed internally. These results then serve as the starting point for the transient simulation

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

On the adequacy of numerical codes for the simulation of vapour cloud explosions

International Nuclear Information System (INIS)

Wingerden, G.J.M.v.; Berg, A.C.v.d.

1984-01-01

Three spherically symmetric blast simulation codes have been evaluated: a low-flame-speed model (Piston model) and two gasdynamic blast simulation codes (BLAST and CLOUD). Self-similar flow fields in front of constant velocity flames and large- and small-scale spherically symmetric explosions experiments were simulated. The Piston model can be used for the simulation of spherically symmetric explosions at flame speeds -1 whereas BLAST and CLOUD are adequate for flame speeds exceeding 100 ms -1 . An adapted Piston code has been investigated with respect to the capability of simulating blast due to explosions of pancake-shaped clouds. Comparison to an acoustic approach showed that the Piston model can be regarded as an acoustic model with the possibility of handling every imaginable flame path. The research was part of the indirect action research programme on LWR Safety of the Commission of the European Communities. (project 12B, contract 008 SRN)

Reliability of cause of death coding: an international comparison.

Science.gov (United States)

Antini, Carmen; Rajs, Danuta; Muñoz-Quezada, María Teresa; Mondaca, Boris Andrés Lucero; Heiss, Gerardo

2015-07-01

This study evaluates the agreement of nosologic coding of cardiovascular causes of death between a Chilean coder and one in the United States, in a stratified random sample of death certificates of persons aged ≥ 60, issued in 2008 in the Valparaíso and Metropolitan regions, Chile. All causes of death were converted to ICD-10 codes in parallel by both coders. Concordance was analyzed with inter-coder agreement and Cohen's kappa coefficient by level of specification ICD-10 code for the underlying cause and the total causes of death coding. Inter-coder agreement was 76.4% for all causes of death and 80.6% for the underlying cause (agreement at the four-digit level), with differences by the level of specification of the ICD-10 code, by line of the death certificate, and by number of causes of death per certificate. Cohen's kappa coefficient was 0.76 (95%CI: 0.68-0.84) for the underlying cause and 0.75 (95%CI: 0.74-0.77) for the total causes of death. In conclusion, causes of death coding and inter-coder agreement for cardiovascular diseases in two regions of Chile are comparable to an external benchmark and with reports from other countries.

Upgrades to the WIMS-ANL code

International Nuclear Information System (INIS)

Woodruff, W. L.

1998-01-01

The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries

Upgrades to the WIMS-ANL code

International Nuclear Information System (INIS)

Woodruff, W.L.; Leopando, L.S.

1998-01-01

The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries. (author)

Increasing the Value of Research: A Comparison of the Literature on Critical Success Factors for Projects, IT Projects and Enterprise Resource Planning Projects

Directory of Open Access Journals (Sweden)

Annie Maddison Warren

2016-11-01

Full Text Available Since the beginning of modern project management in the 1960s, academic researchers have sought to identify a definitive list of Critical Success Factors (CSFs, the key things that project managers must get right in order to deliver a successful product. With the advent of Information Technology (IT projects and, more recently, projects to deliver Enterprise Resource Planning (ERP systems, attention has turned to identifying definitive lists of CSFs for these more specific project types. The purpose of this paper is to take stock of this research effort by examining how thinking about each type of project has evolved over time, before producing a consolidated list of CSFs for each as a basis for comparison. This process reveals a high degree of similarity, leading to the conclusion that the goal of identifying a generic list of CSFs for project management has been achieved. Therefore, rather than continuing to describe lists of CSFs, researchers could increase the value of their contribution by taking a step forward and focusing on why, despite this apparent knowledge of how to ensure their success, ERP projects continue to fail.

Genome-wide identification of coding and non-coding conserved sequence tags in human and mouse genomes

Directory of Open Access Journals (Sweden)

Maggi Giorgio P

2008-06-01

Full Text Available Abstract Background The accurate detection of genes and the identification of functional regions is still an open issue in the annotation of genomic sequences. This problem affects new genomes but also those of very well studied organisms such as human and mouse where, despite the great efforts, the inventory of genes and regulatory regions is far from complete. Comparative genomics is an effective approach to address this problem. Unfortunately it is limited by the computational requirements needed to perform genome-wide comparisons and by the problem of discriminating between conserved coding and non-coding sequences. This discrimination is often based (thus dependent on the availability of annotated proteins. Results In this paper we present the results of a comprehensive comparison of human and mouse genomes performed with a new high throughput grid-based system which allows the rapid detection of conserved sequences and accurate assessment of their coding potential. By detecting clusters of coding conserved sequences the system is also suitable to accurately identify potential gene loci. Following this analysis we created a collection of human-mouse conserved sequence tags and carefully compared our results to reliable annotations in order to benchmark the reliability of our classifications. Strikingly we were able to detect several potential gene loci supported by EST sequences but not corresponding to as yet annotated genes. Conclusion Here we present a new system which allows comprehensive comparison of genomes to detect conserved coding and non-coding sequences and the identification of potential gene loci. Our system does not require the availability of any annotated sequence thus is suitable for the analysis of new or poorly annotated genomes.

LIONS: a new set of Fortran90 codes for the SPIRAL project at GANIL

International Nuclear Information System (INIS)

Bertrand, P.

1995-01-01

In this paper a set of new computer programs developed at GANIL is presented. These codes are used to study different parts of the SPIRAL project, in particular the dynamics in the CIME cyclotron and the new extraction system of the ECR ion sources. Three important modules are described: CHA3D for the evaluation of 3D electric fields with or without space charge effects, LIONS for the motion of ions and EXTRACT for the ECRIS extraction. These modules are written in Fortran90 in a ''data parallel scheme''. They work either on UNIX workstations or parallel and vectorial computers. (orig.)

Cross checking of the new capabilities of the fuel cycle scenario code TR-EVOL - 5229

International Nuclear Information System (INIS)

Merino-Rodriguez, I.; Garcia-Martinez, M.; Alvarez-Velarde, F.

2015-01-01

This work is intended to cross check the new capabilities of the fuel cycle scenario code TR-EVOL by means of comparing its results with those published in bibliography. This process has been divided in two stages as follows. The first stage is dedicated to check the improvements in the material management part of the fuel cycle code (the nuclear fuel mass balance estimation). The Spanish nuclear fuel cycle has been chosen as the model for the mass balance comparison given that the fuel mass per reactor is available in bibliography. The second stage has been focused in verifying the validity of the TR-EVOL economic module. The economic model verification has been carried out by making use of the ARCAS EU project and its economic assessments for advanced reactors and scenarios involving fast reactors and ADS. As conclusions, the main finding from the first stage includes that TR-EVOL provides a prediction of mass values quite accurate after the improvements and when using the proper parameters as input for the code. For the second stage, results were highly satisfactory since a difference smaller than 3% can be found regarding results published by the ARCAS project (NRG estimations). Furthermore, concerning the Decommissioning, Dismantling and Disposal cost, results are highly acceptable (7% difference in the comparison with the final disposal in a once-through scenario and around 11% in a final disposal with a reprocessing strategy) given the difficulties to find in bibliography detailed information about the costs of the final disposals and the significant uncertainties involved in design concepts and related unit costs

Structured Light Based 3d Scanning for Specular Surface by the Combination of Gray Code and Phase Shifting

Science.gov (United States)

Zhang, Yujia; Yilmaz, Alper

2016-06-01

Surface reconstruction using coded structured light is considered one of the most reliable techniques for high-quality 3D scanning. With a calibrated projector-camera stereo system, a light pattern is projected onto the scene and imaged by the camera. Correspondences between projected and recovered patterns are computed in the decoding process, which is used to generate 3D point cloud of the surface. However, the indirect illumination effects on the surface, such as subsurface scattering and interreflections, will raise the difficulties in reconstruction. In this paper, we apply maximum min-SW gray code to reduce the indirect illumination effects of the specular surface. We also analysis the errors when comparing the maximum min-SW gray code and the conventional gray code, which justifies that the maximum min-SW gray code has significant superiority to reduce the indirect illumination effects. To achieve sub-pixel accuracy, we project high frequency sinusoidal patterns onto the scene simultaneously. But for specular surface, the high frequency patterns are susceptible to decoding errors. Incorrect decoding of high frequency patterns will result in a loss of depth resolution. Our method to resolve this problem is combining the low frequency maximum min-SW gray code and the high frequency phase shifting code, which achieves dense 3D reconstruction for specular surface. Our contributions include: (i) A complete setup of the structured light based 3D scanning system; (ii) A novel combination technique of the maximum min-SW gray code and phase shifting code. First, phase shifting decoding with sub-pixel accuracy. Then, the maximum min-SW gray code is used to resolve the ambiguity resolution. According to the experimental results and data analysis, our structured light based 3D scanning system enables high quality dense reconstruction of scenes with a small number of images. Qualitative and quantitative comparisons are performed to extract the advantages of our new

Description of ground motion data processing codes: Volume 3

International Nuclear Information System (INIS)

Sanders, M.L.

1988-02-01

Data processing codes developed to process ground motion at the Nevada Test Site for the Weapons Test Seismic Investigations Project are used today as part of the program to process ground motion records for the Nevada Nuclear Waste Storage Investigations Project. The work contained in this report documents and lists codes and verifies the ''PSRV'' code. 39 figs

Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications

International Nuclear Information System (INIS)

2013-12-01

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

Data model description for the DESCARTES and CIDER codes

International Nuclear Information System (INIS)

Miley, T.B.; Ouderkirk, S.J.; Nichols, W.E.; Eslinger, P.W.

1993-01-01

The primary objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the US Department of Energy's (DOE) Hanford Site near Richland, Washington. One of the major objectives of the HEDR Project is to develop several computer codes to model the airborne releases. transport and envirorunental accumulation of radionuclides resulting from Hanford operations from 1944 through 1972. In July 1992, the HEDR Project Manager determined that the computer codes being developed (DESCARTES, calculation of environmental accumulation from airborne releases, and CIDER, dose calculations from environmental accumulation) were not sufficient to create accurate models. A team of HEDR staff members developed a plan to assure that computer codes would meet HEDR Project goals. The plan consists of five tasks: (1) code requirements definition. (2) scoping studies, (3) design specifications, (4) benchmarking, and (5) data modeling. This report defines the data requirements for the DESCARTES and CIDER codes

Low Complexity List Decoding for Polar Codes with Multiple CRC Codes

Directory of Open Access Journals (Sweden)

Jong-Hwan Kim

2017-04-01

Full Text Available Polar codes are the first family of error correcting codes that provably achieve the capacity of symmetric binary-input discrete memoryless channels with low complexity. Since the development of polar codes, there have been many studies to improve their finite-length performance. As a result, polar codes are now adopted as a channel code for the control channel of 5G new radio of the 3rd generation partnership project. However, the decoder implementation is one of the big practical problems and low complexity decoding has been studied. This paper addresses a low complexity successive cancellation list decoding for polar codes utilizing multiple cyclic redundancy check (CRC codes. While some research uses multiple CRC codes to reduce memory and time complexity, we consider the operational complexity of decoding, and reduce it by optimizing CRC positions in combination with a modified decoding operation. Resultingly, the proposed scheme obtains not only complexity reduction from early stopping of decoding, but also additional reduction from the reduced number of decoding paths.

International assessment of PCA codes

International Nuclear Information System (INIS)

Neymotin, L.; Lui, C.; Glynn, J.; Archarya, S.

1993-11-01

Over the past three years (1991-1993), an extensive international exercise for intercomparison of a group of six Probabilistic Consequence Assessment (PCA) codes was undertaken. The exercise was jointly sponsored by the Commission of European Communities (CEC) and OECD Nuclear Energy Agency. This exercise was a logical continuation of a similar effort undertaken by OECD/NEA/CSNI in 1979-1981. The PCA codes are currently used by different countries for predicting radiological health and economic consequences of severe accidents at nuclear power plants (and certain types of non-reactor nuclear facilities) resulting in releases of radioactive materials into the atmosphere. The codes participating in the exercise were: ARANO (Finland), CONDOR (UK), COSYMA (CEC), LENA (Sweden), MACCS (USA), and OSCAAR (Japan). In parallel with this inter-code comparison effort, two separate groups performed a similar set of calculations using two of the participating codes, MACCS and COSYMA. Results of the intercode and inter-MACCS comparisons are presented in this paper. The MACCS group included four participants: GREECE: Institute of Nuclear Technology and Radiation Protection, NCSR Demokritos; ITALY: ENEL, ENEA/DISP, and ENEA/NUC-RIN; SPAIN: Universidad Politecnica de Madrid (UPM) and Consejo de Seguridad Nuclear; USA: Brookhaven National Laboratory, US NRC and DOE

Translation of ARAC computer codes

International Nuclear Information System (INIS)

Takahashi, Kunio; Chino, Masamichi; Honma, Toshimitsu; Ishikawa, Hirohiko; Kai, Michiaki; Imai, Kazuhiko; Asai, Kiyoshi

1982-05-01

In 1981 we have translated the famous MATHEW, ADPIC and their auxiliary computer codes for CDC 7600 computer version to FACOM M-200's. The codes consist of a part of the Atmospheric Release Advisory Capability (ARAC) system of Lawrence Livermore National Laboratory (LLNL). The MATHEW is a code for three-dimensional wind field analysis. Using observed data, it calculates the mass-consistent wind field of grid cells by a variational method. The ADPIC is a code for three-dimensional concentration prediction of gases and particulates released to the atmosphere. It calculates concentrations in grid cells by the particle-in-cell method. They are written in LLLTRAN, i.e., LLNL Fortran language and are implemented on the CDC 7600 computers of LLNL. In this report, i) the computational methods of the MATHEW/ADPIC and their auxiliary codes, ii) comparisons of the calculated results with our JAERI particle-in-cell, and gaussian plume models, iii) translation procedures from the CDC version to FACOM M-200's, are described. Under the permission of LLNL G-Division, this report is published to keep the track of the translation procedures and to serve our JAERI researchers for comparisons and references of their works. (author)

Validation of One-Dimensional Module of MARS-KS1.2 Computer Code By Comparison with the RELAP5/MOD3.3/patch3 Developmental Assessment Results

International Nuclear Information System (INIS)

Bae, S. W.; Chung, B. D.

2010-07-01

This report records the results of the code validation for the one-dimensional module of the MARS-KS thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 Code Developmental Assessment Problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The result suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

Assessment of systems codes and their coupling with CFD codes in thermal–hydraulic applications to innovative reactors

Energy Technology Data Exchange (ETDEWEB)

Bandini, G., E-mail: giacomino.bandini@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Polidori, M. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Gerschenfeld, A.; Pialla, D.; Li, S. [Commissariat à l’Energie Atomique (CEA) (France); Ma, W.M.; Kudinov, P.; Jeltsov, M.; Kööp, K. [Royal Institute of Technology (KTH) (Sweden); Huber, K.; Cheng, X.; Bruzzese, C.; Class, A.G.; Prill, D.P. [Karlsruhe Institute of Technology (KIT) (Germany); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Geffray, C.; Macian-Juan, R. [Technische Universität München (TUM) (Germany); Maas, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) (France)

2015-01-15

Highlights: • The assessment of RELAP5, TRACE and CATHARE system codes on integral experiments is presented. • Code benchmark of CATHARE, DYN2B, and ATHLET on PHENIX natural circulation experiment. • Grid-free pool modelling based on proper orthogonal decomposition for system codes is explained. • The code coupling methodologies are explained. • The coupling of several CFD/system codes is tested against integral experiments. - Abstract: The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal–hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal–hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal–hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.

A study on the nuclear computer code maintenance and management system

International Nuclear Information System (INIS)

Kim, Yeon Seung; Huh, Young Hwan; Lee, Jong Bok; Choi, Young Gil; Suh, Soong Hyok; Kang, Byong Heon; Kim, Hee Kyung; Kim, Ko Ryeo; Park, Soo Jin

1990-12-01

According to current software development and quality assurance trends. It is necessary to develop computer code management system for nuclear programs. For this reason, the project started in 1987. Main objectives of the project are to establish a nuclear computer code management system, to secure software reliability, and to develop nuclear computer code packages. Contents of performing the project in this year were to operate and maintain computer code information system of KAERI computer codes, to develop application tool, AUTO-i, for solving the 1st and 2nd moments of inertia on polygon or circle, and to research nuclear computer code conversion between different machines. For better supporting the nuclear code availability and reliability, assistance from users who are using codes is required. Lastly, for easy reference about the codes information, we presented list of code names and information on the codes which were introduced or developed during this year. (Author)

GMMIP (v1.0) contribution to CMIP6: Global Monsoons Model Inter-comparison Project

Energy Technology Data Exchange (ETDEWEB)

Zhou, Tianjun; Turner, Andrew G.; Kinter, James L.; Wang, Bin; Qian, Yun; Chen, Xiaolong; Wu, Bo; Wang, Bin; Liu, Bo; Zou, Liwei; He, Bian

2016-10-10

The Global Monsoons Model Inter-comparison Project (GMMIP) has been endorsed by the panel of Coupled Model Inter-comparison Project (CMIP) as one of the participating model inter-comparison projects (MIPs) in the sixth phase of CMIP (CMIP6). The focus of GMMIP is on monsoon climatology, variability, prediction and projection, which is relevant to four of the “Grand Challenges” proposed by the World Climate Research Programme. At present, 21 international modeling groups are committed to joining GMMIP. This overview paper introduces the motivation behind GMMIP and the scientific questions it intends to answer. Three tiers of experiments, of decreasing priority, are designed to examine (a) model skill in simulating the climatology and interannual-to-multidecadal variability of global monsoons forced by the sea surface temperature during historical climate period; (b) the roles of the Interdecadal Pacific Oscillation and Atlantic Multidecadal Oscillation in driving variations of the global and regional monsoons; and (c) the effects of large orographic terrain on the establishment of the monsoons. The outputs of the CMIP6 Diagnostic, Evaluation and Characterization of Klima experiments (DECK), “historical” simulation and endorsed MIPs will also be used in the diagnostic analysis of GMMIP to give a comprehensive understanding of the roles played by different external forcings, potential improvements in the simulation of monsoon rainfall at high resolution and reproducibility at decadal timescales. The implementation of GMMIP will improve our understanding of the fundamental physics of changes in the global and regional monsoons over the past 140 years and ultimately benefit monsoons prediction and projection in the current century.

Fast comparison of IS radar code sequences for lag profile inversion

Directory of Open Access Journals (Sweden)

M. S. Lehtinen

2008-08-01

Full Text Available A fast method for theoretically comparing the posteriori variances produced by different phase code sequences in incoherent scatter radar (ISR experiments is introduced. Alternating codes of types 1 and 2 are known to be optimal for selected range resolutions, but the code sets are inconveniently long for many purposes like ground clutter estimation and in cases where coherent echoes from lower ionospheric layers are to be analyzed in addition to standard F-layer spectra.

The method is used in practice for searching binary code quads that have estimation accuracy almost equal to that of much longer alternating code sets. Though the code sequences can consist of as few as four different transmission envelopes, the lag profile estimation variances are near to the theoretical minimum. Thus the short code sequence is equally good as a full cycle of alternating codes with the same pulse length and bit length. The short code groups cannot be directly decoded, but the decoding is done in connection with more computationally expensive lag profile inversion in data analysis.

The actual code searches as well as the analysis and real data results from the found short code searches are explained in other papers sent to the same issue of this journal. We also discuss interesting subtle differences found between the different alternating codes by this method. We assume that thermal noise dominates the incoherent scatter signal.

Integral large scale experiments on hydrogen combustion for severe accident code validation-HYCOM

International Nuclear Information System (INIS)

Breitung, W.; Dorofeev, S.; Kotchourko, A.; Redlinger, R.; Scholtyssek, W.; Bentaib, A.; L'Heriteau, J.-P.; Pailhories, P.; Eyink, J.; Movahed, M.; Petzold, K.-G.; Heitsch, M.; Alekseev, V.; Denkevits, A.; Kuznetsov, M.; Efimenko, A.; Okun, M.V.; Huld, T.; Baraldi, D.

2005-01-01

A joint research project was carried out in the EU Fifth Framework Programme, concerning hydrogen risk in a nuclear power plant. The goals were: Firstly, to create a new data base of results on hydrogen combustion experiments in the slow to turbulent combustion regimes. Secondly, to validate the partners CFD and lumped parameter codes on the experimental data, and to evaluate suitable parameter sets for application calculations. Thirdly, to conduct a benchmark exercise by applying the codes to the full scale analysis of a postulated hydrogen combustion scenario in a light water reactor containment after a core melt accident. The paper describes the work programme of the project and the partners activities. Significant progress has been made in the experimental area, where test series in medium and large scale facilities have been carried out with the focus on specific effects of scale, multi-compartent geometry, heat losses and venting. The data were used for the validation of the partners CFD and lumped parameter codes, which included blind predictive calculations and pre- and post-test intercomparison exercises. Finally, a benchmark exercise was conducted by applying the codes to the full scale analysis of a hydrogen combustion scenario. The comparison and assessment of the results of the validation phase and of the challenging containment calculation exercise allows a deep insight in the quality, capabilities and limits of the CFD and the lumped parameter tools which are currently in use at various research laboratories

Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Esquivel E, J.

2016-09-01

The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

Status of the CONTAIN computer code for LWR containment analysis

International Nuclear Information System (INIS)

Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

1983-01-01

The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

Status of the CONTAIN computer code for LWR containment analysis

International Nuclear Information System (INIS)

Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

1982-01-01

The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

OPAL reactor calculations using the Monte Carlo code serpent

Energy Technology Data Exchange (ETDEWEB)

Ferraro, Diego; Villarino, Eduardo [Nuclear Engineering Dept., INVAP S.E., Rio Negro (Argentina)

2012-03-15

In the present work the Monte Carlo cell code developed by VTT Serpent v1.1.14 is used to model the MTR fuel assemblies (FA) and control rods (CR) from OPAL (Open Pool Australian Light-water) reactor in order to obtain few-group constants with burnup dependence to be used in the already developed reactor core models. These core calculations are performed using CITVAP 3-D diffusion code, which is well-known reactor code based on CITATION. Subsequently the results are compared with those obtained by the deterministic calculation line used by INVAP, which uses the Collision Probability Condor cell-code to obtain few-group constants. Finally the results are compared with the experimental data obtained from the reactor information for several operation cycles. As a result several evaluations are performed, including a code to code cell comparison at cell and core level and calculation-experiment comparison at core level in order to evaluate the Serpent code actual capabilities. (author)

A study on the nuclear computer codes installation and management system

International Nuclear Information System (INIS)

Kim, Yeon Seung; Huh, Young Hwan; Kim, Hee Kyung; Kang, Byung Heon; Kim, Ko Ryeo; Suh, Soong Hyok; Choi, Young Gil; Lee, Jong Bok

1990-12-01

From 1987 a number of technical transfer related to nuclear power plant had been performed from C-E for YGN 3 and 4 construction. Among them, installation and management of the computer codes for YGN 3 and 4 fuel and nuclear steam supply system was one of the most important project. Main objectives of this project are to establish the nuclear computer code management system, to develop QA procedure for nuclear codes, to secure the nuclear code reliability and to extend techanical applicabilities including the user-oriented utility programs for nuclear codes. Contents of performing the project in this year was to produce 215 transmittal packages of nuclear codes installation including making backup magnetic tape and microfiche for software quality assurance. Lastly, for easy reference about the nuclear codes information we presented list of code names and information on the codes which were introduced from C-E. (Author)

Assessment of thema code against spreading experiments

International Nuclear Information System (INIS)

Spindler, B.; Veteau, J.M.; Cecco, L. de; Montanelli, P.; Pineau, D.

2000-01-01

In the frame work of severe accident research, the spreading code THEMA, developed at CEA/DRN, aims at predicting the spreading extent of molten core after a vessel melt-through. The code solves fluid balance equations integrated over the fluid depth for oxidic and/or metallic phases under the shallow water assumption, using a finite difference scheme. Solidification is taken into account through crust formation on the substrate and at contact with the surroundings, as well as increase of fluid viscosity with solid fraction in the melt. A separate energy equation is solved for the solid substrate, including possible ablation. The assessment of THEMA code against the spreading experiments performed in the framework of the corium spreading and coolability project of the European Union is presented. These experiments use either simulating materials at medium (RIT), or at high temperature (KATS), or corium (VULCANO, FARO), conducted at different mass flow rates and with large or low solidification interval. THEMA appears to be able to simulate the whole set of the experiments investigated. Comparison between experimental and computed spreading lengths and substrate temperatures are quite satisfactory. The results show a rather large sensitivity at mass flow rate and inlet temperature, indicating that, generally, efforts should be made to improve the accuracy of the measurements of such parameters in the experiments. (orig.)

Gap Conductance model Validation in the TASS/SMR-S code using MARS code

International Nuclear Information System (INIS)

Ahn, Sang Jun; Yang, Soo Hyung; Chung, Young Jong; Lee, Won Jae

2010-01-01

Korea Atomic Energy Research Institute (KAERI) has been developing the TASS/SMR-S (Transient and Setpoint Simulation/Small and Medium Reactor) code, which is a thermal hydraulic code for the safety analysis of the advanced integral reactor. An appropriate work to validate the applicability of the thermal hydraulic models within the code should be demanded. Among the models, the gap conductance model which is describes the thermal gap conductivity between fuel and cladding was validated through the comparison with MARS code. The validation of the gap conductance model was performed by evaluating the variation of the gap temperature and gap width as the changed with the power fraction. In this paper, a brief description of the gap conductance model in the TASS/SMR-S code is presented. In addition, calculated results to validate the gap conductance model are demonstrated by comparing with the results of the MARS code with the test case

Financial and clinical governance implications of clinical coding accuracy in neurosurgery: a multidisciplinary audit.

Science.gov (United States)

Haliasos, N; Rezajooi, K; O'neill, K S; Van Dellen, J; Hudovsky, Anita; Nouraei, Sar

2010-04-01

Clinical coding is the translation of documented clinical activities during an admission to a codified language. Healthcare Resource Groupings (HRGs) are derived from coding data and are used to calculate payment to hospitals in England, Wales and Scotland and to conduct national audit and benchmarking exercises. Coding is an error-prone process and an understanding of its accuracy within neurosurgery is critical for financial, organizational and clinical governance purposes. We undertook a multidisciplinary audit of neurosurgical clinical coding accuracy. Neurosurgeons trained in coding assessed the accuracy of 386 patient episodes. Where clinicians felt a coding error was present, the case was discussed with an experienced clinical coder. Concordance between the initial coder-only clinical coding and the final clinician-coder multidisciplinary coding was assessed. At least one coding error occurred in 71/386 patients (18.4%). There were 36 diagnosis and 93 procedure errors and in 40 cases, the initial HRG changed (10.4%). Financially, this translated to pound111 revenue-loss per patient episode and projected to pound171,452 of annual loss to the department. 85% of all coding errors were due to accumulation of coding changes that occurred only once in the whole data set. Neurosurgical clinical coding is error-prone. This is financially disadvantageous and with the coding data being the source of comparisons within and between departments, coding inaccuracies paint a distorted picture of departmental activity and subspecialism in audit and benchmarking. Clinical engagement improves accuracy and is encouraged within a clinical governance framework.

Ontario emissions trading code : emission reduction credit creation, recording and transfer rules, rules for renewable energy projects and conservation projects, and rules for the operation of the Ontario Emissions Trading Registry

International Nuclear Information System (INIS)

2001-12-01

Emissions trading has been an integral part of Ontario's air quality strategy since December 31, 2001. Ontario has adopted the 'cap, credit and trade' type of emissions trading system, a hybrid that takes the best features of pure 'cap-and-trade' and 'baseline-and-credit' type systems. It covers nitric oxide and sulphur dioxide. The Ontario Emissions Trading Code supplements Ontario Regulation 397/01 and sets out rules for renewable energy projects and conservation projects for which applications for emission allowances can be made. This Code describes the rules for the creation and transfer of emission reduction credits (ERCs). It also explains the rules for the operation of the registry that has been established to provide information to the public about the emissions trading program and records decisions about credit creation and credit and allowance retirement. 3 tabs

Investigating the Simulink Auto-Coding Process

Science.gov (United States)

Gualdoni, Matthew J.

2016-01-01

Model based program design is the most clear and direct way to develop algorithms and programs for interfacing with hardware. While coding "by hand" results in a more tailored product, the ever-growing size and complexity of modern-day applications can cause the project work load to quickly become unreasonable for one programmer. This has generally been addressed by splitting the product into separate modules to allow multiple developers to work in parallel on the same project, however this introduces new potentials for errors in the process. The fluidity, reliability and robustness of the code relies on the abilities of the programmers to communicate their methods to one another; furthermore, multiple programmers invites multiple potentially differing coding styles into the same product, which can cause a loss of readability or even module incompatibility. Fortunately, Mathworks has implemented an auto-coding feature that allows programmers to design their algorithms through the use of models and diagrams in the graphical programming environment Simulink, allowing the designer to visually determine what the hardware is to do. From here, the auto-coding feature handles converting the project into another programming language. This type of approach allows the designer to clearly see how the software will be directing the hardware without the need to try and interpret large amounts of code. In addition, it speeds up the programming process, minimizing the amount of man-hours spent on a single project, thus reducing the chance of human error as well as project turnover time. One such project that has benefited from the auto-coding procedure is Ramses, a portion of the GNC flight software on-board Orion that has been implemented primarily in Simulink. Currently, however, auto-coding Ramses into C++ requires 5 hours of code generation time. This causes issues if the tool ever needs to be debugged, as this code generation will need to occur with each edit to any part of

A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

International Nuclear Information System (INIS)

Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun

2013-01-01

The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

A Comparative Study on Seismic Analysis of Bangladesh National Building Code (BNBC) with Other Building Codes

Science.gov (United States)

Bari, Md. S.; Das, T.

2013-09-01

Tectonic framework of Bangladesh and adjoining areas indicate that Bangladesh lies well within an active seismic zone. The after effect of earthquake is more severe in an underdeveloped and a densely populated country like ours than any other developed countries. Bangladesh National Building Code (BNBC) was first established in 1993 to provide guidelines for design and construction of new structure subject to earthquake ground motions in order to minimize the risk to life for all structures. A revision of BNBC 1993 is undergoing to make this up to date with other international building codes. This paper aims at the comparison of various provisions of seismic analysis as given in building codes of different countries. This comparison will give an idea regarding where our country stands when it comes to safety against earth quake. Primarily, various seismic parameters in BNBC 2010 (draft) have been studied and compared with that of BNBC 1993. Later, both 1993 and 2010 edition of BNBC codes have been compared graphically with building codes of other countries such as National Building Code of India 2005 (NBC-India 2005), American Society of Civil Engineering 7-05 (ASCE 7-05). The base shear/weight ratios have been plotted against the height of the building. The investigation in this paper reveals that BNBC 1993 has the least base shear among all the codes. Factored Base shear values of BNBC 2010 are found to have increased significantly than that of BNBC 1993 for low rise buildings (≤20 m) around the country than its predecessor. Despite revision of the code, BNBC 2010 (draft) still suggests less base shear values when compared to the Indian and American code. Therefore, this increase in factor of safety against the earthquake imposed by the proposed BNBC 2010 code by suggesting higher values of base shear is appreciable.

VVER 1000 SBO calculations with pressuriser relief valve stuck open with ASTEC computer code

International Nuclear Information System (INIS)

Atanasova, B.P.; Stefanova, A.E.; Groudev, P.P.

2012-01-01

Highlights: ► We modelled the ASTEC input file for accident scenario (SBO) and focused analyses on the behaviour of core degradation. ► We assumed opening and stuck-open of pressurizer relief valve during performance of SBO scenario. ► ASTEC v1.3.2 has been used as a reference code for the comparison study with the new version of ASTEC code. - Abstract: The objective of this paper is to present the results obtained from performing the calculations with ASTEC computer code for the Source Term evaluation for specific severe accident transient. The calculations have been performed with the new version of ASTEC. The ASTEC V2 code version is released by the French IRSN (Institut de Radioprotection at de surete nucleaire) and Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), Germany. This investigation has been performed in the framework of the SARNET2 project (under the Euratom 7th framework program) by Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Science (INRNE-BAS).

3D-shape of objects with straight line-motion by simultaneous projection of color coded patterns

Science.gov (United States)

Flores, Jorge L.; Ayubi, Gaston A.; Di Martino, J. Matías; Castillo, Oscar E.; Ferrari, Jose A.

2018-05-01

In this work, we propose a novel technique to retrieve the 3D shape of dynamic objects by the simultaneous projection of a fringe pattern and a homogeneous light pattern which are both coded in two of the color channels of a RGB image. The fringe pattern, red channel, is used to retrieve the phase by phase-shift algorithms with arbitrary phase-step, while the homogeneous pattern, blue channel, is used to match pixels from the test object in consecutive images, which are acquired at different positions, and thus, to determine the speed of the object. The proposed method successfully overcomes the standard requirement of projecting fringes of two different frequencies; one frequency to extract object information and the other one to retrieve the phase. Validation experiments are presented.

On-going activities in the European JASMIN project for the development and validation of ASTEC-Na SFR safety simulation code - 15072

International Nuclear Information System (INIS)

Girault, N.; Cloarec, L.; Herranz, L.; Bandini, G.; Perez-Martin, S.; Ammirabile, L.

2015-01-01

The 4-year JASMIN collaborative project (Joint Advanced Severe accidents Modelling and Integration for Na-cooled fast reactors), started in Dec.2011 in the frame of the 7. Framework Programme of the European Commission. It aims at developing a new European simulation code, ASTEC-Na, dealing with the primary phase of SFR core disruptive accidents. The development of a new code, based on a robust advanced simulation tool and able to encompass the in-vessel and in-containment phenomena occurring during a severe accident is indeed of utmost interest for advanced and innovative future SFRs for which an enhanced safety level will be required. This code, based on the ASTEC European code system developed by IRSN and GRS for severe accidents in water-cooled reactors, is progressively integrating and capitalizing the state-of-the-art knowledge of SFR accidents through physical model improvement or development of new ones. New models are assessed on in-pile (CABRI, SCARABEE etc...) and out-of pile experiments conducted during the 70's-80's and code-o-code benchmarking with current accident simulation tools for SFRs is also conducted. During the 2 and a half first years of the project, model specifications and developments were conducted and the validation test matrix was built. The first version of ASTEC-Na available in early 2014 already includes a thermal-hydraulics module able to simulate single and two-phase sodium flow conditions, a zero point neutronic model with simple definition of channel and axial dependences of reactivity feedbacks and models derived from SCANAIR IRSN code for simulating fuel pin thermo-mechanical behaviour and fission gas release/retention. Meanwhile, models have been developed in the source term area for in-containment particle generation and particle chemical transformation, but their implementation is still to be done. As a first validation step, the ASTEC-Na calculations were satisfactorily compared to thermal-hydraulics experimental

Development of new two-dimensional spectral/spatial code based on dynamic cyclic shift code for OCDMA system

Science.gov (United States)

Jellali, Nabiha; Najjar, Monia; Ferchichi, Moez; Rezig, Houria

2017-07-01

In this paper, a new two-dimensional spectral/spatial codes family, named two dimensional dynamic cyclic shift codes (2D-DCS) is introduced. The 2D-DCS codes are derived from the dynamic cyclic shift code for the spectral and spatial coding. The proposed system can fully eliminate the multiple access interference (MAI) by using the MAI cancellation property. The effect of shot noise, phase-induced intensity noise and thermal noise are used to analyze the code performance. In comparison with existing two dimensional (2D) codes, such as 2D perfect difference (2D-PD), 2D Extended Enhanced Double Weight (2D-Extended-EDW) and 2D hybrid (2D-FCC/MDW) codes, the numerical results show that our proposed codes have the best performance. By keeping the same code length and increasing the spatial code, the performance of our 2D-DCS system is enhanced: it provides higher data rates while using lower transmitted power and a smaller spectral width.

Establishment the code for prediction of waste volume on NPP decommissioning

International Nuclear Information System (INIS)

Cho, W. H.; Park, S. K.; Choi, Y. D.; Kim, I. S.; Moon, J. K.

2013-01-01

In practice, decommissioning waste volume can be estimated appropriately by finding the differences between prediction and actual operation and considering the operational problem or supplementary matters. So in the nuclear developed countries such as U.S. or Japan, the decommissioning waste volume is predicted on the basis of the experience in their own decommissioning projects. Because of the contamination caused by radioactive material, decontamination activity and management of radio-active waste should be considered in decommissioning of nuclear facility unlike the usual plant or facility. As the decommissioning activity is performed repeatedly, data for similar activities are accumulated, and optimal strategy can be achieved by comparison with the predicted strategy. Therefore, a variety of decommissioning experiences are the most important. In Korea, there is no data on the decommissioning of commercial nuclear power plants yet. However, KAERI has accumulated the basis decommissioning data of nuclear facility through decommissioning of research reactor (KRR-2) and uranium conversion plant (UCP). And DECOMMIS(DECOMMissioning Information Management System) was developed to provide and manage the whole data of decommissioning project. Two codes, FAC code and WBS code, were established in this process. FAC code is the one which is classified by decommissioning target of nuclear facility, and WBS code is classified by each decommissioning activity. The reason why two codes where created is that the codes used in DEFACS (Decommissioning Facility Characterization management System) and DEWOCS (Decommissioning Work-unit productivity Calculation System) are different from each other, and they were classified each purpose. DEFACS which manages the facility needs the code that categorizes facility characteristics, and DEWOCS which calculates unit productivity needs the code that categorizes decommissioning waste volume. KAERI has accumulated decommissioning data of KRR

Detecting non-coding selective pressure in coding regions

Directory of Open Access Journals (Sweden)

Blanchette Mathieu

2007-02-01

Full Text Available Abstract Background Comparative genomics approaches, where orthologous DNA regions are compared and inter-species conserved regions are identified, have proven extremely powerful for identifying non-coding regulatory regions located in intergenic or intronic regions. However, non-coding functional elements can also be located within coding region, as is common for exonic splicing enhancers, some transcription factor binding sites, and RNA secondary structure elements affecting mRNA stability, localization, or translation. Since these functional elements are located in regions that are themselves highly conserved because they are coding for a protein, they generally escaped detection by comparative genomics approaches. Results We introduce a comparative genomics approach for detecting non-coding functional elements located within coding regions. Codon evolution is modeled as a mixture of codon substitution models, where each component of the mixture describes the evolution of codons under a specific type of coding selective pressure. We show how to compute the posterior distribution of the entropy and parsimony scores under this null model of codon evolution. The method is applied to a set of growth hormone 1 orthologous mRNA sequences and a known exonic splicing elements is detected. The analysis of a set of CORTBP2 orthologous genes reveals a region of several hundred base pairs under strong non-coding selective pressure whose function remains unknown. Conclusion Non-coding functional elements, in particular those involved in post-transcriptional regulation, are likely to be much more prevalent than is currently known. With the numerous genome sequencing projects underway, comparative genomics approaches like that proposed here are likely to become increasingly powerful at detecting such elements.

Calculations of Fission Gas Release During Ramp Tests Using Copernic Code

Energy Technology Data Exchange (ETDEWEB)

Tong, Liu [Nuclear Fuel R and D Center, China Nuclear Power Technology Research Institute (CNPRI) (China)

2013-03-15

The report performed under IAEA research contract No.15951 describes the results of fuel performance evaluation of LWR fuel rods operated at ramp conditions using the COPERNIC code developed by AREVA. The experimental data from the Third Riso Fission Gas Project and the Studsvik SUPER-RAMP Project presented in the IFPE database of the OECD/NEA has been utilized for assessing the code itself during simulation of fission gas release (FGR). Standard code models for LWR fuel were used in simulations with parameters set properly in accordance with relevant test reports. With the help of data adjustment, the input power histories are restructured to fit the real ones, so as to ensure the validity of FGR prediction. The results obtained by COPERNIC show that different models lead to diverse predictions and discrepancies. By comparison, the COPERNIC V2.2 model (95% Upper bound) is selected as the standard FGR model in this report and the FGR phenomenon is properly simulated by the code. To interpret the large discrepancies of some certain PK rods, the burst effect of FGR which is taken into consideration in COPERNIC is described and the influence of the input power histories is extrapolated. In addition, the real-time tracking capability of COPERNIC is tested against experimental data. In the process of investigation, two main dominant factors influencing the measured gas release rate are described and different mechanisms are analyzed. With the limited predicting capacity, accurate predictions cannot be carried out on abrupt changes of FGR during ramp tests by COPERNIC and improvements may be necessary to some relevant models. (author)

Validating the BISON fuel performance code to integral LWR experiments

Energy Technology Data Exchange (ETDEWEB)

Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

2016-05-15

Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

A program for undergraduate research into the mechanisms of sensory coding and memory decay

Energy Technology Data Exchange (ETDEWEB)

Calin-Jageman, R J

2010-09-28

This is the final technical report for this DOE project, entitltled "A program for undergraduate research into the mechanisms of sensory coding and memory decay". The report summarizes progress on the three research aims: 1) to identify phyisological and genetic correlates of long-term habituation, 2) to understand mechanisms of olfactory coding, and 3) to foster a world-class undergraduate neuroscience program. Progress on the first aim has enabled comparison of learning-regulated transcripts across closely related learning paradigms and species, and results suggest that only a small core of transcripts serve truly general roles in long-term memory. Progress on the second aim has enabled testing of several mutant phenotypes for olfactory behaviors, and results show that responses are not fully consistent with the combinitoral coding hypothesis. Finally, 14 undergraduate students participated in this research, the neuroscience program attracted extramural funding, and we completed a successful summer program to enhance transitions for community-college students into 4-year colleges to persue STEM fields.

An experimental comparison of coded modulation strategies for 100 Gb/s transceivers

NARCIS (Netherlands)

Sillekens, E.; Alvarado, A.; Okonkwo, C.; Thomsen, B.C.

2016-01-01

Coded modulation is a key technique to increase the spectral efficiency of coherent optical communication systems. Two popular strategies for coded modulation are turbo trellis-coded modulation (TTCM) and bit-interleaved coded modulation (BICM) based on low-density parity-check (LDPC) codes.

MCNP and GADRAS Comparisons

International Nuclear Information System (INIS)

Klasky, Marc Louis; Myers, Steven Charles; James, Michael R.; Mayo, Douglas R.

2016-01-01

To facilitate the timely execution of System Threat Reviews (STRs) for DNDO, and also to develop a methodology for performing STRs, LANL performed comparisons of several radiation transport codes (MCNP, GADRAS, and Gamma-Designer) that have been previously utilized to compute radiation signatures. While each of these codes has strengths, it is of paramount interest to determine the limitations of each of the respective codes and also to identify the most time efficient means by which to produce computational results, given the large number of parametric cases that are anticipated in performing STR's. These comparisons serve to identify regions of applicability for each code and provide estimates of uncertainty that may be anticipated. Furthermore, while performing these comparisons, examination of the sensitivity of the results to modeling assumptions was also examined. These investigations serve to enable the creation of the LANL methodology for performing STRs. Given the wide variety of radiation test sources, scenarios, and detectors, LANL calculated comparisons of the following parameters: decay data, multiplicity, device (n,γ) leakages, and radiation transport through representative scenes and shielding. This investigation was performed to understand potential limitations utilizing specific codes for different aspects of the STR challenges.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Code Package to Analyze Parameters of the WWER Fuel Rod. TOPRA-2 Code - Verification Data

International Nuclear Information System (INIS)

Scheglov, A.; Proselkov, V.; Passage, G.; Stefanova, S.

2009-01-01

Presented are the data for computer codes to analyze WWER fuel rods, used in the WWER department of RRC 'Kurchatov Institute'. Presented is the description of TOPRA-2 code intended for the engineering analysis of thermophysical and strength parameters of the WWER fuel rod - temperature distributions along the fuel radius, gas pressures under the cladding, stresses in the cladding, etc. for the reactor operation in normal conditions. Presented are some results of the code verification against test problems and the data obtained in the experimental programs. Presented are comparison results of the calculations with TOPRA-2 and TRANSURANUS (V1M1J06) codes. Results obtained in the course of verification demonstrate possibility of application of the methodology and TOPRA-2 code for the engineering analysis of the WWER fuel rods

Direct/Delayed Response Project: Soil-characterization comparison

International Nuclear Information System (INIS)

Fenstermaker, L.K.; Byers, G.E.; Starks, T.H.; Miah, M.J.; Palmer, C.J.

1992-01-01

A large amount of soil characterization data has been collected as a component of the Direct/Delayed Response Project (DDRP) in the acid rain Aquatic Effects Research Program. An interlaboratory comparison study was undertaken to identify the comparability of the data to that obtained from representative soil characterization laboratories. Participating laboratories were selected at random from four regions of the U.S. and two regions of Canada. Two original DDRP contract laboratories also participated. Duplicate samples of six soil audit materials and two liquid soil extracts were sent to each of the laboratories in two separate batches. Laboratories used their own protocols to perform the analyses requested except for the contract laboratories which followed the DDRP protocol. Liquid audits were used in an effort to identify if interlaboratory differences were due to extraction procedures or chemical measurements. A component of the variability in the results was attributed to differences in the methods used such as soil/solution ratios, extractants or extraction procedures. The largest number of different methods used was for the measurement of cation exchange capacity. The results between the DDRP soil survey data and the study's results were compared using Youden-pair plots. In addition, standard statistical tests were performed. Overall, the DDRP data were comparable to the data from the study. However, out of the total 141 comparisons involving results from six or more laboratories, the results from the two contract laboratories did not meet the comparison criteria in 19 cases. Since there was never a case in which both contract laboratories failed, it would appear that the 19 cases which were not comparable were due to random analytical errors, incorrectly reported results, or misapplication of DDRP protocol

Linear and nonlinear verification of gyrokinetic microstability codes

Science.gov (United States)

Bravenec, R. V.; Candy, J.; Barnes, M.; Holland, C.

2011-12-01

Verification of nonlinear microstability codes is a necessary step before comparisons or predictions of turbulent transport in toroidal devices can be justified. By verification we mean demonstrating that a code correctly solves the mathematical model upon which it is based. Some degree of verification can be accomplished indirectly from analytical instability threshold conditions, nonlinear saturation estimates, etc., for relatively simple plasmas. However, verification for experimentally relevant plasma conditions and physics is beyond the realm of analytical treatment and must rely on code-to-code comparisons, i.e., benchmarking. The premise is that the codes are verified for a given problem or set of parameters if they all agree within a specified tolerance. True verification requires comparisons for a number of plasma conditions, e.g., different devices, discharges, times, and radii. Running the codes and keeping track of linear and nonlinear inputs and results for all conditions could be prohibitive unless there was some degree of automation. We have written software to do just this and have formulated a metric for assessing agreement of nonlinear simulations. We present comparisons, both linear and nonlinear, between the gyrokinetic codes GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and GS2 [W. Dorland, F. Jenko, M. Kotschenreuther, and B. N. Rogers, Phys. Rev. Lett. 85, 5579 (2000)]. We do so at the mid-radius for the same discharge as in earlier work [C. Holland, A. E. White, G. R. McKee, M. W. Shafer, J. Candy, R. E. Waltz, L. Schmitz, and G. R. Tynan, Phys. Plasmas 16, 052301 (2009)]. The comparisons include electromagnetic fluctuations, passing and trapped electrons, plasma shaping, one kinetic impurity, and finite Debye-length effects. Results neglecting and including electron collisions (Lorentz model) are presented. We find that the linear frequencies with or without collisions agree well between codes, as do the time averages of

Development of 2-d cfd code

International Nuclear Information System (INIS)

Mirza, S.A.

1999-01-01

In the present study, a two-dimensional computer code has been developed in FORTRAN using CFD technique, which is basically a numerical scheme. This computer code solves the Navier Stokes equations and continuity equation to find out the velocity and pressure fields within a given domain. This analysis has been done for the developed within a square cavity driven by the upper wall which has become a bench mark for testing and comparing the newly developed numerical schemes. Before to handle this task, different one-dimensional cases have been studied by CFD technique and their FORTRAN programs written. The cases studied are Couette flow, Poiseuille flow with and without using symmetric boundary condition. Finally a comparison between CFD results and analytical results has also been made. For the cavity flow the results from the developed code have been obtained for different Reynolds numbers which are finally presented in the form of velocity vectors. The comparison of the developed code results have been made with the results obtained from the share ware version of a commercially available code for Reynolds number of 10.0. The disagreement in the results quantitatively and qualitatively at some grid points of the calculation domain have been discussed and future recommendations in this regard have also been made. (author)

Comparison of CFD Natural Convection and Conduction-only Models for Heat Transfer in the Yucca Mountain Project Drifts

International Nuclear Information System (INIS)

Hadgu, T.; Webb, S.; Itamura, M.

2004-01-01

Yucca Mountain, Nevada has been designated as the nation's high-level radioactive waste repository and the U.S. Department of Energy has been approved to apply to the U.S. Nuclear Regulatory Commission for a license to construct a repository. Heat transfer in the Yucca Mountain Project (YMP) drift enclosures is an important aspect of repository waste emplacement. Canisters containing radioactive waste are to be emplaced in tunnels drilled 500 m below the ground surface. After repository closure, decaying heat is transferred from waste packages to the host rock by a combination of thermal radiation, natural convection and conduction heat transfer mechanism?. Current YMP mountain-scale and drift-scale numerical models often use a simplified porous medium code to model fluid and heat flow in the drift openings. To account for natural convection heat transfer, the thermal conductivity of the air was increased in the porous medium model. The equivalent thermal conductivity, defined as the ratio of total heat flow to conductive heat flow, used in the porous media models was based on horizontal concentric cylinders. Such modeling does not effectively capture turbulent natural convection in the open spaces as discussed by Webb et al. (2003) yet the approach is still widely used on the YMP project. In order to mechanistically model natural convection conditions in YMP drifts, the computational fluid dynamics (CFD) code FLUENT (Fluent, Incorporated, 2001) has been used to model natural convection heat transfer in the YMP emplacement drifts. A two-dimensional (2D) model representative of YMP geometry (e.g., includes waste package, drip shield, invert and drift wall) has been developed and numerical simulations made (Francis et al., 2003). Using CFD simulation results for both natural convection and conduction-only heat transfer in a single phase, single component fluid, equivalent thermal conductivities have been calculated for different Rayleigh numbers. Correlation

Design project of fast subcritical system 'Mala Lasta'; Idejno resenje brzog podkriticnog sistema Mala LASTA

Energy Technology Data Exchange (ETDEWEB)

Milosevic, M; Stefanovic, D; Popovic, D; Pesic, M; Zavaljevski, N; Nikolic, D; Arsenovic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1988-10-15

This report contains two parts. Part one covers the objective and fundamental elements for the choice of fast subcritical system 'Mala Lasta', review of the existing fast subcritical assemblies, and a description of the available domestic computer codes applied for calculating neutron reactor parameters. Comparison of results obtained by these codes for a number of existing subcritical assemblies was used for the choice of the design project described in part two of this report. It contains detailed description of the operating parameters of the chosen subcritical system based on the obtained calculated parameters.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

LIONS: a new set of Fortran 90 codes for the SPIRAL project at GANIL

International Nuclear Information System (INIS)

Bertrand, P.

1994-01-01

A set of new computer programs developed at GANIL is presented; these codes are used to study different parts of the SPIRAL project (a new radioactive ion beam facility), and particularly the dynamics in the CIME cyclotron, its injection inflector, and the new extraction system of the ECR ion sources. Three important modules are described: CHA3D for the evaluation of 3D electric fields with or without space charge effects, LIONS for the motion of ions and EXTRACT for the ECRIS extraction. These modules are written in Fortran 90 in a ''data parallel scheme''. They work either on UNIX workstations or parallel and vectorial computers. (author). 5 figs., 5 refs

Comparison of the two thermal-hydraulic codes Thyc and Vipre-02 on Vatican experiment

International Nuclear Information System (INIS)

Montat, D.; Maurel, F.; Olive, J.; Srikantiah, G.

1993-08-01

EDF's THYC and EPRI's VIPRE-02 3D thermalhydraulics computer codes are based on strongly different approaches (mixing against two-fluid representation for the two-phase flow, and porous media against subchannel approach for the rod bundle geometry description). In order to assess their efficiencies, they were both used to compute the two-phase flow behavior in an EDF experimental set-up, VATICAN. This set-up consisted of a 2.10 m high vertically oriented rod bundle of 40 heated rods in a 10 by 4 matrix (9.5 mm diameter and 12.6 mm square pitch). Refrigerant - 114 was introduced through the bottom left and right sides of the bundle and exited at the top. The lower 1.6 m height of the bundle was separated into two symmetric halves by a vertical wall. Desequilibrium in flow and quality between the two halves could be set, so that strong lateral mixing occurred above the divider wall. Comparison of codes computing and experimental data showed to some extent the superiority of a 6-equations model, but also highlighted the dramatic need for good constitutive relations concerning for instance the turbulent mixing. Some systematic deviations from experimental data were detected, linked to poor accuracy of some chosen closure laws. (authors). 6 figs., 3 tabs., 5 refs

The NIMROD Code

Science.gov (United States)

Schnack, D. D.; Glasser, A. H.

1996-11-01

NIMROD is a new code system that is being developed for the analysis of modern fusion experiments. It is being designed from the beginning to make the maximum use of massively parallel computer architectures and computer graphics. The NIMROD physics kernel solves the three-dimensional, time-dependent two-fluid equations with neo-classical effects in toroidal geometry of arbitrary poloidal cross section. The NIMROD system also includes a pre-processor, a grid generator, and a post processor. User interaction with NIMROD is facilitated by a modern graphical user interface (GUI). The NIMROD project is using Quality Function Deployment (QFD) team management techniques to minimize re-engineering and reduce code development time. This paper gives an overview of the NIMROD project. Operation of the GUI is demonstrated, and the first results from the physics kernel are given.

Comparison of strongly heat-driven flow codes for unsaturated media

International Nuclear Information System (INIS)

Updegraff, C.D.

1989-08-01

Under the sponsorship of the US Nuclear Regulatory Commission, Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal of high-level radioactive waste (HLW) in unsaturated welded tuff. As part of this effort, SNL evaluated existing strongly heat-driven flow computer codes for simulating ground-water flow in unsaturated media. The three codes tested, NORIA, PETROS, and TOUGH, were compared against a suite of problems for which analytical and numerical solutions or experimental results exist. The problems were selected to test the abilities of the codes to simulate situations ranging from simple, uncoupled processes, such as two-phase flow or heat transfer, to fully coupled processes, such as vaporization caused by high temperatures. In general, all three codes were found to be difficult to use because of (1) built-in time stepping criteria, (2) the treatment of boundary conditions, and (3) handling of evaporation/condensation problems. A drawback of the study was that adequate problems related to expected repository conditions were not available in the literature. Nevertheless, the results of this study suggest the need for thorough investigations of the impact of heat on the flow field in the vicinity of an unsaturated HLW repository. Recommendations are to develop a new flow code combining the best features of these three codes and eliminating the worst ones. 19 refs., 49 figs

Neutronic evolution of SENA reactor during the first and second cycles. Comparison between the experimental power distributions obtained from the in-core instrumentation evaluation code CIRCE and the theoretical power values computed with the two-dimensional diffusion-evolution code EVOE

International Nuclear Information System (INIS)

Andrieux, Chantal

1976-03-01

The neutronic evolution of the reacteur Sena during the first and second cycles is presented. The experimental power distributions, obtained from the in-core instrumentation evaluation code CIRCE are compared with the theoretical powers calculated with the two-dimensional diffusion-evolution code EVOE. The CIRCE code allows: the study of the evolution of the principal parameters of the core, the comparison of the results of measured and theoretical estimates. Therefore this study has a great interest for the knowledge of the neutronic evolution of the core, as well as the validation of the refinement of theoretical estimation methods. The core calculation methods and requisite data for the evaluation of the measurements are presented after a brief description of the SENA core and its inner instrumentation. The principle of the in-core instrumentation evaluation code CIRCE, and calculation of the experimental power distributions and nuclear core parameters are then exposed. The results of the evaluation are discussed, with a comparison of the theoretical and experimental results. Taking account of the approximations used, these results, as far as the first and second cycles at SENA are concerned, are satisfactory, the deviations between theoretical and experimental power distributions being lower than 3% at the middle of the reactor and 9% at the periphery [fr

Adaption, validation and application of advanced codes with 3-dimensional neutron kinetics for accident analysis calculations - STC with Bulgaria

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.; Panayotov, D.; Ilieva, B.

2001-08-01

In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.) [de

Leadership Class Configuration Interaction Code - Status and Opportunities

Science.gov (United States)

Vary, James

2011-10-01

With support from SciDAC-UNEDF (www.unedf.org) nuclear theorists have developed and are continuously improving a Leadership Class Configuration Interaction Code (LCCI) for forefront nuclear structure calculations. The aim of this project is to make state-of-the-art nuclear structure tools available to the entire community of researchers including graduate students. The project includes codes such as NuShellX, MFDn and BIGSTICK that run a range of computers from laptops to leadership class supercomputers. Codes, scripts, test cases and documentation have been assembled, are under continuous development and are scheduled for release to the entire research community in November 2011. A covering script that accesses the appropriate code and supporting files is under development. In addition, a Data Base Management System (DBMS) that records key information from large production runs and archived results of those runs has been developed (http://nuclear.physics.iastate.edu/info/) and will be released. Following an outline of the project, the code structure, capabilities, the DBMS and current efforts, I will suggest a path forward that would benefit greatly from a significant partnership between researchers who use the codes, code developers and the National Nuclear Data efforts. This research is supported in part by DOE under grant DE-FG02-87ER40371 and grant DE-FC02-09ER41582 (SciDAC-UNEDF).

SUMMARY OF GENERAL WORKING GROUP A+B+D: CODES BENCHMARKING.

Energy Technology Data Exchange (ETDEWEB)

WEI, J.; SHAPOSHNIKOVA, E.; ZIMMERMANN, F.; HOFMANN, I.

2006-05-29

Computer simulation is an indispensable tool in assisting the design, construction, and operation of accelerators. In particular, computer simulation complements analytical theories and experimental observations in understanding beam dynamics in accelerators. The ultimate function of computer simulation is to study mechanisms that limit the performance of frontier accelerators. There are four goals for the benchmarking of computer simulation codes, namely debugging, validation, comparison and verification: (1) Debugging--codes should calculate what they are supposed to calculate; (2) Validation--results generated by the codes should agree with established analytical results for specific cases; (3) Comparison--results from two sets of codes should agree with each other if the models used are the same; and (4) Verification--results from the codes should agree with experimental measurements. This is the summary of the joint session among working groups A, B, and D of the HI32006 Workshop on computer codes benchmarking.

A comparison of thermal algorithms of fuel rod performance code systems

International Nuclear Information System (INIS)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

A comparison of thermal algorithms of fuel rod performance code systems

Energy Technology Data Exchange (ETDEWEB)

Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

2003-11-01

The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

Verification of HYDRASTAR - A code for stochastic continuum simulation of groundwater flow

International Nuclear Information System (INIS)

Norman, S.

1991-07-01

HYDRASTAR is a code developed at Starprog AB for use in the SKB 91 performance assessment project with the following principal function: - Reads the actual conductivity measurements from a file created from the data base GEOTAB. - Regularizes the measurements to a user chosen calculation scale. - Generates three dimensional unconditional realizations of the conductivity field by using a supplied model of the conductivity field as a stochastic function. - Conditions the simulated conductivity field on the actual regularized measurements. - Reads the boundary conditions from a regional deterministic NAMMU computation. - Calculates the hydraulic head field, Darcy velocity field, stream lines and water travel times by solving the stationary hydrology equation and the streamline equation obtained with the velocities calculated from Darcy's law. - Generates visualizations of the realizations if desired. - Calculates statistics such as semivariograms and expectation values of the output fields by repeating the above procedure by iterations of the Monte Carlo type. When using computer codes for safety assessment purpose validation and verification of the codes are important. Thus this report describes a work performed with the goal of verifying parts of HYDRASTAR. The verification described in this report uses comparisons with two other solutions of related examples: A. Comparison with a so called perturbation solution of the stochastical stationary hydrology equation. This as an analytical approximation of the stochastical stationary hydrology equation valid in the case of small variability of the unconditional random conductivity field. B. Comparison with the (Hydrocoin, 1988), case 2. This is a classical example of a hydrology problem with a deterministic conductivity field. The principal feature of the problem is the presence of narrow fracture zones with high conductivity. the compared output are the hydraulic head field and a number of stream lines originating from a

MODLIB, library of Fortran modules for nuclear reaction codes

International Nuclear Information System (INIS)

Talou, Patrick

2006-01-01

1 - Description of program or function: ModLib is a library of Fortran (90-compatible) modules to be used in existing and future nuclear reaction codes. The development of the library is an international effort being undertaken under the auspices of the long-term Subgroup A of the OECD/NEA Working Party on Evaluation and Cooperation. The aim is to constitute a library of well-tested and well-documented pieces of codes that can be used with confidence in all our coding efforts. This effort will undoubtedly help avoid the duplication of work, and most certainly facilitate the very important inter-comparisons between existing codes. 2 - Methods: - Width f luctuations [Talou, Chadwick]: calculates width fluctuation correction factors (output) for a set of transmission coefficients (input). Three methods are available: HRTW, Moldauer, and Verbaarschot (also called GOE approach). So far, no distinction is made according to the type of the coefficients channels (particle emission, gamma-ray emission, fission). - Gamma s trength [Herman]: calculates gamma-ray transmission coefficients using a Giant Resonance formalism. - Level d ensity [Koning]: computes the Gilbert-Cameron-Ignatyuk formalism for the continuum nuclear level density. - CHECKR, FIZCON, INTER, PSYCHE, STANEF [Dunford]: these modules are used in the MODLIB project but are not included in this package. They are available from the NEA Data Bank Computer Program Service under Package Ids: CHECKR (USCD1208), FIZCON (USCD1209), INTER (USCD1212), PSYCHE (USCD1216), STANEF (USCD1218)

Low-temperature plasma simulations with the LSP PIC code

Science.gov (United States)

Carlsson, Johan; Khrabrov, Alex; Kaganovich, Igor; Keating, David; Selezneva, Svetlana; Sommerer, Timothy

2014-10-01

The LSP (Large-Scale Plasma) PIC-MCC code has been used to simulate several low-temperature plasma configurations, including a gas switch for high-power AC/DC conversion, a glow discharge and a Hall thruster. Simulation results will be presented with an emphasis on code comparison and validation against experiment. High-voltage, direct-current (HVDC) power transmission is becoming more common as it can reduce construction costs and power losses. Solid-state power-electronics devices are presently used, but it has been proposed that gas switches could become a compact, less costly, alternative. A gas-switch conversion device would be based on a glow discharge, with a magnetically insulated cold cathode. Its operation is similar to that of a sputtering magnetron, but with much higher pressure (0.1 to 0.3 Torr) in order to achieve high current density. We have performed 1D (axial) and 2D (axial/radial) simulations of such a gas switch using LSP. The 1D results were compared with results from the EDIPIC code. To test and compare the collision models used by the LSP and EDIPIC codes in more detail, a validation exercise was performed for the cathode fall of a glow discharge. We will also present some 2D (radial/azimuthal) LSP simulations of a Hall thruster. The information, data, or work presented herein was funded in part by the Advanced Research Projects Agency-Energy (ARPA-E), U.S. Department of Energy, under Award Number DE-AR0000298.

Code of accounts. Management overview volume: Environmental restoration

International Nuclear Information System (INIS)

Fox, M.B.; Birkholz, H.L.

1997-10-01

The purpose of this procedure is to provide the requirement for assigning cost collection codes and the structure of these codes for all costs incurred for the Environmental Restoration Contract. The coding structure will be used in the budgeting and control of project costs

Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations

International Nuclear Information System (INIS)

Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I.

2015-01-01

A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.

On the automated assessment of nuclear reactor systems code accuracy

International Nuclear Information System (INIS)

Kunz, Robert F.; Kasmala, Gerald F.; Mahaffy, John H.; Murray, Christopher J.

2002-01-01

An automated code assessment program (ACAP) has been developed to provide quantitative comparisons between nuclear reactor systems (NRS) code results and experimental measurements. The tool provides a suite of metrics for quality of fit to specific data sets, and the means to produce one or more figures of merit (FOM) for a code, based on weighted averages of results from the batch execution of a large number of code-experiment and code-code data comparisons. Accordingly, this tool has the potential to significantly streamline the verification and validation (V and V) processes in NRS code development environments which are characterized by rapidly evolving software, many contributing developers and a large and growing body of validation data. In this paper, a survey of data conditioning and analysis techniques is summarized which focuses on their relevance to NRS code accuracy assessment. A number of methods are considered for their applicability to the automated assessment of the accuracy of NRS code simulations. A variety of data types and computational modeling methods are considered from a spectrum of mathematical and engineering disciplines. The goal of the survey was to identify needs, issues and techniques to be considered in the development of an automated code assessment procedure, to be used in United States Nuclear Regulatory Commission (NRC) advanced thermal-hydraulic T/H code consolidation efforts. The ACAP software was designed based in large measure on the findings of this survey. An overview of this tool is summarized and several NRS data applications are provided. The paper is organized as follows: The motivation for this work is first provided by background discussion that summarizes the relevance of this subject matter to the nuclear reactor industry. Next, the spectrum of NRS data types are classified into categories, in order to provide a basis for assessing individual comparison methods. Then, a summary of the survey is provided, where each

Final Technical Report: Hydrogen Codes and Standards Outreach

Energy Technology Data Exchange (ETDEWEB)

Hall, Karen I.

2007-05-12

This project contributed significantly to the development of new codes and standards, both domestically and internationally. The NHA collaborated with codes and standards development organizations to identify technical areas of expertise that would be required to produce the codes and standards that industry and DOE felt were required to facilitate commercialization of hydrogen and fuel cell technologies and infrastructure. NHA staff participated directly in technical committees and working groups where issues could be discussed with the appropriate industry groups. In other cases, the NHA recommended specific industry experts to serve on technical committees and working groups where the need for this specific industry expertise would be on-going, and where this approach was likely to contribute to timely completion of the effort. The project also facilitated dialog between codes and standards development organizations, hydrogen and fuel cell experts, the government and national labs, researchers, code officials, industry associations, as well as the public regarding the timeframes for needed codes and standards, industry consensus on technical issues, procedures for implementing changes, and general principles of hydrogen safety. The project facilitated hands-on learning, as participants in several NHA workshops and technical meetings were able to experience hydrogen vehicles, witness hydrogen refueling demonstrations, see metal hydride storage cartridges in operation, and view other hydrogen energy products.

Input data required for specific performance assessment codes

International Nuclear Information System (INIS)

Seitz, R.R.; Garcia, R.S.; Starmer, R.J.; Dicke, C.A.; Leonard, P.R.; Maheras, S.J.; Rood, A.S.; Smith, R.W.

1992-02-01

The Department of Energy's National Low-Level Waste Management Program at the Idaho National Engineering Laboratory generated this report on input data requirements for computer codes to assist States and compacts in their performance assessments. This report gives generators, developers, operators, and users some guidelines on what input data is required to satisfy 22 common performance assessment codes. Each of the codes is summarized and a matrix table is provided to allow comparison of the various input required by the codes. This report does not determine or recommend which codes are preferable

The 1992 ENDF Pre-processing codes

International Nuclear Information System (INIS)

Cullen, D.E.

1992-01-01

This document summarizes the 1992 version of the ENDF pre-processing codes which are required for processing evaluated nuclear data coded in the format ENDF-4, ENDF-5, or ENDF-6. Included are the codes CONVERT, MERGER, LINEAR, RECENT, SIGMA1, LEGEND, FIXUP, GROUPIE, DICTION, MIXER, VIRGIN, COMPLOT, EVALPLOT, RELABEL. Some of the functions of these codes are: to calculate cross-sections from resonance parameters; to calculate angular distributions, group average, mixtures of cross-sections, etc; to produce graphical plottings and data comparisons. The codes are designed to operate on virtually any type of computer including PC's. They are available from the IAEA Nuclear Data Section, free of charge upon request, on magnetic tape or a set of HD diskettes. (author)

Thick target benchmark test for the code used in the design of high intensity proton accelerator project

International Nuclear Information System (INIS)

Meigo, Shin-ichiro; Harada, Masatoshi

2003-01-01

In the neutronics design for the JAERI and KEK Joint high intensity accelerator facilities, transport codes of NMTC/JAM, MCNPX and MARS are used. In order to confirm the predict ability for these code, it is important to compare with the experiment result. For the validation of the source term of neutron, the calculations are compared with the experimental spectrum of neutrons produced from thick target, which are carried out at LANL and KEK. As for validation of low energy incident case, the calculations are compared with experiment carried out at LANL, in which target of C, Al, Fe, and 238 U are irradiated with 256-MeV protons. By the comparison, it is found that both NMTC/JAM and MCNPX show good agreement with the experiment within by a factor of 2. MARS shows good agreement for C and Al target. MARS, however, gives rather underestimation for all targets in the neutron energy region higher than 30 MeV. For the validation high incident energy case, the codes are compared with the experiment carried out at KEK. In this experiment, W and Pb targets are bombarded with 0.5- and 1.5-GeV protons. Although slightly disagreement exists, NMTC/JAM, MCNPX and MARS are in good agreement with the experiment within by a factor of 2. (author)

The LIONS code (version 1.0)

International Nuclear Information System (INIS)

Bertrand, P.

1993-01-01

The new LIONS code (Lancement d'IONS or Ion Launching), a dynamical code implemented in the SPIRaL project for the CIME cyclotron studies, is presented. The various software involves a 3D magnetostatic code, 2D or 3D electrostatic codes for generation of realistic field maps, and several dynamical codes for studying the behaviour of the reference particle from the cyclotron center up to the ejection and for launching particles packets complying with given correlations. Its interactions with the other codes are described. The LIONS code, written in Fortran 90 is already used in studying the CIME cyclotron, from the center to the ejection. It is designed to be used, with minor modifications, in other contexts such as for the simulation of mass spectrometer facilities

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

Energy Technology Data Exchange (ETDEWEB)

Paolo Balestra; Carlo Parisi; Andrea Alfonsi

2016-02-01

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.

Verification of 3-D generation code package for neutronic calculations of WWERs

International Nuclear Information System (INIS)

Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.

2000-01-01

Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)

Modeling of FREYA fast critical experiments with the Serpent Monte Carlo code

International Nuclear Information System (INIS)

Fridman, E.; Kochetkov, A.; Krása, A.

2017-01-01

Highlights: • FREYA – the EURATOM project executed to support fast lead-based reactor systems. • Critical experiments in the VENUS-F facility during the FREYA project. • Characterization of the critical VENUS-F cores with Serpent. • Comparison of the numerical Serpent results to the experimental data. - Abstract: The FP7 EURATOM project FREYA has been executed between 2011 and 2016 with the aim of supporting the design of fast lead-cooled reactor systems such as MYRRHA and ALFRED. During the project, a number of critical experiments were conducted in the VENUS-F facility located at SCK·CEN, Mol, Belgium. The Monte Carlo code Serpent was one of the codes applied for the characterization of the critical VENUS-F cores. Four critical configurations were modeled with Serpent, namely the reference critical core, the clean MYRRHA mock-up, the full MYRRHA mock-up, and the critical core with the ALFRED island. This paper briefly presents the VENUS-F facility, provides a detailed description of the aforementioned critical VENUS-F cores, and compares the numerical results calculated by Serpent to the available experimental data. The compared parameters include keff, point kinetics parameters, fission rate ratios of important actinides to that of U235 (spectral indices), axial and radial distribution of fission rates, and lead void reactivity effect. The reported results show generally good agreement between the calculated and experimental values. Nevertheless, the paper also reveals some noteworthy issues requiring further attention. This includes the systematic overprediction of reactivity and systematic underestimation of the U238 to U235 fission rate ratio.

Benchmarking studies for the DESCARTES and CIDER codes

International Nuclear Information System (INIS)

Eslinger, P.W.; Ouderkirk, S.J.; Nichols, W.E.

1993-01-01

The Hanford Envirorunental Dose Reconstruction (HEDR) project is developing several computer codes to model the airborne release, transport, and envirormental accumulation of radionuclides resulting from Hanford operations from 1944 through 1972. In order to calculate the dose of radiation a person may have received in any given location, the geographic area addressed by the HEDR Project will be divided into a grid. The grid size suggested by the draft requirements contains 2091 units called nodes. Two of the codes being developed are DESCARTES and CIDER. The DESCARTES code will be used to estimate the concentration of radionuclides in environmental pathways from the output of the air transport code RATCHET. The CIDER code will use information provided by DESCARTES to estimate the dose received by an individual. The requirements that Battelle (BNW) set for these two codes were released to the HEDR Technical Steering Panel (TSP) in a draft document on November 10, 1992. This document reports on the preliminary work performed by the code development team to determine if the requirements could be met

Contributions to the validation of the ASTEC V1 code

International Nuclear Information System (INIS)

Constantin, Marin; Rizoiu, Andrei; Turcu, Ilie

2004-01-01

In the frame of PHEBEN2 project (Validation of the severe accidents codes for applications to nuclear power plants, based on the PHEBUS FP experiments), a project developed within the EU research Frame Program 5 (FP5), the INR-Pitesti's team has received the task of determining the ASTEC code sensitivity. The PHEBEN2 project has been initiated in 1998 and gathered 13 partners from 6 EU member states. To the project 4 partners from 3 candidate states (Hungary, Bulgaria and Romania) joined later. The works were contracted with the European Commission (under FIKS-CT1999-00009 contract) that supports financially the research effort up to about 50%. According to the contract provisions, INR's team participated in developing the Working Package 1 (WP1) which refers to validation of the integral computation codes that use the PHOEBUS experimental data and the Working Package 3 (WP3) referring to the evaluation of the codes to be applied in nuclear power plants for risk evaluation, nuclear safety margin evaluation and determination/evaluation of the measures to be adopted in case of severe accident. The present work continues the efforts to validate preliminarily the ASTEC code. Focused are the the stand-alone sensitivity analyses applied to two most important modules of the code, namely DIVA and SOPHAEROS

Timing comparison of two-dimensional discrete-ordinates codes for criticality calculations

International Nuclear Information System (INIS)

Miller, W.F. Jr.; Alcouffe, R.E.; Bosler, G.E.; Brinkley, F.W. Jr.; O'dell, R.D.

1979-01-01

The authors compare two-dimensional discrete-ordinates neutron transport computer codes to solve reactor criticality problems. The fundamental interest is in determining which code requires the minimum Central Processing Unit (CPU) time for a given numerical model of a reasonably realistic fast reactor core and peripherals. The computer codes considered are the most advanced available and, in three cases, are not officially released. The conclusion, based on the study of four fast reactor core models, is that for this class of problems the diffusion synthetic accelerated version of TWOTRAN, labeled TWOTRAN-DA, is superior to the other codes in terms of CPU requirements

Spread-spectrum communication using binary spatiotemporal chaotic codes

International Nuclear Information System (INIS)

Wang Xingang; Zhan Meng; Gong Xiaofeng; Lai, C.H.; Lai, Y.-C.

2005-01-01

We propose a scheme to generate binary code for baseband spread-spectrum communication by using a chain of coupled chaotic maps. We compare the performances of this type of spatiotemporal chaotic code with those of a conventional code used frequently in digital communication, the Gold code, and demonstrate that our code is comparable or even superior to the Gold code in several key aspects: security, bit error rate, code generation speed, and the number of possible code sequences. As the field of communicating with chaos faces doubts in terms of performance comparison with conventional digital communication schemes, our work gives a clear message that communicating with chaos can be advantageous and it deserves further attention from the nonlinear science community

CONSUL code package application for LMFR core calculations

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

Case file coding of child maltreatment: Methods, challenges, and innovations in a longitudinal project of youth in foster care☆

Science.gov (United States)

Huffhines, Lindsay; Tunno, Angela M.; Cho, Bridget; Hambrick, Erin P.; Campos, Ilse; Lichty, Brittany; Jackson, Yo

2016-01-01

State social service agency case files are a common mechanism for obtaining information about a child’s maltreatment history, yet these documents are often challenging for researchers to access, and then to process in a manner consistent with the requirements of social science research designs. Specifically, accessing and navigating case files is an extensive undertaking, and a task that many researchers have had to maneuver with little guidance. Even after the files are in hand and the research questions and relevant variables have been clarified, case file information about a child’s maltreatment exposure can be idiosyncratic, vague, inconsistent, and incomplete, making coding such information into useful variables for statistical analyses difficult. The Modified Maltreatment Classification System (MMCS) is a popular tool used to guide the process, and though comprehensive, this coding system cannot cover all idiosyncrasies found in case files. It is not clear from the literature how researchers implement this system while accounting for issues outside of the purview of the MMCS or that arise during MMCS use. Finally, a large yet reliable file coding team is essential to the process, however, the literature lacks training guidelines and methods for establishing reliability between coders. In an effort to move the field toward a common approach, the purpose of the present discussion is to detail the process used by one large-scale study of child maltreatment, the Studying Pathways to Adjustment and Resilience in Kids (SPARK) project, a longitudinal study of resilience in youth in foster care. The article addresses each phase of case file coding, from accessing case files, to identifying how to measure constructs of interest, to dealing with exceptions to the coding system, to coding variables reliably, to training large teams of coders and monitoring for fidelity. Implications for a comprehensive and efficient approach to case file coding are discussed. PMID

SeisCode: A seismological software repository for discovery and collaboration

Science.gov (United States)

Trabant, C.; Reyes, C. G.; Clark, A.; Karstens, R.

2012-12-01

SeisCode is a community repository for software used in seismological and related fields. The repository is intended to increase discoverability of such software and to provide a long-term home for software projects. Other places exist where seismological software may be found, but none meet the requirements necessary for an always current, easy to search, well documented, and citable resource for projects. Organizations such as IRIS, ORFEUS, and the USGS have websites with lists of available or contributed seismological software. Since the authors themselves do often not maintain these lists, the documentation often consists of a sentence or paragraph, and the available software may be outdated. Repositories such as GoogleCode and SourceForge, which are directly maintained by the authors, provide version control and issue tracking but do not provide a unified way of locating geophysical software scattered in and among countless unrelated projects. Additionally, projects are hosted at language-specific sites such as Mathworks and PyPI, in FTP directories, and in websites strewn across the Web. Search engines are only partially effective discovery tools, as the desired software is often hidden deep within the results. SeisCode provides software authors a place to present their software, codes, scripts, tutorials, and examples to the seismological community. Authors can choose their own level of involvement. At one end of the spectrum, the author might simply create a web page that points to an existing site. At the other extreme, an author may choose to leverage the many tools provided by SeisCode, such as a source code management tool with integrated issue tracking, forums, news feeds, downloads, wikis, and more. For software development projects with multiple authors, SeisCode can also be used as a central site for collaboration. SeisCode provides the community with an easy way to discover software, while providing authors a way to build a community around their

Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

International Nuclear Information System (INIS)

Brus, N.A.; Ioussoupov, O.E.

2000-01-01

This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

Writing the Live Coding Book

DEFF Research Database (Denmark)

Blackwell, Alan; Cox, Geoff; Lee, Sang Wong

2016-01-01

This paper is a speculation on the relationship between coding and writing, and the ways in which technical innovations and capabilities enable us to rethink each in terms of the other. As a case study, we draw on recent experiences of preparing a book on live coding, which integrates a wide range...... of personal, historical, technical and critical perspectives. This book project has been both experimental and reflective, in a manner that allows us to draw on critical understanding of both code and writing, and point to the potential for new practices in the future....

Comparison of three gamma ray isotopic determination codes: FRAM, MGA, and TRIFID

International Nuclear Information System (INIS)

Cremers, T.L.; Malcom, J.E.; Bonner, C.A.

1994-01-01

The determination of the isotopic distribution of plutonium and the americium concentration is required for the assay of nuclear material by calorimetry or neutron coincidence counting. The isotopic information is used in calorimetric assay to compute the effective specific power from the measured isotopic fractions and the known specific power of each isotope. The effective specific power is combined with the heat measurement to obtain the mass of plutonium in the assayed nuclear material. The response of neutron coincidence counters is determined by the 240 Pu isotopic fraction with contributions from the other even plutonium isotopes. The effect of the 240 Pu isotopic fraction and the other neutron contributing isotopes are combined as 240 Pu effective. This is used to calculate the mass of nuclear material from the neutron counting data in a manner analogous to the effective specific power in calorimeter. Comparisons of the precision and accuracy of calorimetric assay and neutron coincidence counting often focus only on the precision and accuracy of the heat measurement (calorimetry) compared to the precision and accuracy of the neutron coincidence counting statistics. The major source of uncertainty for both calorimetric assay and neutron coincidence counting often lies in the determination of the plutonium isotopic distribution ad determined by gamma ray spectroscopy. Thus, the selection of the appropriate isotopic distribution code is of paramount importance to good calorimetric assay and neutron coincidence counting. Three gamma ray isotopic distribution codes, FRAM, MGA, and TRIFID have been compared at the Los Alamos Plutonium Facility under carefully controlled conditions of similar count rates, count times, and 240 Pu isotopic fraction

Short-Block Protograph-Based LDPC Codes

Science.gov (United States)

Divsalar, Dariush; Dolinar, Samuel; Jones, Christopher

2010-01-01

Short-block low-density parity-check (LDPC) codes of a special type are intended to be especially well suited for potential applications that include transmission of command and control data, cellular telephony, data communications in wireless local area networks, and satellite data communications. [In general, LDPC codes belong to a class of error-correcting codes suitable for use in a variety of wireless data-communication systems that include noisy channels.] The codes of the present special type exhibit low error floors, low bit and frame error rates, and low latency (in comparison with related prior codes). These codes also achieve low maximum rate of undetected errors over all signal-to-noise ratios, without requiring the use of cyclic redundancy checks, which would significantly increase the overhead for short blocks. These codes have protograph representations; this is advantageous in that, for reasons that exceed the scope of this article, the applicability of protograph representations makes it possible to design highspeed iterative decoders that utilize belief- propagation algorithms.

Design evaluation on sodium piping system and comparison of the design codes

International Nuclear Information System (INIS)

Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon

2015-01-01

A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

Design evaluation on sodium piping system and comparison of the design codes

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon [KAERI, Daejeon (Korea, Republic of)

2015-03-15

A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

The analysis of thermal-hydraulic models in MELCOR code

Energy Technology Data Exchange (ETDEWEB)

Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

1996-07-15

The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.

The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

Directory of Open Access Journals (Sweden)

Surian Pinem

2014-01-01

Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents. (Modelling of steam condensation on the particles)

International Nuclear Information System (INIS)

Bunz, H.; Dunbar, L.H.; Fermandjian, J.; Lhiaubet, G.

1987-11-01

An aerosol code comparison exercise was performed within the framework of the Commission of European Communities (Division of Safety of Nuclear Installations). This exercise, focused on the process of steam condensation onto the aerosols occurring in PWR containment buildings during severe core damage accidents, has allowed to understand the discrepancies between the results obtained. These discrepancies are due, in particular, to whether the curvature effect is modelled or not in the codes

Implementation of a 3D halo neutral model in the TRANSP code and application to projected NSTX-U plasmas

Science.gov (United States)

Medley, S. S.; Liu, D.; Gorelenkova, M. V.; Heidbrink, W. W.; Stagner, L.

2016-02-01

A 3D halo neutral code developed at the Princeton Plasma Physics Laboratory and implemented for analysis using the TRANSP code is applied to projected National Spherical Torus eXperiment-Upgrade (NSTX-U plasmas). The legacy TRANSP code did not handle halo neutrals properly since they were distributed over the plasma volume rather than remaining in the vicinity of the neutral beam footprint as is actually the case. The 3D halo neutral code uses a ‘beam-in-a-box’ model that encompasses both injected beam neutrals and resulting halo neutrals. Upon deposition by charge exchange, a subset of the full, one-half and one-third beam energy components produce first generation halo neutrals that are tracked through successive generations until an ionization event occurs or the descendant halos exit the box. The 3D halo neutral model and neutral particle analyzer (NPA) simulator in the TRANSP code have been benchmarked with the Fast-Ion D-Alpha simulation (FIDAsim) code, which provides Monte Carlo simulations of beam neutral injection, attenuation, halo generation, halo spatial diffusion, and photoemission processes. When using the same atomic physics database, TRANSP and FIDAsim simulations achieve excellent agreement on the spatial profile and magnitude of beam and halo neutral densities and the NPA energy spectrum. The simulations show that the halo neutral density can be comparable to the beam neutral density. These halo neutrals can double the NPA flux, but they have minor effects on the NPA energy spectrum shape. The TRANSP and FIDAsim simulations also suggest that the magnitudes of beam and halo neutral densities are relatively sensitive to the choice of the atomic physics databases.

Implementation of a 3D halo neutral model in the TRANSP code and application to projected NSTX-U plasmas

Energy Technology Data Exchange (ETDEWEB)

Medley, S. S. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Liu, D. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Univ. of California, Irvine, CA (United States). Dept. of Physics and Astronomy; Gorelenkova, M. V. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Heidbrink, W. W. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Univ. of California, Irvine, CA (United States). Dept. of Physics and Astronomy; Stagner, L. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Univ. of California, Irvine, CA (United States). Dept. of Physics and Astronomy

2016-01-12

A 3D halo neutral code developed at the Princeton Plasma Physics Laboratory and implemented for analysis using the TRANSP code is applied to projected National Spherical Torus eXperiment-Upgrade (NSTX-U plasmas). The legacy TRANSP code did not handle halo neutrals properly since they were distributed over the plasma volume rather than remaining in the vicinity of the neutral beam footprint as is actually the case. The 3D halo neutral code uses a 'beam-in-a-box' model that encompasses both injected beam neutrals and resulting halo neutrals. Upon deposition by charge exchange, a subset of the full, one-half and one-third beam energy components produce first generation halo neutrals that are tracked through successive generations until an ionization event occurs or the descendant halos exit the box. The 3D halo neutral model and neutral particle analyzer (NPA) simulator in the TRANSP code have been benchmarked with the Fast-Ion D-Alpha simulation (FIDAsim) code, which provides Monte Carlo simulations of beam neutral injection, attenuation, halo generation, halo spatial diffusion, and photoemission processes. When using the same atomic physics database, TRANSP and FIDAsim simulations achieve excellent agreement on the spatial profile and magnitude of beam and halo neutral densities and the NPA energy spectrum. The simulations show that the halo neutral density can be comparable to the beam neutral density. These halo neutrals can double the NPA flux, but they have minor effects on the NPA energy spectrum shape. The TRANSP and FIDAsim simulations also suggest that the magnitudes of beam and halo neutral densities are relatively sensitive to the choice of the atomic physics databases.

Comparison of two-dimensional and three-dimensional MHD equilibrium and stability codes

International Nuclear Information System (INIS)

Herrnegger, F.; Merkel, P.; Johnson, J.L.

1986-02-01

Stability results obtained with the fully three-dimensional magnetohydrodynamic code BETA, the helically invariant code HERA, and the asymptotic stellarator expansion code STEP agree well for a straight l = 2, M = 5 stellarator model. This good agreement between the BETA and STEP codes persists as toroidal curvature is introduced. This validation provides justification for confidence in work with these models. 20 refs., 11 figs

Coding training for medical students: How good is diagnoses coding with ICD-10 by novices?

Directory of Open Access Journals (Sweden)

Stausberg, Jürgen

2005-04-01

Full Text Available Teaching of knowledge and competence in documentation and coding is an essential part of medical education. Therefore, coding training had been placed within the course of epidemiology, medical biometry, and medical informatics. From this, we can draw conclusions about the quality of coding by novices. One hundred and eighteen students coded diagnoses from 15 nephrological cases in homework. In addition to interrater reliability, validity was calculated by comparison with a reference coding. On the level of terminal codes, 59.3% of the students' results were correct. The completeness was calculated as 58.0%. The results on the chapter level increased up to 91.5% and 87.7% respectively. For the calculation of reliability a new, simple measure was developed that leads to values of 0.46 on the level of terminal codes and 0.87 on the chapter level for interrater reliability. The figures of concordance with the reference coding are quite similar. In contrary, routine data show considerably lower results with 0.34 and 0.63 respectively. Interrater reliability and validity of coding by novices is as good as coding by experts. The missing advantage of experts could be explained by the workload of documentation and a negative attitude to coding on the one hand. On the other hand, coding in a DRG-system is handicapped by a large number of detailed coding rules, which do not end in uniform results but rather lead to wrong and random codes. Anyway, students left the course well prepared for coding.

Linear codes associated to determinantal varieties

DEFF Research Database (Denmark)

Beelen, Peter; Ghorpade, Sudhir R.; Hasan, Sartaj Ul

2015-01-01

We consider a class of linear codes associated to projective algebraic varieties defined by the vanishing of minors of a fixed size of a generic matrix. It is seen that the resulting code has only a small number of distinct weights. The case of varieties defined by the vanishing of 2×2 minors is ...

Plaspp: A New X-Ray Postprocessing Capability for ASCI Codes

International Nuclear Information System (INIS)

Pollak, Gregory

2003-01-01

This report announces the availability of the beta version of a (partly) new code, Plaspp (Plasma Postprocessor). This code postprocesses (graphics) dumps from at least two ASCI code suites: Crestone Project and Shavano Project. The basic structure of the code follows that of TDG, the equivalent postprocessor code for LASNEX. In addition to some new commands, the basic differences between TDG and Plaspp are the following: Plaspp uses a graphics dump instead of the unique TDG dump, it handles the unstructured meshes that the ASCI codes produce, and it can use its own multigroup opacity data. Because of the dump format, this code should be useable by any code that produces Cartesian, cylindrical, or spherical graphics formats. This report details the new commands; the required information to be placed on the dumps; some new commands and edits that are applicable to TDG as well, but have not been documented elsewhere; and general information about execution on the open and secure networks.

The reason for having a code of pharmaceutical ethics: Spanish Pharmacists Code of Ethics

Directory of Open Access Journals (Sweden)

Ana Mulet Alberola

2017-05-01

Full Text Available The pharmacist profession needs its own code of conduct set out in writing to serve as a stimulus to pharmacists in their day-to-day work in the different areas of pharmacy, in conjunction always with each individual pharmacist´s personal commitment to their patients, to other healthcare professionals and to society. An overview is provided of the different codes of ethics for pharmacists on the national and international scale, the most up-to-date code for 2015 being presented as a set of principles which must guide a pharmacutical conduct from the standpoint of deliberative judgment. The difference between codes of ethics and codes of practice is discussed. In the era of massive-scale collaboration, this code is a project holding bright prospects for the future. Each individual pharmacutical attitude in practicing their profession must be identified with the pursuit of excellence in their own personal practice for the purpose of achieving the ethical and professional values above and beyond complying with regulations and code of practice.

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

International Nuclear Information System (INIS)

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

1994-01-01

The ppercase nufreq - ppercase np (G.C. Park et al. NUREG/CR-3375, 1983; S.J. Peng et al. NUREG/CR-4116, 1984; S.J. Peng et al. Nucl. Sci. Eng. 88 (1988) 404-411) code was modified and set up at Westinghouse, USA, for mixed fuel type multichannel core-wide stability analysis. The resulting code, ppercase nufreq - ppercase npw , allows for variable axial power profiles between channel groups and can handle mixed fuel types.Various models incorporated into ppercase nurfreq - ppercase npw were systematically compared against the Westinghouse channel stability analysis code ppercase mazda -ppercase nf (R. Taleyarkhan et al. J. Heat Transfer 107 (February 1985) 175-181; NUREG/CR2972, 1983), for which the mathematical model was developed in an entirely different manner. Excellent agreement was obtained which verified the thermal-hydraulic modeling and coding aspects. Detailed comparisons were also performed against nuclear-coupled reactor core stability data. All 13 Peach Bottom-2 EOC-2/3 low flow stability tests (L.A. Carmichael and R.O. Neimi, EPRI NP-564, Project 1020-1, 1978; F.B. Woffinden and R.O. Neimi, EPRI, NP 0972, Project 1020-2, 1981) were simulated. A key aspect for code qualification involved the development of a physically based empirical algorithm to correct for the effect of core inlet flow development on subcooled boiling. Various other modeling assumptions were tested and sensitivity studies performed. Good agreement was obtained between ppercase nufreq-npw predictions and data. ((orig.))

The intercomparison of aerosol codes

International Nuclear Information System (INIS)

Dunbar, I.H.; Fermandjian, J.; Gauvain, J.

1988-01-01

The behavior of aerosols in a reactor containment vessel following a severe accident could be an important determinant of the accident source term to the environment. Various processes result in the deposition of the aerosol onto surfaces within the containment, from where they are much less likely to be released. Some of these processes are very sensitive to particle size, so it is important to model the aerosol growth processes: agglomeration and condensation. A number of computer codes have been written to model growth and deposition processes. They have been tested against each other in a series of code comparison exercises. These exercises have investigated sensitivities to physical and numerical assumptions and have also proved a useful means of quality control for the codes. Various exercises in which code predictions are compared with experimental results are now under way

On the reliability of predictions of geomechanical response - project Cosa in perspective

International Nuclear Information System (INIS)

Knowles, N.C.; Lowe, M.J.S.; Come, B.

1990-01-01

Project COSA (Comparison of computer codes for Salt) was set up by the CEC as international benchmark exercise to compare the reliability of predictions of thermo-mechanical response of HLW repositories in salt. The first phase (COSA I) was conducted between 1984-1986 and attention was directed at code verification issues. The second phase (COSA II), carried out in the period 1986-1988, addressed code validation and other issues. Specifically, a series of experimental heat and pressure tests carried out at the Asse Mine in Wast Germany were modelled and predictions of the thermo-mechanical behaviour were compared. Ten European organisations participated. A key feature of this exercise was that, as far as possible, the calculations were performed blind (i.e. without any knowledge of the observed behaviour) using the best information available a priori, to describe the physical situation to be modelled. Interest centred around the various constitutive models (of material behaviour) for rock-salt and the assumptions about the in situ state of stress. The paper gives an overview of the project, presents some broad conclusions and attempts to assess their significance. 17 refs., 6 figs., 2 tabs

A Comparison of Athletic Movement Among Talent-Identified Juniors From Different Football Codes in Australia: Implications for Talent Development.

Science.gov (United States)

Woods, Carl T; Keller, Brad S; McKeown, Ian; Robertson, Sam

2016-09-01

Woods, CT, Keller, BS, McKeown, I, and Robertson, S. A comparison of athletic movement among talent-identified juniors from different football codes in Australia: implications for talent development. J Strength Cond Res 30(9): 2440-2445, 2016-This study aimed to compare the athletic movement skill of talent-identified (TID) junior Australian Rules football (ARF) and soccer players. The athletic movement skill of 17 TID junior ARF players (17.5-18.3 years) was compared against 17 TID junior soccer players (17.9-18.7 years). Players in both groups were members of an elite junior talent development program within their respective football codes. All players performed an athletic movement assessment that included an overhead squat, double lunge, single-leg Romanian deadlift (both movements performed on right and left legs), a push-up, and a chin-up. Each movement was scored across 3 essential assessment criteria using a 3-point scale. The total score for each movement (maximum of 9) and the overall total score (maximum of 63) were used as the criterion variables for analysis. A multivariate analysis of variance tested the main effect of football code (2 levels) on the criterion variables, whereas a 1-way analysis of variance identified where differences occurred. A significant effect was noted, with the TID junior ARF players outscoring their soccer counterparts when performing the overhead squat and push-up. No other criterions significantly differed according to the main effect. Practitioners should be aware that specific sporting requirements may incur slight differences in athletic movement skill among TID juniors from different football codes. However, given the low athletic movement skill noted in both football codes, developmental coaches should address the underlying movement skill capabilities of juniors when prescribing physical training in both codes.

Lightweight Detection of Android-specific Code Smells : The aDoctor Project

NARCIS (Netherlands)

Palomba, F.; Di Nucci, D.; Panichella, A.; Zaidman, A.E.; De Lucia, Andrea; Pinzger, Martin; Bavota, Gabriele; Marcus, Andrian

2017-01-01

Code smells are symptoms of poor design solutions applied by programmers during the development of software systems. While the research community devoted a lot of effort to studying and devising approaches for detecting the traditional code smells defined by Fowler, little knowledge and support

Project on comparison of structural parameters and electron density maps of oxalic acid dihydrate

NARCIS (Netherlands)

Coppens, Philip; Dam, J.; Harkema, Sybolt; Feil, D.

1984-01-01

Results obtained from four X-ray and five neutron data sets collected under a project sponsored by the Commission on Charge, Spin and Momentum Densities are analyzed by comparison of thermal parameters, positional parameters and X - N electron density maps. Three sets of theoretical calculations are

COCOA Code for Creating Mock Observations of Star Cluster Models

OpenAIRE

Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Dalessandro, Emanuele

2017-01-01

We introduce and present results from the COCOA (Cluster simulatiOn Comparison with ObservAtions) code that has been developed to create idealized mock photometric observations using results from numerical simulations of star cluster evolution. COCOA is able to present the output of realistic numerical simulations of star clusters carried out using Monte Carlo or \\textit{N}-body codes in a way that is useful for direct comparison with photometric observations. In this paper, we describe the C...

Gene-Auto: Automatic Software Code Generation for Real-Time Embedded Systems

Science.gov (United States)

Rugina, A.-E.; Thomas, D.; Olive, X.; Veran, G.

2008-08-01

This paper gives an overview of the Gene-Auto ITEA European project, which aims at building a qualified C code generator from mathematical models under Matlab-Simulink and Scilab-Scicos. The project is driven by major European industry partners, active in the real-time embedded systems domains. The Gene- Auto code generator will significantly improve the current development processes in such domains by shortening the time to market and by guaranteeing the quality of the generated code through the use of formal methods. The first version of the Gene-Auto code generator has already been released and has gone thought a validation phase on real-life case studies defined by each project partner. The validation results are taken into account in the implementation of the second version of the code generator. The partners aim at introducing the Gene-Auto results into industrial development by 2010.

The international hydrocoin project

International Nuclear Information System (INIS)

1987-01-01

The International HYDROCOIN (Hydrologic Code Intercomparison) Project was started in May 1984. Fourteen organizations participate in the Project, with the Swedish Nuclear Power Inspectorate (SKI) as managing participant and the OECD Nuclear Energy Agency (NEA) taking part as a member of the Project secretariat. HYDROCOIN is concerned with the assessment of groundwater movements at potential nuclear waste disposal sites with the help of mathematical models and computer codes. The Project is divided into three levels. The objective of HYDROCOIN Level 1 is to verify the accuracy of groundwater flow codes. HYDROCOIN Level 2 and Level 3 are concerned with validation of hydrological models and uncertainty/sensitivity analysis, respectively. This report, which has been prepared by the NEA Secretariat and SKI in consultation with the co-ordinating group of the HYDROCOIN Project summarizes the background and objectives of the Project and presents the results of the work performed up to the middle of 1987. It is intended to provide general information on HYDROCOIN to interested parties beyond the group of directly involved specialists

Overall simulation of a HTGR plant with the gas adapted MANTA code

International Nuclear Information System (INIS)

Emmanuel Jouet; Dominique Petit; Robert Martin

2005-01-01

Full text of publication follows: AREVA's subsidiary Framatome ANP is developing a Very High Temperature Reactor nuclear heat source that can be used for electricity generation as well as cogeneration including hydrogen production. The selected product has an indirect cycle architecture which is easily adapted to all possible uses of the nuclear heat source. The coupling to the applications is implemented through an Intermediate Heat exchanger. The system code chosen to calculate the steady-state and transient behaviour of the plant is based on the MANTA code. The flexible and modular MANTA code that is originally a system code for all non LOCA PWR plant transients, has been the subject of new developments to simulate all the forced convection transients of a nuclear plant with a gas cooled High Temperature Reactor including specific core thermal hydraulics and neutronics modelizations, gas and water steam turbomachinery and control structure. The gas adapted MANTA code version is now able to model a total HTGR plant with a direct Brayton cycle as well as indirect cycles. To validate these new developments, a modelization with the MANTA code of a real plant with direct Brayton cycle has been performed and steady-states and transients compared with recorded thermal hydraulic measures. Finally a comparison with the RELAP5 code has been done regarding transient calculations of the AREVA indirect cycle HTR project plant. Moreover to improve the user-friendliness in order to use MANTA as a systems conception, optimization design tool as well as a plant simulation tool, a Man- Machine-Interface is available. Acronyms: MANTA Modular Advanced Neutronic and Thermal hydraulic Analysis; HTGR High Temperature Gas-Cooled Reactor. (authors)

Comparison of codes and standards for radiographic inspection

International Nuclear Information System (INIS)

Bingoeldag, M. M.; Aksu, M.; Akguen, A. F.

1995-01-01

This report compares the procedurel requirements and acceptance criteria for radiographic inspections specified in the relevant national and international codes and standards. In particular, detailed analysis of inspection conditions such as exposure arrangements, and contrast requirements are given

Validation and Improvement of the FEMAXI-JNES Code by Using PIE Data at Extended Burnup. Final Report for FUMEX-III

International Nuclear Information System (INIS)

Hirose, Tsutomu; Miura, Hiromichi; Kitamura, Toshiya; Kamimura, Katsuichiro

2013-01-01

Japan Nuclear Energy Safety Organization (JNES) has participated in the IAEA FUMEX-III Coordinated Research Project (CRP) on the Improvement of Computer Codes Used for Fuel Behaviour Simulation for the following purpose. 1. Cooperate between member states and exchange information and expertise for understanding of fuel modelling and improvement 2. Develop and improve the FEMAXI-JNES code as an audit code for Japanese safety licensing review of fuel rod design, especially, - High burnup fuel - MOX fuel 3. Set the standard models for the FEMAXI-JNES code to provide best-estimate predictions of the thermal and mechanical performance of LWR fuel rod This is the JNES's final report for the FUMEX-III CRP. During the period of the CRP, JNES has modified pellet swelling and fission gas release models, and demonstrated the predictive capability relative to fuel centerline temperature, fission gas release, fuel rod internal gas pressure, cladding diametral deformation and cladding elongation by comparisons of integral code predictions of these parameters to experimental (measured) data from OECD/NEA IFPE database. (author)

Application of RASCAL code for multiunit accident in domestic nuclear sites

Energy Technology Data Exchange (ETDEWEB)

Park, Sang Hyun; Jeong, Seung Young [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-10-15

All of domestic nuclear power plant sites are multiunit site (at least 5 - 6 reactors are operating), so this capability has to be quickly secured for nuclear licensee and institutes responsible for nuclear emergency response. In this study, source term and offsite dose from multiunit event were assessed using a computer code, RASCAL. An emergency exercise scenario was chosen to verify applicability of the codes to domestic nuclear site accident. Employing tools and new features of the code, such as merging more than two individual source terms and source term estimate for long term progression accident, main parameters and information in the scenario, release estimates and dose projections were performed. Radiological releases and offsite doses from multiunit accident were calculated using RASCAL.. A scenario, in which three reactors were damaged coincidently by a great natural disaster, was considered. Surrogate plants were chosen for the code calculation. Source terms of each damaged unit were calculated individually first, and then total source term and integrated offsite dose assessment data was acquired using a source term merge function in the code. Also comparison between LTSBO and LOCA source term estimate options was performed. Differences in offsite doses were caused by release characteristics. From LTSBO option, iodines were released much higher than LOCA. Also LTSBO source term release was delayed and the duration was longer than LOCA. This option would be useful to accidents which progress with much longer time frame than LOCA. RASCAL can be useful tool for radiological consequence assessment in domestic nuclear site accidents.

Automated delivery of codes for charge in radiotherapy

International Nuclear Information System (INIS)

Sauer, Michael; Volz, Steffen; Hall, Markus; Roehner, Fred; Frommhold, Hermann; Grosu, Anca-Ligia; Heinemann, Felix

2010-01-01

Background and purpose: for the medical billing of Radiotherapy every fraction has to be encoded, including date and time of all administered treatments. With fractions averaging 30 per patient and about 2,500 new patients every year the number of Radiotherapy codes reaches an amount of 70,000 and more. Therefore, an automated proceeding for transferring and processing therapy codes has been developed at the Department of Radiotherapy Freiburg, Germany. This is a joint project of the Department of Radiotherapy, the Administration Department, and the Central II Department of the University Hospital of Freiburg. Material and methods: the project consists of several modules whose collaboration makes the projected automated transfer of treatment codes possible. The first step is to extract the data from the department's Clinical Information System (MOSAIQ). These data are transmitted to the Central IT Department via an HL7 interface, where a check for corresponding hospitalization data is performed. In the further processing of the data, a matching table plays an important role allowing the transformation of a treatment code into a valid medical billing code. In a last step, the data are transferred to the medical billing system. Results and conclusion: after assembling and implementing the particular modules successfully, a first beta test was launched. In order to test the modules separately as well as the interaction of the components, extensive tests were performed during March 2006. Soon it became clear that the tested procedure worked efficiently and accurately. In April 2006, a pilot project with a few qualities of treatment (e.g., computed tomography, simulation) was put into practice. Since October 2006, nearly all Radiation Therapy codes (∝ 75,000) are being transferred to the comprehensive Hospital Information System (HIS) automatically in a daily routine. (orig.)

Comparison of the THYC and FLICA-3M codes by the pseudo-cubic thin-plate method; Comparaison par la methode des plaques de predicteurs de flux critique obtenus a l`aide des codes THYC et FLICA-3

Energy Technology Data Exchange (ETDEWEB)

Banner, D; Crecy, F de

1993-06-01

The pseudo cubic Spline method (PCSM) is a statistical tool developed by the CEA. It is designed to analyse experimental points and in particular thermalhydraulic data. Predictors of the occurrence of critical heat flux are obtained by using Spline functions. In this paper, predictors have been computed from the same CHF databases by using two different flow analyses to derive local thermal-hydraulic variables at the CHF location. In fact, CEA`s FLICA-3M represents rod bundles by interconnected subchannels whereas EDF`s THYC code uses a porous 3D approach. In a first step, the PCSM is briefly presented as well as the two codes studied here. Then, the comparison methodology is explained in order to prove that advanced analysis of thermalhydraulic codes can be achieved with the PCSM. (authors). 6 figs., 2 tabs., 5 refs.

Application of the three-dimensional Oak Ridge transport code

International Nuclear Information System (INIS)

Rhoades, W.A.; Childs, R.L.; Emmett, M.B.; Cramer, S.N.

1984-01-01

TORT, a 3-d extension of the DOT discrete ordinates transport code is now in production use for studies of radiation penetration into large concrete and masonry structures. This paper discusses certain features of the new code and shows representative results, including comparisons with Monte Carlo calculations

Country Report on Building Energy Codes in China

Energy Technology Data Exchange (ETDEWEB)

Shui, Bin; Evans, Meredydd; Lin, H.; Jiang, Wei; Liu, Bing; Song, Bo; Somasundaram, Sriram

2009-04-15

This report is part of a series of reports on building energy efficiency codes in countries associated with the Asian Pacific Partnership (APP) - Australia, South Korea, Japan, China, India, and the United States of America (U.S.). This reports gives an overview of the development of building energy codes in China, including national energy policies related to building energy codes, history of building energy codes, recent national projects and activities to promote building energy codes. The report also provides a review of current building energy codes (such as building envelope and HVAC) for commercial and residential buildings in China.

Country Report on Building Energy Codes in Japan

Energy Technology Data Exchange (ETDEWEB)

Evans, Meredydd; Shui, Bin; Takagi, T.

2009-04-15

This report is part of a series of reports on building energy efficiency codes in countries associated with the Asian Pacific Partnership (APP) - Australia, South Korea, Japan, China, India, and the United States of America (U.S.). This reports gives an overview of the development of building energy codes in Japan, including national energy policies related to building energy codes, history of building energy codes, recent national projects and activities to promote building energy codes. The report also provides a review of current building energy codes (such as building envelope, HVAC, and lighting) for commercial and residential buildings in Japan.

Country Report on Building Energy Codes in Australia

Energy Technology Data Exchange (ETDEWEB)

Shui, Bin; Evans, Meredydd; Somasundaram, Sriram

2009-04-02

This report is part of a series of reports on building energy efficiency codes in countries associated with the Asian Pacific Partnership (APP) - Australia, South Korea, Japan, China, India, and the United States of America (U.S.). This reports gives an overview of the development of building energy codes in Australia, including national energy policies related to building energy codes, history of building energy codes, recent national projects and activities to promote building energy codes. The report also provides a review of current building energy codes (such as building envelope, HVAC, and lighting) for commercial and residential buildings in Australia.

Country Report on Building Energy Codes in Canada

Energy Technology Data Exchange (ETDEWEB)

Shui, Bin; Evans, Meredydd

2009-04-06

This report is part of a series of reports on building energy efficiency codes in countries associated with the Asian Pacific Partnership (APP) - Australia, South Korea, Japan, China, India, and the United States of America . This reports gives an overview of the development of building energy codes in Canada, including national energy policies related to building energy codes, history of building energy codes, recent national projects and activities to promote building energy codes. The report also provides a review of current building energy codes (such as building envelope, HVAC, lighting, and water heating) for commercial and residential buildings in Canada.

Development status of the lattice physics code in COSINE project

Energy Technology Data Exchange (ETDEWEB)

Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y. [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software NEKLS, North Third Ring Road, Beijing 100029 (China)

2013-07-01

LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)

Development status of the lattice physics code in COSINE project

International Nuclear Information System (INIS)

Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y.

2013-01-01

LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)

Comparison of spectra for validation of Penelope code for the energy range used in mammography; Comparacao de espectros para validacao do codigo PENELOPE para faixa de energia usada em mamografia

Energy Technology Data Exchange (ETDEWEB)

Albuquerque, M.A.G.; Ferreira, N.M.P.D., E-mail: malbuqueque@hotmail.co [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Pires, E.; Ganizeu, M.D.; Almeida, C.E. de, E-mail: marianogd@uol.com.b [Universidade do Estado do Rio de Janeiro (UERJ), RJ (Brazil); Prizio, R.; Peixoto, J.G., E-mail: guilherm@ird.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2009-07-01

The spectra simulated by the Penelope code were compared with the spectra experimentally obtained through the silicon PIN photodiode detector, and with spectra calculated by the code of IPEN, and the comparison exhibited a concordance of 93.3 %, and make them an option for study of X-ray spectroscopy in the voltage range used in mammography

Analyses of CsI aerosol deposition in aerosol behavior tests in WIND project

International Nuclear Information System (INIS)

Kudo, Tamotsu; Shibazaki, Hiroaki; Hidaka, Akihide

1999-01-01

The aerosol deposition tests have been performed in WIND project at JAERI to characterize the aerosol behavior. The aerosol deposition tests named WAV1-D and WAV2-D were analyzed by aerosol behavior analysis codes, JAERI's ART and SNL's VICTORIA. The comparison calculation was performed for the confirmation of the analytical capabilities of the both codes and improvement of the models in ART. The deposition mass calculated by ART was larger than that by VICTORIA. This discrepancy is caused by differences in model for FP vapor condensation onto the wall surface. In the WAV2-D test, in which boric acid was placed on the floor area of the test section prior to the deposition phase to simulate the PWR primary coolant, there was a discrepancy in deposition mass between analytical results in both codes and experimental results. The discrepancy may be caused by existence of boric acid which is not considered in the codes. (author)

Improving coding accuracy in an academic practice.

Science.gov (United States)

Nguyen, Dana; O'Mara, Heather; Powell, Robert

2017-01-01

Practice management has become an increasingly important component of graduate medical education. This applies to every practice environment; private, academic, and military. One of the most critical aspects of practice management is documentation and coding for physician services, as they directly affect the financial success of any practice. Our quality improvement project aimed to implement a new and innovative method for teaching billing and coding in a longitudinal fashion in a family medicine residency. We hypothesized that implementation of a new teaching strategy would increase coding accuracy rates among residents and faculty. Design: single group, pretest-posttest. military family medicine residency clinic. Study populations: 7 faculty physicians and 18 resident physicians participated as learners in the project. Educational intervention: monthly structured coding learning sessions in the academic curriculum that involved learner-presented cases, small group case review, and large group discussion. overall coding accuracy (compliance) percentage and coding accuracy per year group for the subjects that were able to participate longitudinally. Statistical tests used: average coding accuracy for population; paired t test to assess improvement between 2 intervention periods, both aggregate and by year group. Overall coding accuracy rates remained stable over the course of time regardless of the modality of the educational intervention. A paired t test was conducted to compare coding accuracy rates at baseline (mean (M)=26.4%, SD=10%) to accuracy rates after all educational interventions were complete (M=26.8%, SD=12%); t24=-0.127, P=.90. Didactic teaching and small group discussion sessions did not improve overall coding accuracy in a residency practice. Future interventions could focus on educating providers at the individual level.

Attention Filtering in the Design of Electronic Map Displays: A Comparison of Color-Coding, Intensity Coding, and Decluttering Techniques

National Research Council Canada - National Science Library

Yeh, Michelle; Wickens, Christopher D

2000-01-01

In a series of experiments, the use of color-coding, intensity coding, and decluttering were compared order to assess their potential benefits for accessing information from electronic map displays...

Predictive Bias and Sensitivity in NRC Fuel Performance Codes

Energy Technology Data Exchange (ETDEWEB)

Geelhood, Kenneth J.; Luscher, Walter G.; Senor, David J.; Cunningham, Mitchel E.; Lanning, Donald D.; Adkins, Harold E.

2009-10-01

The latest versions of the fuel performance codes, FRAPCON-3 and FRAPTRAN were examined to determine if the codes are intrinsically conservative. Each individual model and type of code prediction was examined and compared to the data that was used to develop the model. In addition, a brief literature search was performed to determine if more recent data have become available since the original model development for model comparison.

Comparison of SAND-II and FERRET

International Nuclear Information System (INIS)

Wootan, D.W.; Schmittroth, F.

1981-01-01

A comparison was made of the advantages and disadvantages of two codes, SAND-II and FERRET, for determining the neutron flux spectrum and uncertainty from experimental dosimeter measurements as anticipated in the FFTF Reactor Characterization Program. This comparison involved an examination of the methodology and the operational performance of each code. The merits of each code were identified with respect to theoretical basis, directness of method, solution uniqueness, subjective influences, and sensitivity to various input parameters

LFSC - Linac Feedback Simulation Code

Energy Technology Data Exchange (ETDEWEB)

Ivanov, Valentin; /Fermilab

2008-05-01

The computer program LFSC (Code>) is a numerical tool for simulation beam based feedback in high performance linacs. The code LFSC is based on the earlier version developed by a collective of authors at SLAC (L.Hendrickson, R. McEwen, T. Himel, H. Shoaee, S. Shah, P. Emma, P. Schultz) during 1990-2005. That code was successively used in simulation of SLC, TESLA, CLIC and NLC projects. It can simulate as pulse-to-pulse feedback on timescale corresponding to 5-100 Hz, as slower feedbacks, operating in the 0.1-1 Hz range in the Main Linac and Beam Delivery System. The code LFSC is running under Matlab for MS Windows operating system. It contains about 30,000 lines of source code in more than 260 subroutines. The code uses the LIAR ('Linear Accelerator Research code') for particle tracking under ground motion and technical noise perturbations. It uses the Guinea Pig code to simulate the luminosity performance. A set of input files includes the lattice description (XSIF format), and plane text files with numerical parameters, wake fields, ground motion data etc. The Matlab environment provides a flexible system for graphical output.

Analyses of CsI aerosol deposition tests in WIND project with ART and VICTORIA codes

International Nuclear Information System (INIS)

Yuchi, Y.; Shibazaki, H.; Kudo, T.

2000-01-01

Deposition behavior of cesium iodide (CsI) was analyzed with ART and VICTORIA-92 codes for a test of the aerosol re-vaporization test series performed in WIND project at JAERI. In the test analyzed, CsI aerosol was injected into piping of test section where metaboric acid (HBO 2 ) was placed in advance on the floor area. It was confirmed in the present analysis that similar results on the CsI deposition were obtained between ART and VICTORIA when influences of chemical interactions were negligibly small. The analysis with VICTORIA agreed satisfactorily with the test results in analytical cases that cesium metaborate (CsBO 2 ) was injected into the test section instead of CsI to simulate the pre-existence of HBO 2 effect. (author)

Introduction of gadolinium in the library of Leopard code

International Nuclear Information System (INIS)

Claro, L.H.; Menezes, A.

1989-12-01

The materials Gd-154, Gd-155, Gd-156 and Gd-157 were included in the LEOPARD code library at the request of FURNAS Centrais Eletricas S.A. Results from comparison of LEOPARD and WIMSD/4 codes for a typical cell with 7 burnup steps, are presented. (author) [pt

Code-To-Code Benchmarking Of The Porflow And GoldSim Contaminant Transport Models Using A Simple 1-D Domain - 11191

International Nuclear Information System (INIS)

Hiergesell, R.; Taylor, G.

2010-01-01

An investigation was conducted to compare and evaluate contaminant transport results of two model codes, GoldSim and Porflow, using a simple 1-D string of elements in each code. Model domains were constructed to be identical with respect to cell numbers and dimensions, matrix material, flow boundary and saturation conditions. One of the codes, GoldSim, does not simulate advective movement of water; therefore the water flux term was specified as a boundary condition. In the other code, Porflow, a steady-state flow field was computed and contaminant transport was simulated within that flow-field. The comparisons were made solely in terms of the ability of each code to perform contaminant transport. The purpose of the investigation was to establish a basis for, and to validate follow-on work that was conducted in which a 1-D GoldSim model developed by abstracting information from Porflow 2-D and 3-D unsaturated and saturated zone models and then benchmarked to produce equivalent contaminant transport results. A handful of contaminants were selected for the code-to-code comparison simulations, including a non-sorbing tracer and several long- and short-lived radionuclides exhibiting both non-sorbing to strongly-sorbing characteristics with respect to the matrix material, including several requiring the simulation of in-growth of daughter radionuclides. The same diffusion and partitioning coefficients associated with each contaminant and the half-lives associated with each radionuclide were incorporated into each model. A string of 10-elements, having identical spatial dimensions and properties, were constructed within each code. GoldSim's basic contaminant transport elements, Mixing cells, were utilized in this construction. Sand was established as the matrix material and was assigned identical properties (e.g. bulk density, porosity, saturated hydraulic conductivity) in both codes. Boundary conditions applied included an influx of water at the rate of 40 cm/yr at one

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

The reason for having a code of pharmaceutical ethics: Spanish Pharmacists Code of Ethics.

Science.gov (United States)

Barreda Hernández, Dolores; Mulet Alberola, Ana; González Bermejo, Diana; Soler Company, Enrique

2017-05-01

The pharmacist profession needs its own code of conduct set out in writing to serve as a stimulus to pharmacists in their day-to-day work in the different areas of pharmacy, in conjunction always with each individual pharmacist´s personal commitment to their patients, to other healthcare professionals and to society. An overview is provided of the different codes of ethics for pharmacists on the national and international scale, the most up-to-date code for 2015 being presented as a set of principles which must guide a pharmacutical conduct from the standpoint of deliberative judgment. The difference between codes of ethics and codes of practice is discussed. In the era of massive-scale collaboration, this code is a project holding bright prospects for the future. Each individual pharmacutical attitude in practicing their profession must be identified with the pursuit of excellence in their own personal practice for the purpose of achieving the ethical and professional values above and beyond complying with regulations and code of practice. Copyright AULA MEDICA EDICIONES 2017. Published by AULA MEDICA. All rights reserved.

Deciphering the genetic regulatory code using an inverse error control coding framework.

Energy Technology Data Exchange (ETDEWEB)

Rintoul, Mark Daniel; May, Elebeoba Eni; Brown, William Michael; Johnston, Anna Marie; Watson, Jean-Paul

2005-03-01

We have found that developing a computational framework for reconstructing error control codes for engineered data and ultimately for deciphering genetic regulatory coding sequences is a challenging and uncharted area that will require advances in computational technology for exact solutions. Although exact solutions are desired, computational approaches that yield plausible solutions would be considered sufficient as a proof of concept to the feasibility of reverse engineering error control codes and the possibility of developing a quantitative model for understanding and engineering genetic regulation. Such evidence would help move the idea of reconstructing error control codes for engineered and biological systems from the high risk high payoff realm into the highly probable high payoff domain. Additionally this work will impact biological sensor development and the ability to model and ultimately develop defense mechanisms against bioagents that can be engineered to cause catastrophic damage. Understanding how biological organisms are able to communicate their genetic message efficiently in the presence of noise can improve our current communication protocols, a continuing research interest. Towards this end, project goals include: (1) Develop parameter estimation methods for n for block codes and for n, k, and m for convolutional codes. Use methods to determine error control (EC) code parameters for gene regulatory sequence. (2) Develop an evolutionary computing computational framework for near-optimal solutions to the algebraic code reconstruction problem. Method will be tested on engineered and biological sequences.

Dual Coding and Bilingual Memory.

Science.gov (United States)

Paivio, Allan; Lambert, Wallace

1981-01-01

Describes study which tested a dual coding approach to bilingual memory using tasks that permit comparison of the effects of bilingual encoding with verbal-nonverbal dual encoding items. Results provide strong support for a version of the independent or separate stories view of bilingual memory. (Author/BK)

RAID-6Plus: A Comprised Methodology for Extending RAID-6 Codes

Directory of Open Access Journals (Sweden)

Ming-Zhu Deng

2017-01-01

Full Text Available Existing RAID-6 code extensions assume that failures are independent and instantaneous, overlooking the underlying mechanism of multifailure occurrences. Also, the effect of reconstruction window is ignored. Additionally, these coding extensions have not been adapted to occurrence patterns of failure in real-world applications. As a result, the third parity drive is set to handle the triple-failure scenario; however, the lower level failure situations have been left unattended. Therefore, a new methodology of extending RAID-6 codes named RAID-6Plus with better compromise has been studied in this paper. RAID-6Plus (Deng et al., 2015 employs short combinations which can greatly reuse overlapped elements during reconstruction to remake the third parity drive. A sample extension code called RDP+ is given based on RDP. Moreover, we extended the study to present another extension example called X-code+ which has better update penalty and load balance. The analysis shows that RAID-6Plus is a balanced tradeoff of reliability, performance, and practicality. For instance, RDP+ could achieve speedups as high as 33.4% in comparison to the RTP with conventional rebuild, 11.9% in comparison to RTP with the optimal rebuild, 47.7% in comparison to STAR with conventional rebuild, and 26.2% for a single failure rebuild.

A NEM diffusion code for fuel management and time average core calculation

International Nuclear Information System (INIS)

Mishra, Surendra; Ray, Sherly; Kumar, A.N.

2005-01-01

A computer code based on Nodal expansion method has been developed for solving two groups three dimensional diffusion equation. This code can be used for fuel management and time average core calculation. Explicit Xenon and fuel temperature estimation are also incorporated in this code. TAPP-4 phase-B physics experimental results were analyzed using this code and a code based on FD method. This paper gives the comparison of the observed data and the results obtained with this code and FD code. (author)

AMZ, library of multigroup constants for EXPANDA computer codes, generated by NJOY computer code from ENDF/B-IV

International Nuclear Information System (INIS)

Chalhoub, E.S.; Moraes, M. de.

1984-01-01

A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author) [pt

Acceptance and validation test report for HANSF code version 1.3.2

International Nuclear Information System (INIS)

PIEPHO, M.G.

2001-01-01

The HANSF code, Version 1.3.2, is a stand-along code that runs only in DOS. As a result, it runs on any Windows' platform, since each Windows(trademark) platform can create a DOS-prompt window and execute HANSF in the DOS window. The HANSF code is proprietary to Fauske and Associates, Inc., (FAI) of Burr Ridge, IL, the developers of the code. The SNF Project has a license from FAI to run the HANSF code on any computer for only work related to SNF Project. The SNF Project owns the MCO.FOR routine, which is the main routine in HANSF for CVDF applications. The HANSF code calculates physical variables such as temperature, pressure, oxidation rates due to chemical reactions of uranium metal/fuel with water or oxygen. The code is used by the Spent Nuclear Fuel (SNF) Project at Hanford; for example, the report Thermal Analysis of Cold Vacuum Drying of Spent Nuclear Fuel (HNF-SD-SNF-CN-023). The primary facilities of interest are the K-Basins, Cold Vacuum Drying Facility (CVDF), Canister Storage Building (CSB) and T Plant. The overall Summary is presented in Section 2.0, Variances in Section 3.0, Comprehensive Assessment in Section 4.0, Results in Section 5.0, Evaluation in Section 6.0, and Summary of Activities in Section 7.0

Comparison between the Findings from the TROI Experiments and the Sensitivity Studies by Using the TEXAS-V Code

International Nuclear Information System (INIS)

Park, I. K.; Kim, J. H.; Hong, S. W.; Min, B. T.; Hong, S. H.; Song, J. H.; Kim, H. D.

2006-01-01

Since a steam explosion may breach the integrity of a reactor vessel and containment, it is one of the most important severe accident issues. So, a lot of experimental and analytical researches on steam explosions have been performed. Although many findings from the steam explosion researches have been obtained, there still exist unsolved issues such as the explosivity of the real core material(corium) and the conversion ratio from the thermal energy to the mechanical energy. TROI experiments were carried out to provide the experimental data for these issues. The TROI experiments were performed with a prototypic material such as ZrO 2 melt and a mixture of ZrO 2 and UO 2 melt (corium). Several steam explosion codes including TEXAS-V had been developed by considering the findings in the past steam explosion experiments. However, some unique findings on steam explosions have been obtained from a series of TROI experiments. These findings should be considered in the application to a reactor safety analysis by using a computational code. In this paper, several findings from TROI experiments are discussed and the sensitivity studies on the TROI experimental parameters were conducted by using TEXAS-V code and TROI-13 test. The comparison between the TROI experimental findings and the results of the sensitivity study might allow us to know which parameter is important and which model is uncertain for steam explosions

Comparison of two numerical modelling codes for hydraulic and transport calculations in the near-field

International Nuclear Information System (INIS)

Kalin, J.; Petkovsek, B.; Montarnal, Ph.; Genty, A.; Deville, E.; Krivic, J.; Ratej, J.

2011-01-01

In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.

Comparison of two numerical modelling codes for hydraulic and transport calculations in the near-field

Energy Technology Data Exchange (ETDEWEB)

Kalin, J., E-mail: jan.kalin@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Petkovsek, B., E-mail: borut.petkovsek@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Montarnal, Ph., E-mail: philippe.montarnal@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Genty, A., E-mail: alain.genty@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Deville, E., E-mail: estelle.deville@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Krivic, J., E-mail: jure.krivic@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia); Ratej, J., E-mail: joze.ratej@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia)

2011-04-15

In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.

Towards Effective Intra-flow Network Coding in Software Defined Wireless Mesh Networks

Directory of Open Access Journals (Sweden)

Donghai Zhu

2016-01-01

Full Text Available Wireless Mesh Networks (WMNs have potential to provide convenient broadband wireless Internet access to mobile users.With the support of Software-Defined Networking (SDN paradigm that separates control plane and data plane, WMNs can be easily deployed and managed. In addition, by exploiting the broadcast nature of the wireless medium and the spatial diversity of multi-hop wireless networks, intra-flow network coding has shown a greater benefit in comparison with traditional routing paradigms in data transmission for WMNs. In this paper, we develop a novel OpenCoding protocol, which combines the SDN technique with intra-flow network coding for WMNs. Our developed protocol can simplify the deployment and management of the network and improve network performance. In OpenCoding, a controller that works on the control plane makes routing decisions for mesh routers and the hop-by-hop forwarding function is replaced by network coding functions in data plane. We analyze the overhead of OpenCoding. Through a simulation study, we show the effectiveness of the OpenCoding protocol in comparison with existing schemes. Our data shows that OpenCoding outperforms both traditional routing and intra-flow network coding schemes.

The new lattice code Paragon and its qualification for PWR core applications

International Nuclear Information System (INIS)

Ouisloumen, M.; Huria, H.C.; Mayhue, L.T.; Smith, R.M.; Kichty, M.J.; Matsumoto, H.; Tahara, Y.

2003-01-01

Paragon is a new two-dimensional transport code based on collision probability with interface current method and written entirely in Fortran 90/95. The qualification of Paragon has been completed and the results are very good. This qualification included a number of critical experiments. Comparisons to the Monte Carlo code MCNP for a wide variety of PWR assembly lattice types were also performed. In addition, Paragon-based core simulator models have been compared against PWR plant startup and operational data for a large number of plants. Some results of these calculations and also comparisons against models developed with a licensed Westinghouse lattice code, Phoenix-P, are presented. The qualification described in this paper provided the basis for the qualification of Paragon both as a validated transport code and as the nuclear data source for core simulator codes

Development of an advanced fluid-dynamic analysis code: α-flow

International Nuclear Information System (INIS)

Akiyama, Mamoru

1990-01-01

A Project for development of large scale three-dimensional fluid-dynamic analysis code, α-FLOW, coping with the recent advancement of supercomputers and workstations, has been in progress. This project is called the α-Project, which has been promoted by the Association for Large Scale Fluid Dynamics Analysis Code comprising private companies and research institutions such as universities. The developmental period for the α-FLOW is four years, March 1989 to March 1992. To date, the major portions of basic design and program preparation have been completed and the project is in the stage of testing each module. In this paper, the present status of the α-Project, design policy and outline of α-FLOW are described. (author)

On transform coding tools under development for VP10

Science.gov (United States)

Parker, Sarah; Chen, Yue; Han, Jingning; Liu, Zoe; Mukherjee, Debargha; Su, Hui; Wang, Yongzhe; Bankoski, Jim; Li, Shunyao

2016-09-01

Google started the WebM Project in 2010 to develop open source, royaltyfree video codecs designed specifically for media on the Web. The second generation codec released by the WebM project, VP9, is currently served by YouTube, and enjoys billions of views per day. Realizing the need for even greater compression efficiency to cope with the growing demand for video on the web, the WebM team embarked on an ambitious project to develop a next edition codec, VP10, that achieves at least a generational improvement in coding efficiency over VP9. Starting from VP9, a set of new experimental coding tools have already been added to VP10 to achieve decent coding gains. Subsequently, Google joined a consortium of major tech companies called the Alliance for Open Media to jointly develop a new codec AV1. As a result, the VP10 effort is largely expected to merge with AV1. In this paper, we focus primarily on new tools in VP10 that improve coding of the prediction residue using transform coding techniques. Specifically, we describe tools that increase the flexibility of available transforms, allowing the codec to handle a more diverse range or residue structures. Results are presented on a standard test set.

VIPRE-01: A thermal-hydraulic code for reactor cores

International Nuclear Information System (INIS)

Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.

1989-08-01

The VIPRE-01 thermal hydraulics code for PWR and BWR analysis has undergone significant modifications and error correction. This manual for the updated code, designated as VIPRE-01 Mod-02, describes improvements that eliminate problems of slow convergence with the drift flux model in transient simulation. To update the VIPRE-01 code and its documentation the drift flux model of two-phase flow was implemented and error corrections developed during VIPRE-01 application were included. The project team modified the existing VIPRE-01 equations into drift flux model equations by developing additional terms. They also developed and implemented corrections for the errors identified during the last four years. They then validated the modified code against standard test data using selected test cases. The project team prepared documentation revisions reflecting code improvements and corrections to replace the corresponding sections in the original VIPRE documents. The revised VIPRE code, designated VIPRE-01 Mod-02, incorporates improvements that eliminate many shortcomings of the previous version. During the validation, the code produced satisfactory output compared with test data. The revised documentation is in the form of binder pages to replace existing pages in three of the original manuals

Calculation of neutron spectra produced in neutron generator target: Code testing.

Science.gov (United States)

Gaganov, V V

2018-03-01

DT-neutron spectra calculated using the SRIANG code was benchmarked against the results obtained by widely used Monte Carlo codes: PROFIL, SHORIN, TARGET, ENEA-JSI, MCUNED, DDT and NEUSDESC. The comparison of the spectra obtained by different codes confirmed the correctness of SRIANG calculations. The cross-checking of the compared spectra revealed some systematic features and possible errors of analysed codes. Copyright © 2017 Elsevier Ltd. All rights reserved.

New code of conduct

CERN Multimedia

Laëtitia Pedroso

2010-01-01

During his talk to the staff at the beginning of the year, the Director-General mentioned that a new code of conduct was being drawn up. What exactly is it and what is its purpose? Anne-Sylvie Catherin, Head of the Human Resources (HR) Department, talked to us about the whys and wherefores of the project.   Drawing by Georges Boixader from the cartoon strip “The World of Particles” by Brian Southworth. A code of conduct is a general framework laying down the behaviour expected of all members of an organisation's personnel. “CERN is one of the very few international organisations that don’t yet have one", explains Anne-Sylvie Catherin. “We have been thinking about introducing a code of conduct for a long time but lacked the necessary resources until now”. The call for a code of conduct has come from different sources within the Laboratory. “The Equal Opportunities Advisory Panel (read also the "Equal opportuni...

The numerical comparison of fire combustion model and water-mist suppression with experiments by FDS code

International Nuclear Information System (INIS)

Li Hsuennien; Ferng Yuhming; Shih Chunkuan; Hsu Wensheng

2007-01-01

FDS [1] code numerically solves a form of the Navier-Stokes equations appropriate for low-speed, thermally driven flow with an emphasis on smoke and heat transport from fires. FDS uses a mixture fraction combustion model. The mixture fraction is a conserved scalar quantity that is defined as the fraction of fuel gas at a given point in the flow field. The model assumes that combustion is mixing-controlled, and that the reaction of fuel and oxygen is infinitely fast. In FDS, Lagrangian particles are used to simulate smoke movement and sprinkling water-mist discharge. In order to evaluate the combustion model and water-mist suppression function of the code, FDS analyses are conducted to simulate two enclosure fire cases available in the literature. Comparisons with other combustion models are also made. For fires suppression by water-mist in FDS, parametric studies are performed to compare various water-mist injection characteristics for maximum suppression. Numerical results indicate that the flame suppression is closely related to characteristics of the water mist, such as droplet diameter, mist injection velocity, injection density. Our present investigations show that the combustion model and water-mist suppression in FDS can provide simulation results that are comparable with the experiments. (author)

A reflexive exploration of two qualitative data coding techniques

Directory of Open Access Journals (Sweden)

Erik Blair

2016-01-01

Full Text Available In an attempt to help find meaning within qualitative data, researchers commonly start by coding their data. There are a number of coding systems available to researchers and this reflexive account explores my reflections on the use of two such techniques. As part of a larger investigation, two pilot studies were undertaken as a means to examine the relative merits of open coding and template coding for examining transcripts. This article does not describe the research project per se but attempts to step back and offer a reflexive account of the development of data coding tools. Here I reflect upon and evaluate the two data coding techniques that were piloted, and discuss how using appropriate aspects of both led to the development of my final data coding approach. My exploration found there was no clear-cut ‘best’ option but that the data coding techniques needed to be reflexively-aligned to meet the specific needs of my project. This reflection suggests that, when coding qualitative data, researchers should be methodologically thoughtful when they attempt to apply any data coding technique; that they do not assume pre-established tools are aligned to their particular paradigm; and that they consider combining and refining established techniques as a means to define their own specific codes. DOI: 10.2458/azu_jmmss.v6i1.18772DOI: 10.2458/azu_jmmss.v6i1.18772

Comparison of the results of several heat transfer computer codes when applied to a hypothetical nuclear waste repository

International Nuclear Information System (INIS)

Claiborne, H.C.; Wagner, R.S.; Just, R.A.

1979-12-01

A direct comparison of transient thermal calculations was made with the heat transfer codes HEATING5, THAC-SIP-3D, ADINAT, SINDA, TRUMP, and TRANCO for a hypothetical nuclear waste repository. With the exception of TRUMP and SINDA (actually closer to the earlier CINDA3G version), the other codes agreed to within +-5% for the temperature rises as a function of time. The TRUMP results agreed within +-5% up to about 50 years, where the maximum temperature occurs, and then began an oscillary behavior with up to 25% deviations at longer times. This could have resulted from time steps that were too large or from some unknown system problems. The available version of the SINDA code was not compatible with the IBM compiler without using an alternative method for handling a variable thermal conductivity. The results were about 40% low, but a reasonable agreement was obtained by assuming a uniform thermal conductivity; however, a programming error was later discovered in the alternative method. Some work is required on the IBM version to make it compatible with the system and still use the recommended method of handling variable thermal conductivity. TRANCO can only be run as a 2-D model, and TRUMP and CINDA apparently required longer running times and did not agree in the 2-D case; therefore, only HEATING5, THAC-SIP-3D, and ADINAT were used for the 3-D model calculations. The codes agreed within +-5%; at distances of about 1 ft from the waste canister edge, temperature rises were also close to that predicted by the 3-D model

Description and applicability of the BEFEM-CODE

Energy Technology Data Exchange (ETDEWEB)

Groth, T.

1980-05-15

The BEFEM-CODE, developed for rock mechanics problems in hard rock with joints, is a simple FEM code constructed using triangular and quadrilateral elements. As an option, a joint element of the Goodman type may be used. The Cook-Pian type quadrilateral stress hybrid element was introduced into the version of the code used for the Naesliden project, to replace the constant stress quadrilateral elements. This hybrid element, derived with assumed stress distributions, simplifies the excavation process for use in non-linear models. The shear behavior of the Goodman 1976 joint element has been replaced by Goodman's 1968 formulation. This element makes it possible to take dilation into account, but it was not considered necessary to use dilation to simulate proper joint behavior in the Naesliden project. The code uses Barton's shear strength criteria. Excessive nodal forces due to failure and non-linearities in the joint elements are redistributed with stress transfer iterations. Convergence can be speeded up by dividing each excavation sequence into several loadsteps in which the stiffness matrix is recalculated.

Generic Containment: Detailed comparison of containment simulations performed on plant scale

International Nuclear Information System (INIS)

Kelm, St.; Klauck, M.; Beck, S.; Allelein, H.-J.; Preusser, G.; Sangiorgi, M.; Klein-Hessling, W.; Bakalov, I.; Bleyer, A.; Bentaib, A.; Kljenak, I.; Stempniewicz, M.; Kostka, P.; Morandi, S.; Ada del Corno, B.; Bratfisch, C.; Risken, T.; Denk, L.; Parduba, Z.; Paci, S.

2014-01-01

Highlights: • Consequent implementation of the recommendations derived from the OECD/NEA ISP-47. • Phenomenological code-to-code comparison performed on plant scale. • Systematic identification and elimination of the user effect. • Identification of fundamental differences in the model basis. • Application to PAR system analysis. - Abstract: One outcome of the OECD/NEA ISP-47 activity was the recommendation to elaborate a ‘Generic Containment’ in order to allow comparing and rating the results obtained by different lumped-parameter models on plant scale. Within the European SARNET2 project ( (http://www.sar-net.eu)), such a Generic Containment nodalisation, based on a German PWR (1300 MW el ), was defined. This agreement on the nodalisation allows investigating the remaining differences among the results, especially the ‘user-effect’, related to the modelling choices, as well as fundamental differences in the underlying model basis in detail. The methodology applied in order to compare the different code predictions consisted of a series of three benchmark steps with increasing complexity as well as a systematic comparison of characteristic variables and observations. This paper summarises the benchmark series, the lessons learned during specifying the steps, comparing and discussing the results and finally gives an outlook on future steps

Comparison of Radiation Transport Codes, HZETRN, HETC and FLUKA, Using the 1956 Webber SPE Spectrum

Science.gov (United States)

Heinbockel, John H.; Slaba, Tony C.; Blattnig, Steve R.; Tripathi, Ram K.; Townsend, Lawrence W.; Handler, Thomas; Gabriel, Tony A.; Pinsky, Lawrence S.; Reddell, Brandon; Clowdsley, Martha S.;

Nondestructive testing standards and the ASME code

International Nuclear Information System (INIS)

Spanner, J.C.

1991-04-01

Nondestructive testing (NDT) requirements and standards are an important part of the ASME Boiler and Pressure Vessel Code. In this paper, the evolution of these requirements and standards is reviewed in the context of the unique technical and legal stature of the ASME Code. The coherent and consistent manner by which the ASME Code rules are organized is described, and the interrelationship between the various ASME Code sections, the piping codes, and the ASTM Standards is discussed. Significant changes occurred in ASME Sections 5 and 11 during the 1980s, and these are highlighted along with projections and comments regarding future trends and changes in these important documents. 4 refs., 8 tabs

Ensemble Weight Enumerators for Protograph LDPC Codes

Science.gov (United States)

Divsalar, Dariush

2006-01-01

Recently LDPC codes with projected graph, or protograph structures have been proposed. In this paper, finite length ensemble weight enumerators for LDPC codes with protograph structures are obtained. Asymptotic results are derived as the block size goes to infinity. In particular we are interested in obtaining ensemble average weight enumerators for protograph LDPC codes which have minimum distance that grows linearly with block size. As with irregular ensembles, linear minimum distance property is sensitive to the proportion of degree-2 variable nodes. In this paper the derived results on ensemble weight enumerators show that linear minimum distance condition on degree distribution of unstructured irregular LDPC codes is a sufficient but not a necessary condition for protograph LDPC codes.

Identification of coding and non-coding mutational hotspots in cancer genomes.

Science.gov (United States)

Piraino, Scott W; Furney, Simon J

2017-01-05

The identification of mutations that play a causal role in tumour development, so called "driver" mutations, is of critical importance for understanding how cancers form and how they might be treated. Several large cancer sequencing projects have identified genes that are recurrently mutated in cancer patients, suggesting a role in tumourigenesis. While the landscape of coding drivers has been extensively studied and many of the most prominent driver genes are well characterised, comparatively less is known about the role of mutations in the non-coding regions of the genome in cancer development. The continuing fall in genome sequencing costs has resulted in a concomitant increase in the number of cancer whole genome sequences being produced, facilitating systematic interrogation of both the coding and non-coding regions of cancer genomes. To examine the mutational landscapes of tumour genomes we have developed a novel method to identify mutational hotspots in tumour genomes using both mutational data and information on evolutionary conservation. We have applied our methodology to over 1300 whole cancer genomes and show that it identifies prominent coding and non-coding regions that are known or highly suspected to play a role in cancer. Importantly, we applied our method to the entire genome, rather than relying on predefined annotations (e.g. promoter regions) and we highlight recurrently mutated regions that may have resulted from increased exposure to mutational processes rather than selection, some of which have been identified previously as targets of selection. Finally, we implicate several pan-cancer and cancer-specific candidate non-coding regions, which could be involved in tumourigenesis. We have developed a framework to identify mutational hotspots in cancer genomes, which is applicable to the entire genome. This framework identifies known and novel coding and non-coding mutional hotspots and can be used to differentiate candidate driver regions from

LFSC - Linac Feedback Simulation Code

International Nuclear Information System (INIS)

Ivanov, Valentin; Fermilab

2008-01-01

The computer program LFSC ( ) is a numerical tool for simulation beam based feedback in high performance linacs. The code LFSC is based on the earlier version developed by a collective of authors at SLAC (L.Hendrickson, R. McEwen, T. Himel, H. Shoaee, S. Shah, P. Emma, P. Schultz) during 1990-2005. That code was successively used in simulation of SLC, TESLA, CLIC and NLC projects. It can simulate as pulse-to-pulse feedback on timescale corresponding to 5-100 Hz, as slower feedbacks, operating in the 0.1-1 Hz range in the Main Linac and Beam Delivery System. The code LFSC is running under Matlab for MS Windows operating system. It contains about 30,000 lines of source code in more than 260 subroutines. The code uses the LIAR ('Linear Accelerator Research code') for particle tracking under ground motion and technical noise perturbations. It uses the Guinea Pig code to simulate the luminosity performance. A set of input files includes the lattice description (XSIF format), and plane text files with numerical parameters, wake fields, ground motion data etc. The Matlab environment provides a flexible system for graphical output

Axisym finite element code: modifications for pellet-cladding mechanical interaction analysis

International Nuclear Information System (INIS)

Engelman, G.P.

1978-10-01

Local strain concentrations in nuclear fuel rods are known to be potential sites for failure initiation. Assessment of such strain concentrations requires a two-dimensional analysis of stress and strain in both the fuel and the cladding during pellet-cladding mechanical interaction. To provide such a capability in the FRAP (Fuel Rod Analysis Program) codes, the AXISYM code (a small finite element program developed at the Idaho National Engineering Laboratory) was modified to perform a detailed fuel rod deformation analysis. This report describes the modifications which were made to the AXISYM code to adapt it for fuel rod analysis and presents comparisons made between the two-dimensional AXISYM code and the FRACAS-II code. FRACAS-II is the one-dimensional (generalized plane strain) fuel rod mechanical deformation subcode used in the FRAP codes. Predictions of these two codes should be comparable away from the fuel pellet free ends if the state of deformation at the pellet midplane is near that of generalized plane strain. The excellent agreement obtained in these comparisons checks both the correctness of the AXISYM code modifications as well as the validity of the assumption of generalized plane strain upon which the FRACAS-II subcode is based

Cavitation Modeling in Euler and Navier-Stokes Codes

Science.gov (United States)

Deshpande, Manish; Feng, Jinzhang; Merkle, Charles L.

1993-01-01

Many previous researchers have modeled sheet cavitation by means of a constant pressure solution in the cavity region coupled with a velocity potential formulation for the outer flow. The present paper discusses the issues involved in extending these cavitation models to Euler or Navier-Stokes codes. The approach taken is to start from a velocity potential model to ensure our results are compatible with those of previous researchers and available experimental data, and then to implement this model in both Euler and Navier-Stokes codes. The model is then augmented in the Navier-Stokes code by the inclusion of the energy equation which allows the effect of subcooling in the vicinity of the cavity interface to be modeled to take into account the experimentally observed reduction in cavity pressures that occurs in cryogenic fluids such as liquid hydrogen. Although our goal is to assess the practicality of implementing these cavitation models in existing three-dimensional, turbomachinery codes, the emphasis in the present paper will center on two-dimensional computations, most specifically isolated airfoils and cascades. Comparisons between velocity potential, Euler and Navier-Stokes implementations indicate they all produce consistent predictions. Comparisons with experimental results also indicate that the predictions are qualitatively correct and give a reasonable first estimate of sheet cavitation effects in both cryogenic and non-cryogenic fluids. The impact on CPU time and the code modifications required suggests that these models are appropriate for incorporation in current generation turbomachinery codes.

Data exchange between zero dimensional code and physics platform in the CFETR integrated system code

Energy Technology Data Exchange (ETDEWEB)

Xu, Guoliang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Shi, Nan [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Zhou, Yifu; Mao, Shifeng [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Jian, Xiang [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronics Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chen, Jiale [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Liu, Li; Chan, Vincent [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Ye, Minyou, E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China)

2016-11-01

Highlights: • The workflow of the zero dimensional code and the multi-dimension physics platform of CFETR integrated system codeis introduced. • The iteration process among the codes in the physics platform. • The data transfer between the zero dimensionalcode and the physical platform, including data iteration and validation, and justification for performance parameters.. - Abstract: The China Fusion Engineering Test Reactor (CFETR) integrated system code contains three parts: a zero dimensional code, a physics platform and an engineering platform. We use the zero dimensional code to identify a set of preliminary physics and engineering parameters for CFETR, which is used as input to initiate multi-dimension studies using the physics and engineering platform for design, verification and validation. Effective data exchange between the zero dimensional code and the physical platform is critical for the optimization of CFETR design. For example, in evaluating the impact of impurity radiation on core performance, an open field line code is used to calculate the impurity transport from the first-wall boundary to the pedestal. The impurity particle in the pedestal are used as boundary conditions in a transport code for calculating impurity transport in the core plasma and the impact of core radiation on core performance. Comparison of the results from the multi-dimensional study to those from the zero dimensional code is used to further refine the controlled radiation model. The data transfer between the zero dimensional code and the physical platform, including data iteration and validation, and justification for performance parameters will be presented in this paper.

Fatigue and rupture codified rules comparison

International Nuclear Information System (INIS)

Faidy, C.

2004-01-01

The European Directive on Pressure Equipment requests risk studies and in particular to assure no risk of fatigue and rupture in operation. The answers to these questions are different in the different existing design codes (EN Standards, ASME III and VIII or RCC-M or CODAP-CODETI codes) and corresponding in operation codes (ASME or RSE-M). Design safety factors, material properties, fabrication, refinement in the analysis methods, monitoring in operation, hydro-proof test level... Around these Codes, different rules are under development. A16 in France, R6 in UK or FITNET at the EC level. This paper is concerned by a comparison between 2 different Codes to analyze the risk of fatigue or rupture of pressure equipments and mainly a comparison between RCC-M Code and EN 13445 standard for pressure vessel. Recommendations for future work will be proposed. (authors)

Comparison of computer codes (CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT) with data from the NRU-LOCA thermal hydraulic tests

International Nuclear Information System (INIS)

Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

1981-07-01

The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times

CoRoT/ESTA TASK 1 and TASK 3 comparison of the internal structure and seismic properties of representative stellar models. Comparisons between the ASTEC, CESAM, CLES, GARSTEC and STAROX codes

Science.gov (United States)

Lebreton, Yveline; Montalbán, Josefina; Christensen-Dalsgaard, Jørgen; Roxburgh, Ian W.; Weiss, Achim

2008-08-01

We compare stellar models produced by different stellar evolution codes for the CoRoT/ESTA project, comparing their global quantities, their physical structure, and their oscillation properties. We discuss the differences between models and identify the underlying reasons for these differences. The stellar models are representative of potential CoRoT targets. Overall we find very good agreement between the five different codes, but with some significant deviations. We find noticeable discrepancies (though still at the per cent level) that result from the handling of the equation of state, of the opacities and of the convective boundaries. The results of our work will be helpful in interpreting future asteroseismology results from CoRoT.

Model comparisons of the reactive burn model SURF in three ASC codes

Energy Technology Data Exchange (ETDEWEB)

Whitley, Von Howard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stalsberg, Krista Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reichelt, Benjamin Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Shipley, Sarah Jayne [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-01-12

A study of the SURF reactive burn model was performed in FLAG, PAGOSA and XRAGE. In this study, three different shock-to-detonation transition experiments were modeled in each code. All three codes produced similar model results for all the experiments modeled and at all resolutions. Buildup-to-detonation time, particle velocities and resolution dependence of the models was notably similar between the codes. Given the current PBX 9502 equations of state and SURF calibrations, each code is equally capable of predicting the correct detonation time and distance when impacted by a 1D impactor at pressures ranging from 10-16 GPa, as long as the resolution of the mesh is not too coarse.

CATHARE code development and assessment methodologies

International Nuclear Information System (INIS)

Micaelli, J.C.; Barre, F.; Bestion, D.

1995-01-01

The CATHARE thermal-hydraulic code has been developed jointly by Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), and Framatorne for safety analysis. Since the beginning of the project (September 1979), development and assessment activities have followed a methodology supported by two series of experimental tests: separate effects tests and integral effects tests. The purpose of this paper is to describe this methodology, the code assessment status, and the evolution to take into account two new components of this program: the modeling of three-dimensional phenomena and the requirements of code uncertainty evaluation

Development of a graphical interface computer code for reactor fuel reloading optimization

International Nuclear Information System (INIS)

Do Quang Binh; Nguyen Phuoc Lan; Bui Xuan Huy

2007-01-01

This report represents the results of the project performed in 2007. The aim of this project is to develop a graphical interface computer code that allows refueling engineers to design fuel reloading patterns for research reactor using simulated graphical model of reactor core. Besides, this code can perform refueling optimization calculations based on genetic algorithms as well as simulated annealing. The computer code was verified based on a sample problem, which relies on operational and experimental data of Dalat research reactor. This code can play a significant role in in-core fuel management practice at nuclear research reactor centers and in training. (author)

The InterFrost benchmark of Thermo-Hydraulic codes for cold regions hydrology - first inter-comparison results

Science.gov (United States)

Grenier, Christophe; Roux, Nicolas; Anbergen, Hauke; Collier, Nathaniel; Costard, Francois; Ferrry, Michel; Frampton, Andrew; Frederick, Jennifer; Holmen, Johan; Jost, Anne; Kokh, Samuel; Kurylyk, Barret; McKenzie, Jeffrey; Molson, John; Orgogozo, Laurent; Rivière, Agnès; Rühaak, Wolfram; Selroos, Jan-Olof; Therrien, René; Vidstrand, Patrik

2015-04-01

The impacts of climate change in boreal regions has received considerable attention recently due to the warming trends that have been experienced in recent decades and are expected to intensify in the future. Large portions of these regions, corresponding to permafrost areas, are covered by water bodies (lakes, rivers) that interact with the surrounding permafrost. For example, the thermal state of the surrounding soil influences the energy and water budget of the surface water bodies. Also, these water bodies generate taliks (unfrozen zones below) that disturb the thermal regimes of permafrost and may play a key role in the context of climate change. Recent field studies and modeling exercises indicate that a fully coupled 2D or 3D Thermo-Hydraulic (TH) approach is required to understand and model the past and future evolution of landscapes, rivers, lakes and associated groundwater systems in a changing climate. However, there is presently a paucity of 3D numerical studies of permafrost thaw and associated hydrological changes, and the lack of study can be partly attributed to the difficulty in verifying multi-dimensional results produced by numerical models. Numerical approaches can only be validated against analytical solutions for a purely thermic 1D equation with phase change (e.g. Neumann, Lunardini). When it comes to the coupled TH system (coupling two highly non-linear equations), the only possible approach is to compare the results from different codes to provided test cases and/or to have controlled experiments for validation. Such inter-code comparisons can propel discussions to try to improve code performances. A benchmark exercise was initialized in 2014 with a kick-off meeting in Paris in November. Participants from USA, Canada, Germany, Sweden and France convened, representing altogether 13 simulation codes. The benchmark exercises consist of several test cases inspired by existing literature (e.g. McKenzie et al., 2007) as well as new ones. They

KEY COMPARISON: Final report of EUROMET Project 696: pH determination of a phthalate buffer

Science.gov (United States)

Spitzer, Petra; Charlet, Philippe; Eberhard, Ralf; Karpov, Oleg V.; Philippe, Rachel; Rivier, Cedric; Maximov, Igor; Sudmeier, Uwe

2005-01-01

The EUROMET project 696, a trilateral comparison between PTB, Germany, LNE, France and VNIIFTRI, Russia was performed in order to demonstrate and document the capability of the participants to measure the pH of a phthalate buffer by the primary measurement procedure for pH. Good agreement of the reported results was observed. The sample was very similar to the one used in the comparison CCQM-K17. PTB acts as pilot laboratory in CCQM-K17 and in EUROMET 696. This comparison allows one to link the results obtained by LNE to the CCQM-K17 key comparison through the degree of equivalence of PTB. On the other hand, the discrepancy between measured pH values at the VNIIFTRI and PTB for the same type of buffer solution decreased, as compared with a bilateral comparison in 1997. Main text. To reach the main text of this paper, click on Final Report. The final report has been peer-reviewed and approved for publication by the CCQM, according to the provisions of the Mutual Recognition Arrangement (MRA).

Benchmarking (Code2Code) of the 1Hs 3-Bladed Onshore VAWT

DEFF Research Database (Denmark)

Galinos, Christos; Schmidt Paulsen, Uwe

This study is part of the Inflow project. In this report the Nenuphar’s onshore 3-bladed Vertical Axis Wind Turbine (VAWT) prototype (1HS) is modelled in HAWC2 aeroelastic code. In the first part the model properties are summarized. Then the analysis is focused on the rotor performance and various...

User instructions for the DESCARTES environmental accumulation code

International Nuclear Information System (INIS)

Miley, T.B.; Eslinger, P.W.; Nichols, W.E.; Lessor, K.S.; Ouderkirk, S.J.

1994-05-01

The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the Hanford Site near Richland, Washington. The HEDR Project work is conducted under several technical and administrative tasks, among which is the Environmental Pathways and Dose Estimates task. The staff on this task have developed a suite of computer codes which are used to estimate doses to individuals in the public. This document contains the user instructions for the DESCARTES (Dynamic estimates of concentrations and Accumulated Radionuclides in Terrestrial Environments) suite of codes. In addition to the DESCARTES code, this includes two air data preprocessors, a database postprocessor, and several utility routines that are used to format input data needed for DESCARTES

Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

Energy Technology Data Exchange (ETDEWEB)

Seo, Kwi Hyun; Song, Wan Jung [ENERGEO Inc. Sungnam, (Korea, Republic of); Song, Dong Soo; Byun, Choong Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2006-07-01

The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences.

Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

International Nuclear Information System (INIS)

Seo, Kwi Hyun; Song, Wan Jung; Song, Dong Soo; Byun, Choong Sup

2006-01-01

The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences

Applications of the Los Alamos High Energy Transport code

International Nuclear Information System (INIS)

Waters, L.; Gavron, A.; Prael, R.E.

1992-01-01

Simulation codes reliable through a large range of energies are essential to analyze the environment of vehicles and habitats proposed for space exploration. The LAHET monte carlo code has recently been expanded to track high energy hadrons with FLUKA, while retaining the original Los Alamos version of HETC at lower energies. Electrons and photons are transported with EGS4, and an interface to the MCNP monte carlo code is provided to analyze neutrons with kinetic energies less than 20 MeV. These codes are benchmarked by comparison of LAHET/MCNP calculations to data from the Brookhaven experiment E814 participant calorimeter

AECL's advanced code program

Energy Technology Data Exchange (ETDEWEB)

McGee, G.; Ball, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2012-07-01

This paper discusses the advanced code project at AECL.Current suite of Analytical, Scientific and Design (ASD) computer codes in use by Canadian Nuclear Power Industry is mostly developed 20 or more years ago. It is increasingly difficult to develop and maintain. It consist of many independent tools and integrated analysis is difficult, time consuming and error-prone. The objectives of this project is to demonstrate that nuclear facility systems, structures and components meet their design objectives in terms of function, cost, and safety; demonstrate that the nuclear facility meets licensing requirements in terms of consequences of off-normal events; dose to public, workers, impact on environment and demonstrate that the nuclear facility meets operational requirements with respect to on-power fuelling and outage management.

PACTEL OECD project planning (PACO). PACTEL OECD project planning

Energy Technology Data Exchange (ETDEWEB)

Kouhia, V.; Purhonen, H. [Lappeenranta University of Technology (Finland)

2004-07-01

OECD launched the SETH project to investigate issues relevant for accident prevention and management and to ensure the existence of integral thermal hydraulic test facilities. The facilities included in the SETH project are PKL from Germany and PANDA from Switzerland. At the early stages of the SETH project an idea was raised to exploit the PACTEL facility in a similar OECD project. Without any external funding the analytical work in the required extent would not be possible within Lappeenranta University of Technology, the party responsible of operating PACTEL. This fact directed the PACO project proposal to be conducted for the SAFIR programme. The aim of the PACO project is to prepare a project proposal to OECD of a PACTEL related project. To attain this objective some preliminary analyses have to be performed to ensure the relevancy of the proposed topic. The low power situation, i.e. midloop state was chosen to be the topic in the PACO studies and project planning basis. The plan is to use PACTEL to examine vertical steam generator behaviour during the midloop operation and the following loss of residual heat removal system transient. Such a possibility is acknowledged with special alterations to PACTEL. The APROS code version 5.04.07 was selected as a tool for the preanalyses. The virtual simulation of the chosen experimental situation would give a preconception on the phenomena to be expected and the progression of the transient. Originally the PACO project was planned to continue only for a few months, ending up with the project proposal to OECD during the summer time 2004. During the pre-calculation process it became obvious that the time expected was not enough to establish good pre-calculation results. The reasons for this relates to time used to learn and adapt the use of the chosen code, improvements and corrections in modelling as well as the code ability to manage the special conditions defined for the project topic. Another aspect on completing a

The WEST project mechanical analysis of the divertor structure according to the nuclear construction code

Energy Technology Data Exchange (ETDEWEB)

Larroque, S., E-mail: sebastien.larroque@cea.fr [CEA Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France); Portafaix, C. [ITER Organization, 13108 Saint-Paul-lez-Durance (France); Saille, A.; Doceul, L.; Bucalossi, J.; Samaille, F.; Freslon, S. de [CEA Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France)

2014-10-15

Highlights: • Divertor structure is mainly loaded by electromagnetical forces. • A simplified FEM analysis give the stresses in the structure. • RCCM criteria are required for the sizing. • Refined finite element models are used for local overstresses. - Abstract: The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST project, launched in support to the ITER tungsten divertor strategy. The installation of coils inside the vacuum vessel led to the design of a divertor supporting platform able to meet the project requirements and the associated electromagnetic loads. This paper illustrates the design, the method and the results of the thermomechanical elastic stress analyses performed in 2012. The validation of the integrity of the structure is based on the compliance with RCCMR design criteria (even though these Design and Construction rules for Mechanical Components of nuclear installations are not required for such experimental fusion device). Several 3D analyses are performed with the ANSYS code. The major one is a global analysis of half structure which determinates the stresses in the main part of the components. It gives an idea of the areas which needs local analyses. It also provides the interface loads for junction studies or simplified local model.

The code of ethics for nurses.

Science.gov (United States)

Zahedi, F; Sanjari, M; Aala, M; Peymani, M; Aramesh, K; Parsapour, A; Maddah, Ss Bagher; Cheraghi, Ma; Mirzabeigi, Gh; Larijani, B; Dastgerdi, M Vahid

2013-01-01

Nurses are ever-increasingly confronted with complex concerns in their practice. Codes of ethics are fundamental guidance for nursing as many other professions. Although there are authentic international codes of ethics for nurses, the national code would be the additional assistance provided for clinical nurses in their complex roles in care of patients, education, research and management of some parts of health care system in the country. A national code can provide nurses with culturally-adapted guidance and help them to make ethical decisions more closely to the Iranian-Islamic background. Given the general acknowledgement of the need, the National Code of Ethics for Nurses was compiled as a joint project (2009-2011). The Code was approved by the Health Policy Council of the Ministry of Health and Medical Education and communicated to all universities, healthcare centers, hospitals and research centers early in 2011. The focus of this article is on the course of action through which the Code was compiled, amended and approved. The main concepts of the code will be also presented here. No doubt, development of the codes should be considered as an ongoing process. This is an overall responsibility to keep the codes current, updated with the new progresses of science and emerging challenges, and pertinent to the nursing practice.

Toward a CFD nose-to-tail capability - Hypersonic unsteady Navier-Stokes code validation

Science.gov (United States)

Edwards, Thomas A.; Flores, Jolen

1989-01-01

Computational fluid dynamics (CFD) research for hypersonic flows presents new problems in code validation because of the added complexity of the physical models. This paper surveys code validation procedures applicable to hypersonic flow models that include real gas effects. The current status of hypersonic CFD flow analysis is assessed with the Compressible Navier-Stokes (CNS) code as a case study. The methods of code validation discussed to beyond comparison with experimental data to include comparisons with other codes and formulations, component analyses, and estimation of numerical errors. Current results indicate that predicting hypersonic flows of perfect gases and equilibrium air are well in hand. Pressure, shock location, and integrated quantities are relatively easy to predict accurately, while surface quantities such as heat transfer are more sensitive to the solution procedure. Modeling transition to turbulence needs refinement, though preliminary results are promising.

Micromagnetic Code Development of Advanced Magnetic Structures Final Report CRADA No. TC-1561-98

Energy Technology Data Exchange (ETDEWEB)

Cerjan, Charles J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shi, Xizeng [Read-Rite Corporation, Fremont, CA (United States)

2017-11-09

The specific goals of this project were to: Further develop the previously written micromagnetic code DADIMAG (DOE code release number 980017); Validate the code. The resulting code was expected to be more realistic and useful for simulations of magnetic structures of specific interest to Read-Rite programs. We also planned to further the code for use in internal LLNL programs. This project complemented LLNL CRADA TC-840-94 between LLNL and Read-Rite, which allowed for simulations of the advanced magnetic head development completed under the CRADA. TC-1561-98 was effective concurrently with LLNL non-exclusive copyright license (TL-1552-98) to Read-Rite for DADIMAG Version 2 executable code.

Code Team Training: Demonstrating Adherence to AHA Guidelines During Pediatric Code Blue Activations.

Science.gov (United States)

Stewart, Claire; Shoemaker, Jamie; Keller-Smith, Rachel; Edmunds, Katherine; Davis, Andrew; Tegtmeyer, Ken

2017-10-16

Pediatric code blue activations are infrequent events with a high mortality rate despite the best effort of code teams. The best method for training these code teams is debatable; however, it is clear that training is needed to assure adherence to American Heart Association (AHA) Resuscitation Guidelines and to prevent the decay that invariably occurs after Pediatric Advanced Life Support training. The objectives of this project were to train a multidisciplinary, multidepartmental code team and to measure this team's adherence to AHA guidelines during code simulation. Multidisciplinary code team training sessions were held using high-fidelity, in situ simulation. Sessions were held several times per month. Each session was filmed and reviewed for adherence to 5 AHA guidelines: chest compression rate, ventilation rate, chest compression fraction, use of a backboard, and use of a team leader. After the first study period, modifications were made to the code team including implementation of just-in-time training and alteration of the compression team. Thirty-eight sessions were completed, with 31 eligible for video analysis. During the first study period, 1 session adhered to all AHA guidelines. During the second study period, after alteration of the code team and implementation of just-in-time training, no sessions adhered to all AHA guidelines; however, there was an improvement in percentage of sessions adhering to ventilation rate and chest compression rate and an improvement in median ventilation rate. We present a method for training a large code team drawn from multiple hospital departments and a method of assessing code team performance. Despite subjective improvement in code team positioning, communication, and role completion and some improvement in ventilation rate and chest compression rate, we failed to consistently demonstrate improvement in adherence to all guidelines.

ClinicalCodes: an online clinical codes repository to improve the validity and reproducibility of research using electronic medical records.

Science.gov (United States)

Springate, David A; Kontopantelis, Evangelos; Ashcroft, Darren M; Olier, Ivan; Parisi, Rosa; Chamapiwa, Edmore; Reeves, David

2014-01-01

Lists of clinical codes are the foundation for research undertaken using electronic medical records (EMRs). If clinical code lists are not available, reviewers are unable to determine the validity of research, full study replication is impossible, researchers are unable to make effective comparisons between studies, and the construction of new code lists is subject to much duplication of effort. Despite this, the publication of clinical codes is rarely if ever a requirement for obtaining grants, validating protocols, or publishing research. In a representative sample of 450 EMR primary research articles indexed on PubMed, we found that only 19 (5.1%) were accompanied by a full set of published clinical codes and 32 (8.6%) stated that code lists were available on request. To help address these problems, we have built an online repository where researchers using EMRs can upload and download lists of clinical codes. The repository will enable clinical researchers to better validate EMR studies, build on previous code lists and compare disease definitions across studies. It will also assist health informaticians in replicating database studies, tracking changes in disease definitions or clinical coding practice through time and sharing clinical code information across platforms and data sources as research objects.

Development of multi-physics code systems based on the reactor dynamics code DYN3D

Energy Technology Data Exchange (ETDEWEB)

Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

2011-07-15

The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

Development of multi-physics code systems based on the reactor dynamics code DYN3D

International Nuclear Information System (INIS)

Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

2011-01-01

The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

Repository seal materials performance for a SALT Repository Project 5-year code/model development plan: Draft

International Nuclear Information System (INIS)

1987-06-01

This document describes an integrated laboratory testing and model development effort for the seal system for a high-level nuclear waste repository in salt. The testing and modeling efforts are designed to determine seal material response in the repository environment, to provide models of seal system components for performance assessment, and to assist in the development of seal system designs. A code/model development and performance analysis program will be performed to predict the short- and long-term response of seal materials and seal components. The results from these analyses will be used to support the material testing activities on this contract and to support performance assessment activities that are conducted in other parts of the Salt Repository Project (SRP). 48 refs., 15 figs., 4 tabs

10Gbps 2D MGC OCDMA Code over FSO Communication System

Science.gov (United States)

Professor Urmila Bhanja, Associate, Dr.; Khuntia, Arpita; Alamasety Swati, (Student

2017-08-01

Currently, wide bandwidth signal dissemination along with low latency is a leading requisite in various applications. Free space optical wireless communication has introduced as a realistic technology for bridging the gap in present high data transmission fiber connectivity and as a provisional backbone for rapidly deployable wireless communication infrastructure. The manuscript highlights on the implementation of 10Gbps SAC-OCDMA FSO communications using modified two dimensional Golomb code (2D MGC) that possesses better auto correlation, minimum cross correlation and high cardinality. A comparison based on pseudo orthogonal (PSO) matrix code and modified two dimensional Golomb code (2D MGC) is developed in the proposed SAC OCDMA-FSO communication module taking different parameters into account. The simulative outcome signifies that the communication radius is bounded by the multiple access interference (MAI). In this work, a comparison is made in terms of bit error rate (BER), and quality factor (Q) based on modified two dimensional Golomb code (2D MGC) and PSO matrix code. It is observed that the 2D MGC yields better results compared to the PSO matrix code. The simulation results are validated using optisystem version 14.

FONESYS: The FOrum and NEtwork of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics

Energy Technology Data Exchange (ETDEWEB)

Ahn, S.H., E-mail: k175ash@kins.re.kr [Korea Institute of Nuclear Safety (KINS) (Korea, Republic of); Aksan, N., E-mail: nusr.aksan@gmail.com [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Austregesilo, H., E-mail: henrique.austregesilo@grs.de [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Bestion, D., E-mail: dominique.bestion@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Chung, B.D., E-mail: bdchung@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); D’Auria, F., E-mail: f.dauria@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Emonot, P., E-mail: philippe.emonot@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Gandrille, J.L., E-mail: jeanluc.gandrille@areva.com [AREVA NP (France); Hanninen, M., E-mail: markku.hanninen@vtt.fi [VTT Technical Research Centre of Finland (VTT) (Finland); Horvatović, I., E-mail: i.horvatovic@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Kim, K.D., E-mail: kdkim@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); Kovtonyuk, A., E-mail: a.kovtonyuk@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Petruzzi, A., E-mail: a.petruzzi@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy)

2015-01-15

Highlights: • We briefly presented the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). • We presented FONESYS participants and their codes. • We explained FONESYS projects motivation, its main targets and working modalities. • We presented FONESYS position about projects topics and subtopics. - Abstract: The purpose of this article is to present briefly the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS), its participants, the motivation for the project, its main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor technology, safety and design. Namely, the number of code-users can easily be predicted to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework the FONESYS initiative has been created. The main targets of FONESYS are: • To promote the use of SYS-TH Codes and the application of the BEPU approaches. • To establish acceptable and recognized procedures and thresholds for Verification and Validation (V and V). • To create a common ground for discussing envisaged improvements in various areas, including user-interface, and the connection with other numerical tools, including Computational Fluid Dynamics (CFD) Codes.

Rapid installation of numerical models in multiple parent codes

Energy Technology Data Exchange (ETDEWEB)

Brannon, R.M.; Wong, M.K.

1996-10-01

A set of``model interface guidelines``, called MIG, is offered as a means to more rapidly install numerical models (such as stress-strain laws) into any parent code (hydrocode, finite element code, etc.) without having to modify the model subroutines. The model developer (who creates the model package in compliance with the guidelines) specifies the model`s input and storage requirements in a standardized way. For portability, database management (such as saving user inputs and field variables) is handled by the parent code. To date, NUG has proved viable in beta installations of several diverse models in vectorized and parallel codes written in different computer languages. A NUG-compliant model can be installed in different codes without modifying the model`s subroutines. By maintaining one model for many codes, MIG facilitates code-to-code comparisons and reduces duplication of effort potentially reducing the cost of installing and sharing models.

Method for quantitative assessment of nuclear safety computer codes

International Nuclear Information System (INIS)

Dearien, J.A.; Davis, C.B.; Matthews, L.J.

1979-01-01

A procedure has been developed for the quantitative assessment of nuclear safety computer codes and tested by comparison of RELAP4/MOD6 predictions with results from two Semiscale tests. This paper describes the developed procedure, the application of the procedure to the Semiscale tests, and the results obtained from the comparison

Accuracy assessment of a new Monte Carlo based burnup computer code

International Nuclear Information System (INIS)

El Bakkari, B.; ElBardouni, T.; Nacir, B.; ElYounoussi, C.; Boulaich, Y.; Meroun, O.; Zoubair, M.; Chakir, E.

2012-01-01

Highlights: ► A new burnup code called BUCAL1 was developed. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► Validation of BUCAL1 was done by code to code comparison using VVER-1000 LEU Benchmark Assembly. ► Differences from BM value were found to be ± 600 pcm for k ∞ and ±6% for the isotopic compositions. ► The effect on reactivity due to the burnup of Gd isotopes is well reproduced by BUCAL1. - Abstract: This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a “VVER-1000 LEU Assembly Computational Benchmark”, at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.

Analysis of ATLAS FLB-EC6 Experiment using SPACE Code

International Nuclear Information System (INIS)

Lee, Donghyuk; Kim, Yohan; Kim, Seyun

2013-01-01

The new code is named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). As a part of code validation effort, simulation of ATLAS FLB(Feedwater Line Break) experiment using SPACE code has been performed. The FLB-EC6 experiment is economizer break of a main feedwater line. The calculated results using the SPACE code are compared with those from the experiment. The ATLAS FLB-EC6 experiment, which is economizer feedwater line break, was simulated using the SPACE code. The calculated results were compared with those from the experiment. The comparisons of break flow rate and steam generator water level show good agreement with the experiment. The SPACE code is capable of predicting physical phenomena occurring during ATLAS FLB-EC6 experiment

Gamma streaming experiments for validation of Monte Carlo code

International Nuclear Information System (INIS)

Thilagam, L.; Mohapatra, D.K.; Subbaiah, K.V.; Iliyas Lone, M.; Balasubramaniyan, V.

2012-01-01

In-homogeneities in shield structures lead to considerable amount of leakage radiation (streaming) increasing the radiation levels in accessible areas. Development works on experimental as well as computational methods for quantifying this streaming radiation are still continuing. Monte Carlo based radiation transport code, MCNP is usually a tool for modeling and analyzing such problems involving complex geometries. In order to validate this computational method for streaming analysis, it is necessary to carry out some experimental measurements simulating these inhomogeneities like ducts and voids present in the bulk shields for typical cases. The data thus generated will be analysed by simulating the experimental set up employing MCNP code and optimized input parameters for the code in finding solutions for similar radiation streaming problems will be formulated. Comparison of experimental data obtained from radiation streaming experiments through ducts will give a set of thumb rules and analytical fits for total radiation dose rates within and outside the duct. The present study highlights the validation of MCNP code through the gamma streaming experiments carried out with the ducts of various shapes and dimensions. Over all, the present study throws light on suitability of MCNP code for the analysis of gamma radiation streaming problems for all duct configurations considered. In the present study, only dose rate comparisons have been made. Studies on spectral comparison of streaming radiation are in process. Also, it is planned to repeat the experiments with various shield materials. Since the penetrations and ducts through bulk shields are unavoidable in an operating nuclear facility the results on this kind of radiation streaming simulations and experiments will be very useful in the shield structure optimization without compromising the radiation safety

In-situ recording of ionic currents in projection neurons and Kenyon cells in the olfactory pathway of the honeybee.

Science.gov (United States)

Kropf, Jan; Rössler, Wolfgang

2018-01-01

The honeybee olfactory pathway comprises an intriguing pattern of convergence and divergence: ~60.000 olfactory sensory neurons (OSN) convey olfactory information on ~900 projection neurons (PN) in the antennal lobe (AL). To transmit this information reliably, PNs employ relatively high spiking frequencies with complex patterns. PNs project via a dual olfactory pathway to the mushroom bodies (MB). This pathway comprises the medial (m-ALT) and the lateral antennal lobe tract (l-ALT). PNs from both tracts transmit information from a wide range of similar odors, but with distinct differences in coding properties. In the MBs, PNs form synapses with many Kenyon cells (KC) that encode odors in a spatially and temporally sparse way. The transformation from complex information coding to sparse coding is a well-known phenomenon in insect olfactory coding. Intrinsic neuronal properties as well as GABAergic inhibition are thought to contribute to this change in odor representation. In the present study, we identified intrinsic neuronal properties promoting coding differences between PNs and KCs using in-situ patch-clamp recordings in the intact brain. We found very prominent K+ currents in KCs clearly differing from the PN currents. This suggests that odor coding differences between PNs and KCs may be caused by differences in their specific ion channel properties. Comparison of ionic currents of m- and l-ALT PNs did not reveal any differences at a qualitative level.

SU-E-T-323: The FLUKA Monte Carlo Code in Ion Beam Therapy

Energy Technology Data Exchange (ETDEWEB)

Rinaldi, I [Heidelberg University Hospital (Germany); Ludwig-Maximilian University Munich (Germany)

2014-06-01

Purpose: Monte Carlo (MC) codes are increasingly used in the ion beam therapy community due to their detailed description of radiation transport and interaction with matter. The suitability of a MC code demands accurate and reliable physical models for the transport and the interaction of all components of the mixed radiation field. This contribution will address an overview of the recent developments in the FLUKA code oriented to its application in ion beam therapy. Methods: FLUKA is a general purpose MC code which allows the calculations of particle transport and interactions with matter, covering an extended range of applications. The user can manage the code through a graphic interface (FLAIR) developed using the Python programming language. Results: This contribution will present recent refinements in the description of the ionization processes and comparisons between FLUKA results and experimental data of ion beam therapy facilities. Moreover, several validations of the largely improved FLUKA nuclear models for imaging application to treatment monitoring will be shown. The complex calculation of prompt gamma ray emission compares favorably with experimental data and can be considered adequate for the intended applications. New features in the modeling of proton induced nuclear interactions also provide reliable cross section predictions for the production of radionuclides. Of great interest for the community are the developments introduced in FLAIR. The most recent efforts concern the capability of importing computed-tomography images in order to build automatically patient geometries and the implementation of different types of existing positron-emission-tomography scanner devices for imaging applications. Conclusion: The FLUA code has been already chosen as reference MC code in many ion beam therapy centers, and is being continuously improved in order to match the needs of ion beam therapy applications. Parts of this work have been supported by the European

Projection-based curve clustering

International Nuclear Information System (INIS)

Auder, Benjamin; Fischer, Aurelie

2012-01-01

This paper focuses on unsupervised curve classification in the context of nuclear industry. At the Commissariat a l'Energie Atomique (CEA), Cadarache (France), the thermal-hydraulic computer code CATHARE is used to study the reliability of reactor vessels. The code inputs are physical parameters and the outputs are time evolution curves of a few other physical quantities. As the CATHARE code is quite complex and CPU time-consuming, it has to be approximated by a regression model. This regression process involves a clustering step. In the present paper, the CATHARE output curves are clustered using a k-means scheme, with a projection onto a lower dimensional space. We study the properties of the empirically optimal cluster centres found by the clustering method based on projections, compared with the 'true' ones. The choice of the projection basis is discussed, and an algorithm is implemented to select the best projection basis among a library of orthonormal bases. The approach is illustrated on a simulated example and then applied to the industrial problem. (authors)

Comparison report of the OECD/CSNI international standard problem 21 (Piper-one experiment PO-SB-7). Volume 1 comparison report. Volume 2 evaluation of code accuracy in the prediction of ISP 21