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Sample records for mod2 computer codes

  1. VARSKIN MOD 2 and SADDE MOD2: Computer codes for assessing skin dose from skin contamination

    International Nuclear Information System (INIS)

    Durham, J.S.

    1992-12-01

    The computer code VARSKIN has been modified to calculate dose to skin from three-dimensional sources, sources separated from the skin by layers of protective clothing, and gamma dose from certain radionuclides correction for backscatter has also been incorporated for certain geometries. This document describes the new code, VARSKIN Mod 2, including installation and operation instructions, provides detailed descriptions of the models used, and suggests methods for avoiding misuse of the code. The input data file for VARSKIN Mod 2 has been modified to reflect current physical data, to include the contribution to dose from internal conversion and Auger electrons, and to reflect a correction for low-energy electrons. In addition, the computer code SADDE: Scaled Absorbed Dose Distribution Evaluator has been modified to allow the generation of scaled absorbed dose distributions for mixtures of radionuclides and intereat conversion and Auger electrons. This new code, SADDE Mod 2, is also described in this document. Instructions for installation and operation of the code and detailed descriptions of the models used in the code are provided

  2. Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

    1985-04-01

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

  3. Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

    International Nuclear Information System (INIS)

    Megahed, M.M.

    1992-01-01

    An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

  4. Hot particle dose calculations using the computer code VARSKIN Mod 2

    International Nuclear Information System (INIS)

    Durham, J.S.

    1991-01-01

    The only calculational model recognised by the Nuclear Regulatory Commission (NRC) for hot particle dosimetry is VARSKIN Mod 1. Because the code was designed to calculate skin dose from distributed skin contamination and not hot particles, it is assumed that the particle has no thickness and, therefore, that no self-absorption occurs within the source material. For low energy beta particles such as those emitted from 60 Co, a significant amount of self-shielding occurs in hot particles and VARSKIN Mod 1 overestimates the skin dose. In addition, the presence of protective clothing, which will reduce the calculated skin dose for both high and low energy beta emitters, is not modelled in VARSKIN Mod 1. Finally, there is no provision in VARSKIN Mod 1 to calculate the gamma contribution to skin dose from radionuclides that emit both beta and gamma radiation. The computer code VARSKIN Mod 1 has been modified to model three-dimensional sources, insertion of layers of protective clothing between the source and skin, and gamma dose from appropriate radionuclides. The new code, VARSKIN Mod 2, is described and the sensitivity of the calculated dose to source geometry, diameter, thickness, density, and protective clothing thickness are discussed. Finally, doses calculated using VARSKIN Mod 2 are compared to doses measured from hot particles found in nuclear power plants. (author)

  5. RELAP5/MOD2 code assessment

    International Nuclear Information System (INIS)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

    1985-01-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G

  6. RELAP5/MOD2 code assessment

    Energy Technology Data Exchange (ETDEWEB)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

    1985-11-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G.

  7. Qualification of the AUTOBUS Mod. 2 Code

    International Nuclear Information System (INIS)

    Ciarniello, U.; Peroni, P.

    1988-01-01

    The paper presents the qualification of AUTOBUS MOD.2 code. After a brief description of the code itself, all the critical experiments simulated by the code are illustrated to prove the accuracy of criticality calculation and power distribution. An interpretation of the results and a conclusion close this presentation

  8. Using computer program RELAP5/MOD2 on microcomputers

    International Nuclear Information System (INIS)

    Grgic, D.; Bajs, T; Cavlina, N.; Debrecin, N.

    1990-01-01

    Our work on installation of RELAP5/MOD2 code on IBM4341, mVAX 11, MGT-386 and COMPAQ-386/20e computers is described. Main characteristics of RELAP5/MOD2 structure programming style and differences between FORTRAN VS, VAX-11 FORTRAN and NDP FORTRAN 386 are presented. We discussed basic philosophy used in modification and testing and test results. (author)

  9. Steady-state simulations of a 30-tube once-through steam generator with the RELAP5/MOD3 and RELAP5/MOD2 computer codes

    International Nuclear Information System (INIS)

    Hassan, Y.A.; Salim, P.

    1991-01-01

    This paper reports on a steady-state analysis of a 30-tube once-through steam generator that has been performed on the RELAPS/MOD3 and RELAPS/MOD2 computer codes for 100, 75, and 65% loads. Results obtained are compared with experimental data. The RELAP5/MOD3 results for the test facility generally agree reasonably well with the data for the primary-side temperature profiles. The secondary-side temperature profile predicted by RELAP5/MOD3 at 75 and 65% loads agrees fairly well with the data and is better than the RELAP5/MOD2 results. However, the RELAP5/MOD3 calculated secondary-side temperature profile does not compare well with the 100% load data

  10. THYDE-B1/MOD2: a computer code for analysis of small-break loss-of-coolant accidents of boiling water reactors

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Muramatsu, Ken; Kukita, Yutaka; Tasaka, Kanji

    1988-04-01

    THYDE-B1/MOD2 is a fast-running best estimate (BE) computer code to analyze thermal-hydraulic behaviors of the reactor cooling system of a boiling water reactor (BWR), mainly, during a small-break loss-of-coolant accident (SBLOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions consist of subcooled liquid, saturated mixture and saturated steam regions from the volume bottom. The regions are separated by two horizontal moving boundaries which are tracked by mass and energy balances for each region. With this three region node model, the interior of the pressure vessel can be represented by only two volumes: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous node model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SBLOCAs in which the thermal-hydraulic behavior is relatively slow and gravity controlled. The code has been improved and modified from the last version of the code, THYDE-B1/MOD1, especially in the phase separation model which is used in the mixture level calculation in the three region node model. Then, a good predictability of the code has been indicated through the comparison of calculated results with various SBLOCA test data including ROSA-III of JAERI and FIST of the General Electric Co. This report presents the code modifications and input data requirements of the THYDE-B1/MOD2 code. (author)

  11. An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

    International Nuclear Information System (INIS)

    Hall, D.G.; Watkins, J.C.

    1987-01-01

    This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

  12. The WECHSL-Mod2 code: A computer program for the interaction of a core melt with concrete including the long term behavior

    International Nuclear Information System (INIS)

    Reimann, M.; Stiefel, S.

    1989-06-01

    The WECHSL-Mod2 code is a mechanistic computer code developed for the analysis of the thermal and chemical interaction of initially molten LWR reactor materials with concrete in a two-dimensional, axisymmetrical concrete cavity. The code performs calculations from the time of initial contact of a hot molten pool over start of solidification processes until long term basemat erosion over several days with the possibility of basemat penetration. The code assumes that the metallic phases of the melt pool form a layer at the bottom overlayed by the oxide melt atop. Heat generation in the melt is by decay heat and chemical reactions from metal oxidation. Energy is lost to the melting concrete and to the upper containment by radiation or evaporation of sumpwater possibly flooding the surface of the melt. Thermodynamic and transport properties as well as criteria for heat transfer and solidification processes are internally calculated for each time step. Heat transfer is modelled taking into account the high gas flux from the decomposing concrete and the heat conduction in the crusts possibly forming in the long term at the melt/concrete interface. The WECHSL code in its present version was validated by the BETA experiments. The test samples include a typical BETA post test calculation and a WECHSL application to a reactor accident. (orig.) [de

  13. Analysis of the reflood experiment by RELAP5/MOD2 code

    International Nuclear Information System (INIS)

    Prosek, A.; Stritar, A.

    1990-01-01

    The analysis of the reflood experiment on the test rig Achilles has been performed. The analysis has been done by the RELAP5/MOD2 code after the results of the experiment had been released. The experiment has been analyze in several other laboratories around the world. Our results are comparable to other analyses and are in the range of RELAP5/MOD2 capabilities. Two analyses have been done: the core only and the complete system. Computed clad temperatures in the first case are higher than measured, in the second case they are somewhat lower. (author)

  14. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code

    International Nuclear Information System (INIS)

    Ramsthaler, J.A.; Lime, J.F.; Sahota, M.S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A

  15. RELAP5/MOD2 code assessment for the Semiscale Mod-2C Test S-LH-1

    International Nuclear Information System (INIS)

    Fineman, C.P.

    1986-01-01

    RELAP5/MOD2, Cycle 36.02, was assessed using data from Semiscale Mod-2C experiment S-LH-1. The major phenomena that occurred during the experiment were calculated by RELAP5/MOD2, although the duration and the magnitude of their effect on the transient were not always well calculated. Areas defined where further work was needed to improve the RELAP5 calculation include: (1) the system energy balance, (2) core interfacial drag, and 3) the heat transfer logic rod dryout criterion

  16. Vectorization, parallelization and implementation of nuclear codes =MVP/GMVP, QMDRELP, EQMD, HSABC, CURBAL, STREAM V3.1, TOSCA, EDDYCAL, RELAP5/MOD2/C36-05, RELAP5/MOD3= on the VPP500 computer system. Progress report 1995 fiscal year

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Toshiyuki; Watanabe, Hideo; Fujita, Toyozo [Fujitsu Ltd., Tokyo (Japan); Kawai, Wataru; Harada, Hiroo; Gorai, Kazuo; Yamasaki, Kazuhiko; Shoji, Makoto; Fujii, Minoru

    1996-06-01

    At Center for Promotion of Computational Science and Engineering, time consuming eight nuclear codes suggested by users have been vectorized, parallelized on the VPP500 computer system. In addition, two nuclear codes used on the VP2600 computer system were implemented on the VPP500 computer system. Neutron and photon transport calculation code MVP/GMVP and relativistic quantum molecular dynamics code QMDRELP have been parallelized. Extended quantum molecular dynamics code EQMD and adiabatic base calculation code HSABC have been parallelized and vectorized. Ballooning turbulence simulation code CURBAL, 3-D non-stationary compressible fluid dynamics code STREAM V3.1, operating plasma analysis code TOSCA and eddy current analysis code EDDYCAL have been vectorized. Reactor safety analysis code RELAP5/MOD2/C36-05 and RELAP5/MOD3 were implemented on the VPP500 computer system. (author)

  17. Vectorization, parallelization and implementation of nuclear codes [MVP/GMVP, QMDRELP, EQMD, HSABC, CURBAL, STREAM V3.1, TOSCA, EDDYCAL, RELAP5/MOD2/C36-05, RELAP5/MOD3] on the VPP500 computer system. Progress report 1995 fiscal year

    International Nuclear Information System (INIS)

    Nemoto, Toshiyuki; Watanabe, Hideo; Fujita, Toyozo; Kawai, Wataru; Harada, Hiroo; Gorai, Kazuo; Yamasaki, Kazuhiko; Shoji, Makoto; Fujii, Minoru.

    1996-07-01

    At Center for Promotion of Computational Science and Engineering, time consuming eight nuclear codes suggested by users have been vectorized, parallelized on the VPP500 computer system. In addition, two nuclear codes used on the VP2600 computer system were implemented on the VPP500 computer system. Neutron and photon transport calculation code MVP/GMVP and relativistic quantum molecular dynamics code QMDRELP have been parallelized. Extended quantum molecular dynamics code EQMD and adiabatic base calculation code HSABC have been parallelized and vectorized. Ballooning turbulence simulation code CURBAL, 3-D non-stationary compressible fluid dynamics code STREAM V3.1, operating plasma analysis code TOSCA and eddy current analysis code EDDYCAL have been vectorized. Reactor safety analysis code RELAP5/MOD2/C36-05 and RELAP5/MOD3 were implemented on the VPP500 computer system. (author)

  18. FUDA MOD-2: a computer program for simulation the performance of fuel element validation exercise

    International Nuclear Information System (INIS)

    Chouhan, S.K.; Tripathi, R.M.; Prasad, P.N.; Chauhan, Ashok

    2014-01-01

    The PHWR fuel element performance is evaluated using the fuel analysis computer code FUDA MOD2. It is specifically written for performance simulation of UO 2 fuel pellet, located in zirconium alloy sheath operating under coolant pressure. For specific element power histories, the code investigates the variables and their interactions that govern fuel element performance. The input data requires pellet dimensions, element dimensions, sheath properties, heat transfer data, thermal hydraulic parameters of coolant, the inner filler gas composition, flux gradient and linear heat ratings (LHR) at different burn up. The output data generated by the code are radial temperature profile of fuel and sheath, fuel sheath-gap heat transfer coefficient, fission gas generated and released, fission gas pressure, sheath stress and strain for different burn-up zones. The code has been verified against literature data and post irradiation examinations carried out. It is also bench marked against various international fuel element simulation programmes available with water cooled reactors operating countries. The present paper describes the FUDA MOD2 code verification studies carried out using the literature data and post irradiation examination data. (author)

  19. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    Allison, C.M.; Johnson, E.C.

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and II to acquaint the user with the modeling base and thus aid in effective use of the code

  20. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    Allison, C.M.; Johnson, E.C.

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and in this document, Volume II, to acquaint the user with the modeling base and thus aid in effective use of the code. 135 refs., 48 figs., 8 tabs

  1. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    Hohorst, J.K.

    1990-02-01

    This report describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various light water reactor (LWR) accident analysis computer programs. Formulation of the materials properties are generally semiempirical in nature. The materials properties subcodes contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, and fill gas mixtures. 452 refs., 230 figs., 139 tabs

  2. SCDAP/RELAP5/MOD2 code manual

    Energy Technology Data Exchange (ETDEWEB)

    Hohorst, J.K. (ed.) (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-02-01

    This report describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various light water reactor (LWR) accident analysis computer programs. Formulation of the materials properties are generally semiempirical in nature. The materials properties subcodes contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, and fill gas mixtures. 452 refs., 230 figs., 139 tabs.

  3. RELAP5/MOD2 code modifications to obtain better predictions for the once-through steam generator

    International Nuclear Information System (INIS)

    Blanchat, T.; Hassan, Y.

    1989-01-01

    The steam generator is a major component in pressurized water reactors. Predicting the response of a steam generator during both steady-state and transient conditions is essential in studying the thermal-hydraulic behavior of a nuclear reactor coolant system. Therefore, many analytical and experimental efforts have been performed to investigate the thermal-hydraulic behavior of the steam generators during operational and accident transients. The objective of this study is to predict the behavior of the secondary side of the once-through steam generator (OTSG) using the RELAP5/MOD2 computer code. Steady-state conditions were predicted with the current version of the RELAP5/MOD2 code and compared with experimental plant data. The code predictions consistently underpredict the degree of superheat. A new interface friction model has been implemented in a modified version of RELAP5/MOD2. This modification, along with changes to the flow regime transition criteria and the heat transfer correlations, correctly predicts the degree of superheat and matches plant data

  4. Double blind post-test prediction for LOBI-MOD2 small break experiment A2-81 using RELAP5/MOD1/19 computer code as contribution to international CSNI-standardproblem no. 18

    International Nuclear Information System (INIS)

    Jacobs, G.; Mansoor, S.H.

    1986-06-01

    The first small break experiment A2-81 performed in the LOBI-MOD2 test facility was the base of the 18th international CSNI standard problem (ISP 18). Taking part in this exercise, a blind post-test prediction was performed using the light water reactor transient analysis code RELAP5/MOD1. This paper describes the input model preparation and summarizes the findings of the pre-calculation comparing the calculational results with the experimental data. The results show that there was a good agreement between prediction and experiment in the initial stage (up to 250 sec) of the transient and an adequate prediction of the global behaviour (thermal response of the core), which is important for safety related considerations. However, the prediction confirmed some deficiencies of the models in the code concerning vertical and horizontal stratification resulting in a high break mass flow and an erroneous distribution of mass over the primary loops. (orig.) [de

  5. Assessment of RELAP5/MOD2 and RELAP5/MOD1-EUR codes on the basis of LOBI-MOD2 test results

    International Nuclear Information System (INIS)

    D'Auria, F.; Mazzini, M.; Oriolo, F.; Galassi, G.M.

    1989-10-01

    The present report deals with an overview of the application of RELAP5/MOD2 and RELAP5/MOD1-EUR codes to tests performed in the LOBI/MOD2 facility. The work has been carried out in the frame of a contract between Dipartimento di Costruzioni Meccaniche e Nucleari (DCMN) of Pisa University and CEC. The Universities of Roma, Pisa, Bologna and Palermo and the Polytechnic of Torino performed the post-test analysis of the LOBI experiment under the supervision of DCMN. In the report the main outcomes from the analysis of the LOBI experiments are given with the attempt to identify deficiencies in the modelling capabilities of the used codes

  6. Assessment of critical flow models of RELAP5-MOD2 and CATHARE codes

    International Nuclear Information System (INIS)

    Hao Laomi; Zhu Zhanchuan

    1992-01-01

    The critical flow tests for the long and short nozzles conducted on the SUPER MOBY-DICK facility were analyzed using the RELAP5-MOD2 and CATHARE 1.3 codes to assess the critical flow models of two codes. The critical mass flux calculated for two nozzles are given. The CATHARE code has used the thermodynamic nonequilibrium sound velocity of the two-phase fluid as the critical flow criterion, and has the better interphase transfer models and calculates the critical flow velocities with the completely implicit solution. Therefore, it can well calculate the critical flowrate and can describe the effect of the geometry L/D on the critical flowrate

  7. Vectorization and improvement of nuclear codes (MEUDAS4, FORCE, STREAM V2.6, HEATING7-VP, SCDAP/RELAP5/MOD2.5, NBI3DGFN)

    International Nuclear Information System (INIS)

    Nemoto, Toshiyuki; Suzuki, Koichiro; Isobe, Nobuo; Machida, Masahiko; Osanai, Seiji; Yokokawa, Mitsuo

    1992-09-01

    Eight nuclear codes have been vectorized and modified to improve their performance. These codes are magnetic fluid equilibrium code MEUDAS4 (CR and FFT versions), the magnetic field analysis code FORCE, the three-dimensional heat fluid analysis code STREAM V2.6, the three-dimensional heat analysis code HEATING 7-VP, the severe accident transient analysis code SCDAP/RELAP 5/MOD 2.5 for light water reactors, the ion beam orbital analysis code NBI3DGFN, and a free electron laser analysis code. The speedup ratios of the vectorized versions to the original ones in scalar mode are 2.3-4.9, 1.9-5.4, 2.6-6.2, and 1.9 for the MEUDAS4, STREAM, FORCE, and free electron laser analysis code, respectively. The definition method of the computational regions in the HEATING7-VP is improved. The SCDAP/RELAP5/MOD2.5 is modified to use extended memory regions of the computer. In this report, outlines of the codes, techniques used in the vectorization and reorganization of the codes, verification of computed results, and improvement on the performance are presented. (author)

  8. RELOS.MOD2: a code system for the determination of instationary fission product releases from molten pools

    International Nuclear Information System (INIS)

    Kortz, Ch.; Koch, M.K.; Unger, H.; Funke, F.

    1999-01-01

    For the assessment of molten corium pool source terms, a mechanistic model has been developed to describe the transport of fission products from liquid corium pool surfaces into a colder gas atmosphere. Modelling is based on an approach for diffusive and convective transport processes coupled with thermochemical equilibrium considerations enabling detailed speciation analyses of the fission products released. Both have been implemented into the code system RELOS.MOD2. RELOS.MOD2 sensitivity calculations on possible effects of anticipated uncertainties in the thermo-chemical data on the fission product release predictions are presented. (author)

  9. RELAP5/MOD2 code assessment using a LOFT L2-3 loss of coolant experiment

    International Nuclear Information System (INIS)

    Bang, Young Seok; Chung, Bub Dong; Kim, Hho Jung

    1990-01-01

    The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of the PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core in a reasonable range and that the code had deficiencies in the critical flow model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. A Sensitivity calculation with an updated version from RELAP5/MOD2 Cycle 36.04 improved the prediction of the rewet phenomena

  10. RELAP5/MOD2 development

    International Nuclear Information System (INIS)

    Miller, C.S.

    1986-01-01

    Status of the RELAP5/MOD2 computer code is discussed. While the code is undergoing international assessment, emphasis is on user support and code maintenance with modifications allowed only for error correction and user convenience improvements. User support discussed is the response to user inquiries, maintenance of manuals and the implementation of a PC based newletter service. The major 1986 user convenience improvement is the self-initialization option. The method is discussed and examples for PWR ''U tube'' and ''once through'' plants are illustrated. Future plans are also outlined

  11. RELAP5/MOD2 development

    International Nuclear Information System (INIS)

    Miller, C.S.

    1987-01-01

    Status of the RELAP5/MOD2 computer code is discussed. While the code is undergoing international assessment, emphasis is on user support and code maintenance with modifications allowed only for error correction and user convenience improvements. User support discussed is the response to user inquiries, maintenance of manuals and the implementation of a PC based newsletter service. The major 1986 user convenience improvement is the self-initialization option. The method is discussed and examples for PWR U tube and once through plants are illustrated. Future plans are also outlined

  12. International Code Assessment and Applications Program: Summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B

    International Nuclear Information System (INIS)

    Schultz, R.R.

    1993-12-01

    Members of the International Code Assessment Program (ICAP) have assessed the US Nuclear Regulatory Commission (USNRC) advanced thermal-hydraulic codes over the past few years in a concerted effort to identify deficiencies, to define user guidelines, and to determine the state of each code. The results of sixty-two code assessment reviews, conducted at INEL, are summarized. Code deficiencies are discussed and user recommended nodalizations investigated during the course of conducting the assessment studies and reviews are listed. All the work that is summarized was done using the RELAP5/MOD2, RELAP5/MOD3, and TRAC-B codes

  13. Assessment of RELAP5/MOD2 code using loss of offsite power transient data of KNU [Korea Nuclear Unit] No. 1 Plant

    International Nuclear Information System (INIS)

    Chung, Bud-Dong; Kim, Hho-Jung

    1990-04-01

    This report presents a code assessment study based on a real plant transient that occurred on June 9, 1981 at the KNU number-sign 1 (Korea Nuclear Unit Number 1). KNU number-sign 1 is a two-loop Westinghouse PWR plant of 587 Mwe. The loss of offsite power transient occurred at the 77.5% reactor power with 0.5%/hr power ramp. The real plant data were collected from available on-line plant records and computer diagnostics. The transient was simulated by RELAP5/MOD2/36.05 and the results were compared with the plant data to assess the code weaknesses and strengths. Some nodalization studies were performed to contribute to developing a guideline for PWR nodalization for the transient analysis. 5 refs., 18 figs., 3 tabs

  14. Implementation of an implicit method into heat conduction calculation of TRAC-PF1/MOD2 code

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Abe, Yutaka; Ohnuki, Akira; Murao, Yoshio

    1990-08-01

    A two-dimensional unsteady heat conduction equation is solved in the TRAC-PF/MOD2 code to calculate temperature transients in fuel rod. A large CPU time is often required to get stable solution of temperature transients in the TRAC calculation with a small axial node size (less than 1.0 mm), because the heat conduction equation is discretized explicitly. To eliminate the restriction of the maximum time step size by the heat conduction calculation, an implicit method for solving the heat condition equation was developed and implemented into the TRAC code. Several assessment calculations were performed with the original and modified TRAC codes. It is confirmed that the implicit method is reliable and is successfully implemented into the TRAC code through comparison with theoretical solutions and assessment calculation results. It is demonstrated that the implicit method makes the heat conduction calculation practical even for the analyses of temperature transients with the axial node size less than 0.1 mm. (author)

  15. BEACON/MOD2A: a computer program for subcompartment analysis of nuclear reactor containment. A user's manual

    International Nuclear Information System (INIS)

    Wells, R.A.

    1979-03-01

    The BEACON code is a Best Estimate Advanced Containment code which being developed by EG and G, Idaho, Inc., at the Idaho National Engineering Laboratory. The program is designed to perform a best estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped-parameter representations for the various parts of the system. The current version of BEACON, which is designated BEACON/MOD2A, contains mass and heat transfer models for wall film and for wall conduction. It is suitable for the evaluation of short term transients in PWR dry containment systems. This manual describes the models employed in BEACON/MOD2A and specifies code implementation requirements. It provides application information for input data preparation and for output data interpretation

  16. Quality control of the packet of RELAP5/MOD2 code

    International Nuclear Information System (INIS)

    Pomier Baez, L.E.

    1993-01-01

    The methodology that should be used to perform the quality control of entrance data set of RELAP5 calculation code is expounded in this work with this control method an extreme reliability is quarantined in the calculation model established to perform the safety thermohydraulic analysis with the help of RELAP5. This makes possible the complex simulation studies of a nuclear power plant with the quality required

  17. RELAP5/MOD2 assessment at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Turk, C.

    1986-01-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G Idaho, Inc. and the NRC assessing the RELAP5/MOD2 computer code by simulating selected separate effects tests. The purpose of this B and W Owners Group-sponsored assessment was to evaluate RELAP5/MOD2 for use in design calculations for the MIST and OTIS integral system tests and in predicting pressurized water reactor (PWR) transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (Cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve specific predictive capabilities of RELAP5/MOD2

  18. RELAP5/MOD2 models and correlations

    International Nuclear Information System (INIS)

    Dimenna, R.A.; Larson, J.R.; Johnson, R.W.; Larson, T.K.; Miller, C.S.; Streit, J.E.; Hanson, R.G.; Kiser, D.M.

    1988-08-01

    A review of the RELAP5/MOD2 computer code has been performed to assess the basis for the models and correlations comprising the code. The review has included verification of the original data base, including thermodynamic, thermal-hydraulic, and geothermal conditions; simplifying assumptions in implementation or application; and accuracy of implementation compared to documented descriptions of each of the models. An effort has been made to provide the reader with an understanding of what is in the code and why it is there and to provide enough information that an analyst can assess the impact of the correlation or model on the ability of the code to represent the physics of a reactor transient. Where assessment of the implemented versions of the models or correlations has been accomplished and published, the assessment results have been included

  19. Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code

    International Nuclear Information System (INIS)

    Martin, L.; Saenz Tejada, P.

    1993-01-01

    When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

  20. Analysis of Rod Withdrawal at Power (RWAP) Accident using ATHLET Mod 2.2 Cycle A and RELAP5/mod 3.3 Codes

    International Nuclear Information System (INIS)

    Bencik, V.; Cavlina, N.; Grgic, D.

    2012-01-01

    The system code ATHLET is being developed at Gesellschaft fuer Anlagen-und Reaktorsicherheit (GRS) in Germany. In 1996, the NPP Krsko (NEK) input deck for ATHLET Mod 1.1 Cycle C has been developed at Faculty of Electrical Engineering (FER), University of Zagreb. The input deck was tested by analyzing the realistic plant event 'Main Steam Isolation Valve Closure' and the results were assessed against the measured data. The input deck was established before plant modernization that took place in 2000 and included the power uprate and SG replacement. The released ATHLET version (Mod 2.2 Cycle A) is now being available at FER Zagreb. Accordingly, the NEK input deck for ATHLET Mod 2.2 Cycle A has been developed. A completely new input deck has been created taking into account the large number of changes due to power uprate and SG replacement as well as taking advantage of developmental work on NEK data base performed at FER. The new NEK input deck for ATHLET code has been tested by analyzing the Rod Withdrawal Power (RWAP) accident and the results were assessed against the analysis performed by RELAP5/mod 3.3 code. The RWAP accident can be either Departure from Nucleate Boiling (DNB) ratio or overpower limiting accident depending on initial power and reactivity insertion rate. Since the automatic rod control system is assumed unavailable, the only negative reactivity is due to Doppler and moderator feedback. Consequently, the nuclear power and the transferred heat in the steam generators (SGs) increase. Since the steam flow to the turbine and the extracted power from the SGs remain constant, the SG secondary pressure and the temperatures on the primary side increase. Unless terminated by manual or automatic action, the power mismatch between primary and secondary side and the resultant coolant temperature rise could eventually result in DNB ratio and/or fuel centreline melt. In order to avoid core damage, the reactor protection system is designed to automatically

  1. RELAP5/MOD2 Overview and Developmental. Assessment Results from TMl-1 Plant Transient Analysis

    International Nuclear Information System (INIS)

    Lin, J. C.; Tsai, C. C.; Ransom, V. H.; Johnsen, G. W.

    2013-01-01

    RELAP5/MOD2 is a new version of the RELAP5 thermal-hydraulic computer code containing improved modeling features that provide a generic capability for pressurized water reactor transient simulation. The objective of this paper is to provide code users with an overview of the code and to report developmental assessment results obtained from a Three Mile Island Unit One plant transient analysis. The assessment shows that the injection of highly sub-cooled water into a high-pressure primary coolant system does not cause unphysical results or pose a problem for RELAP5/MOD2. (author)

  2. Assessment of RELAP5/MOD2 against a natural circulation experiment in Nuclear Power Plant Borssele

    International Nuclear Information System (INIS)

    Winters, L.

    1993-07-01

    As part of the ICAP (International Code Assessment and Applications Program) agreement between ECN (Netherlands Energy Research Foundation) and USNRC, ECN has performed a number of assessment calculations for the thermohydraulic system analysis code RELAP5/MOD2/36.05. This document describes the assessment of this computer program versus a natural circulation experiment as conducted at the Borssele Nuclear Power Plant. The results of this comparison show that the code RELAP5/MOD2 predicts well the natural circulation behaviour of Nuclear Power Plant Borssele

  3. Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code; Implicacion de las capacidades de union fenosa dentro del area de termohidraulica en el APS de la C.N. Jose Cabrera. Aplicaciones del codigo RELAP5/MOD2

    Energy Technology Data Exchange (ETDEWEB)

    Martin, L; Saenz Tejada, P [Empresarios Agrupados, A.I.E., Madrid (Spain)

    1993-12-15

    When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

  4. Vector models in RETRAN-02 MOD 2

    International Nuclear Information System (INIS)

    Kinnersly, S.R.

    1985-06-01

    The vector momentum model in RETRAN-02 allows momentum flux to be modelled in two dimensions. Vector models in RETRAN-2 are described, including both the actual implementation in the code and the specification given in the code manual. The vector momentum model is described in detail. Other models which use vector quantities include models for volume average flow, volume average slip velocity, volume average phase velocities and fill junction flows. Both code implementations and code manual descriptions are described and inconsistencies noted. The differences between the standard RETRA-02 Mod 2 version and the Winfrith version RETN2204 are noted. (U.K.)

  5. RELAP5/MOD2: for PWR transient analysis

    International Nuclear Information System (INIS)

    Ransom, V.H.

    1983-01-01

    RELAP5 is a light water reactor system transient simulation code for use in nuclear plant safety analysis. Development of a new version, RELAP5/MOD2, has been completed and will be released to the United States Nuclear Regulatory Commission during September of 1983. The new and improved modeling capability of RELAP5/MOD2 is described and some developmental assessment results are presented. The future plans for extension to severe accident modeling are briefly discussed

  6. Analysis of experiments performed at University of Hannover with Relap5/Mod2 and Cathare codes on fluid dynamic effects in the fuel element top nozzle area during refilling and reflooding

    International Nuclear Information System (INIS)

    Ambrosini, W.; D'Auria, F.; Di Marco, P.; Fantappie, G.; Giot, G.; Emmerechts, D.; Seynhaeve, J.M.; Zhang, J.

    1989-11-01

    The experimental data of flooding and CCFL in the fuel element top nozzle area collected at the University of Hannover have been analyzed with RELAP5/MOD2 and CATHARE V.1.3 codes. Preliminary sensitivity calculations have been performed to evaluate the influence of various parameters and code options on the results. However, an a priori rational assessment procedure has been performed for those parameters non specific in experimental data (e.g. energy loss coefficients in flow restrictions). This procedure is based on single phase flow pressure drops and no further tuning has been performed to fit experimental data. The reported experimental data and some others demonstrate the complex relation-ship among the involved physical quantities (film thickness, pressure drop etc.) even in a simple geometrical condition with well defined boundary conditions. In the application of the two advanced codes to the selected CCFL experiments it appears that sophisticated models do not simulate satisfactorily the measured phenomena mainly when situations similar to nuclear reactors are dealt with (rod bundles). This result should be evaluated considering that: - dimensional phenomena occurring in flooding experiments are not well reproducible with one dimensional models implemented in the two codes; - a rational and reproducible procedure has been used to fix some boundary conditions (K-tuning); there is the evidence that more tuning can be used to get results closer to the experimental ones in each specific situation; - the uncertainty bands in measured experimental results are not (entirely) specified. The work performed demonstrated that further applications to CCFL experiments of present codes appear to be unuseful. New models should be tested and implemented before any attempt to reproduce CCFL in experimental facilities by system codes

  7. ICAP [International Code Assessment and Applications Program] assessment of RELAP5/MOD2, Cycle 36.05 against LOFT [Loss of Fluid Test] Small Break Experiment L3-7

    International Nuclear Information System (INIS)

    Lee, Euy-Joon; Chung, Bud-Dong; Kim, Hho-Jung

    1990-04-01

    The LOFT small break (1 in-dia) experiment L3-7 has been analyzed using the reactor thermal hydraulic analysis code RELAP5/MOD2, Cycle 36.05. The base calculation (Case A) was completed and compared with the experimental data. Three types of sensitivity studies (Cases B, Cm, and D) were carried out to investigate the effects of (1) break discharge coefficient Cd, (2) pump two-phase difference multiplier and (3) High Pressure Injection System (HPIS) capacity on major thermal and hydraulic (T/H) parameters. A nodalization study (Case E) was conducted to assess the phenomena with a simplified nodalization. The results indicate that Cd of 0.9 and 0.1 fit to the single discharge flow rate of Test L3-7 best among the tried cases. The pump two-phase multiplier has little effects on the T/H parameters because of the low discharge flow rate and the early pump coast down in this smaller size SBLOCA. But HPIS capacity has a very strong influence on parameters such as pressure, flow and temperature. It is also shown that a simplified nodalization could accomodate the dominant T/H phenomena with the same degree of code accuracy and efficiency

  8. Evaluation of CNA I coolant channel behaviour during an accidental transient using ICARE2 V2 mod2.3 code

    International Nuclear Information System (INIS)

    Marino, Edgardo J.L.

    1999-01-01

    Using the input data language of ICARE2 V2 Mod.3 code, the fuel element and coolant channel assembly of CNA I type was described. This input data was utilized to analyze the system behavior and determine the degradation produced during a hypothetical accidental transient at CNA I. The boundary conditions were determined through a previous calculation with RELAP5/MOD 3.2 code. The results had shown characteristic degradation phenomena's. The temperature of bundle components increases fast after 6.11 h in the first case and 5.28 h in the second case, due to the energy release by cladding oxidation. It was correlated with instantaneous hydrogen production and energy contribution. The cumulated hydrogen production was estimated as 0.15 Kg in the first case and ∼ 5 times greater in the second case. Fission product release from the gap due to cladding rupture took place from 6.25 h in the first case and 5.65 h in the second. Relocation started after 6.81 h in the first case and 5.68 in the second, because the cladding dislocation condition is reached. UO 2 dissolution by molten Zircaloy was observed at different levels in the calculation domain. (author)

  9. TRAC-PF1/MOD2 status and plans

    International Nuclear Information System (INIS)

    Spore, J.W.; Steinke, R.G.; Nelson, R.A.; Cappiello, M.W.; Jenks, R.

    1989-01-01

    The development of the TRAC-PF1/MOD1 code was completed in July 1988 with the release of Version 14.4. A TRAC-PF1/MOD2 code development plan addresses code deficiencies identified in the MOD1 code in order to provide an accurate and defensible tool that can be used to simulate large-break loss-of-coolant accidents (LOCAs), small-break LOCAs, and operational transients. The MOD2 code development plan is an international cooperative effort that includes contributions from Los Alamos National Laboratory, Idaho National Engineering Laboratory (INEL), Japanese Atomic Energy Research Institute (JAERI), Cray Research, Central Electricity Generating Board (CEGB), and United Kingdom Atomic Energy Authority (UKAEA)

  10. CONCEPT computer code

    International Nuclear Information System (INIS)

    Delene, J.

    1984-01-01

    CONCEPT is a computer code that will provide conceptual capital investment cost estimates for nuclear and coal-fired power plants. The code can develop an estimate for construction at any point in time. Any unit size within the range of about 400 to 1300 MW electric may be selected. Any of 23 reference site locations across the United States and Canada may be selected. PWR, BWR, and coal-fired plants burning high-sulfur and low-sulfur coal can be estimated. Multiple-unit plants can be estimated. Costs due to escalation/inflation and interest during construction are calculated

  11. CONTEMPT4/MOD2: a multicompartment containment system analysis program

    International Nuclear Information System (INIS)

    Metcalfe, L.J.; Mings, W.J.; Hartman, J.E.; Crail, A.C.

    1978-02-01

    CONTEMPT4/MOD2 is a digital computer program, written in FORTRAN IV, which describes the behavior of multicompartment pressurized water reactor (PWR) containment systems and experimental containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and intercompartment mass and energy exchange based on user-supplied values for compartment descriptions, time step and edit controls, and selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, fan coolers, heat conducting structures, sump drain, and PWR ice condensers. Dynamic storage allocations (DSA) is used to limit the amount of computer core used for each problem. Optional automatic time step control allows the code to determine time step sizes within limits dictated by the user. Multicompartment capability (up to 999 individual compartments) and generalized, user-oriented input data descriptions permit improved flexibility over previous codes in the CONTEMPT series. Analytical model descriptions, input instructions, and sample problem results are presented

  12. On mod 2 and higher elliptic genera

    International Nuclear Information System (INIS)

    Liu Kefeng

    1992-01-01

    In the first part of this paper, we construct mod 2 elliptic genera on manifolds of dimensions 8k+1, 8k+2 by mod 2 index formulas of Dirac operators. They are given by mod 2 modular forms or mod 2 automorphic functions. We also obtain an integral formula for the mod 2 index of the Dirac operator. As a by-product we find topological obstructions to group actions. In the second part, we construct higher elliptic genera and prove some of their rigidity properties under group actions. In the third part we write down characteristic series for all Witten genera by Jacobi theta-functions. The modular property and transformation formulas of elliptic genera then follow easily. We shall also prove that Krichever's genera, which come from integrable systems, can be written as indices of twisted Dirac operators for SU-manifolds. Some general discussions about elliptic genera are given. (orig.)

  13. Computer code FIT

    International Nuclear Information System (INIS)

    Rohmann, D.; Koehler, T.

    1987-02-01

    This is a description of the computer code FIT, written in FORTRAN-77 for a PDP 11/34. FIT is an interactive program to decude position, width and intensity of lines of X-ray spectra (max. length of 4K channels). The lines (max. 30 lines per fit) may have Gauss- or Voigt-profile, as well as exponential tails. Spectrum and fit can be displayed on a Tektronix terminal. (orig.) [de

  14. MELCOR computer code manuals

    Energy Technology Data Exchange (ETDEWEB)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.

  15. MELCOR computer code manuals

    International Nuclear Information System (INIS)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

  16. Computer codes for safety analysis

    International Nuclear Information System (INIS)

    Holland, D.F.

    1986-11-01

    Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans

  17. TRAC-PF1/MOD2 best-estimate analysis of a large-break LOCA in a 15 x 15 generic four-loop Westinghouse nuclear power plant

    International Nuclear Information System (INIS)

    Spore, J.W.; Lin, J.C.; Schnurr, N.M.; White, J.R.; Cappiello, M.C.

    1992-01-01

    Calculations of a large-break loss-of-coolant accident (LOCA) in a 15 x 15 generic four-loop Westinghouse nuclear power plant with both the TRAC-PF1/MOD1 and TRAC-PF1/MOD2 computer codes will be presented. The Transient Reactor Analysis Code (TRAC) has been developed by Los Alamos National Laboratory to provide advanced best-estimate simulations of real postulated transients in pressurized light-water reactors (LWRs) and for many related thermal-hydraulic facilities. The latest released version of TRAC is TRAC-PF1/MOD2. Significant improvements and enhancements over the MOD1 version were implemented in the MOD2 heat-transfer and constitutive models. One of the most significant improvements in the MOD2 code has been the implementation of the two-step numerics method in the three-dimensional components, which can significantly reduce run times for long, slow transients. A very important area of improvement has been in the reflood heat-transfer models. Developmental assessment results (i.e., code comparisons with experimental data) will be discussed for several separate-effects and integral test, including analysis of the Upper Plenum Test Facility (UPTF), the Cylindrical Core Test Facility (CCTF), and the Loss-of-Fluid Test Facility (LOFT). The assessment results provide information on the anticipated accuracy for the best-estimate models in the MOD2 computer code. The MOD1 to MOD2 comparison will provide an estimate for the effect of improved heat-transfer models on predicted peak cladding temperatures

  18. Geochemical computer codes. A review

    International Nuclear Information System (INIS)

    Andersson, K.

    1987-01-01

    In this report a review of available codes is performed and some code intercomparisons are also discussed. The number of codes treating natural waters (groundwater, lake water, sea water) is large. Most geochemical computer codes treat equilibrium conditions, although some codes with kinetic capability are available. A geochemical equilibrium model consists of a computer code, solving a set of equations by some numerical method and a data base, consisting of thermodynamic data required for the calculations. There are some codes which treat coupled geochemical and transport modeling. Some of these codes solve the equilibrium and transport equations simultaneously while other solve the equations separately from each other. The coupled codes require a large computer capacity and have thus as yet limited use. Three code intercomparisons have been found in literature. It may be concluded that there are many codes available for geochemical calculations but most of them require a user that us quite familiar with the code. The user also has to know the geochemical system in order to judge the reliability of the results. A high quality data base is necessary to obtain a reliable result. The best results may be expected for the major species of natural waters. For more complicated problems, including trace elements, precipitation/dissolution, adsorption, etc., the results seem to be less reliable. (With 44 refs.) (author)

  19. Evaluation and assessment of reflooding models in RELAP5/Mod2.5 and RELAP5/Mod3 codes using Lehigh University and PSI-Neptun bundle experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Sencar, M.; Aksan, N. [Paul Scherrer Institute, Villigen (Switzerland)

    1995-09-01

    An extensive analysis and assessment work on reflooding models of RELAP5/Mod2.5 and, RELAP5/Mod3/v5m5 and RELAP/Mod3/v7j have been performed. Experimental data from LehighUniversityv. and PSI-NEPTUN bundle reflooding experiments have been used for the assessment, since both of these tests cover a broad range of initial conditions. Within the range of these initial conditions, it was tried to identify their separate impacts on the calculated results. A total of six Lehigh University reflooding bundle tests and two PSI-NEPTUN tests with bounding initial conditions are selected for the analysis. Detailed nodalisation studies both for hydraulic and conduction heat transfer were done. On the basis of the results obtained from these cases, a base nodalisation scheme was established. All the other analysis work was performed by using this base nodalisation. RELAP5/Mod2.5 results do not change with renodalisation but RELAP5/Mod3 results are more sensitive to renodalisation. The results of RELAP5/Mod2.5 versions show very large deviations from the used experimental data. These results indicate that some of the phenomenology of the events occurring during the reflooding could not be identified. In the paper, detailed discussions on the main reasons of the deviations from the experimental data will be presented. Since, the results and findings of this study are meant to be a developmental aid, some recommendations have been drawn and some of these have already been implemented at PSI with promising results.

  20. Automotive Stirling engine: Mod 2 design report

    Science.gov (United States)

    Nightingale, Noel P.

    1986-01-01

    The design of an automotive Stirling engine that achieves the superior fuel economy potential of the Stirling cycle is described. As the culmination of a 9-yr development program, this engine, designated the Mod 2, also nullifies arguments that Stirling engines are heavy, expensive, unreliable, demonstrating poor performance. Installed in a General Motors Chevrolet Celebrity car, this engine has a predicted combined fuel economy on unleaded gasoline of 17.5 km/l (41 mpg)- a value 50% above the current vehicle fleet average. The Mod 2 Stirling engine is a four-cylinder V-drive design with a single crankshaft. The engine is also equipped with all the controls and auxiliaries necessary for automotive operation.

  1. Computer code abstract: NESTLE

    International Nuclear Information System (INIS)

    Turinsky, P.J.; Al-Chalabi, R.M.K.; Engrand, P.; Sarsour, H.N.; Faure, F.X.; Guo, W.

    1995-01-01

    NESTLE is a few-group neutron diffusion equation solver utilizing the nodal expansion method (NEM) for eigenvalue, adjoint, and fixed-source steady-state and transient problems. The NESTLE code solve the eigenvalue (criticality), eigenvalue adjoint, external fixed-source steady-state, and external fixed-source or eigenvalue initiated transient problems. The eigenvalue problem allows criticality searches to be completed, and the external fixed-source steady-state problem can search to achieve a specified power level. Transient problems model delayed neutrons via precursor groups. Several core properties can be input as time dependent. Two- or four-energy groups can be utilized, with all energy groups being thermal groups (i.e., upscatter exits) is desired. Core geometries modeled include Cartesian and hexagonal. Three-, two-, and one-dimensional models can be utilized with various symmetries. The thermal conditions predicted by the thermal-hydraulic model of the core are used to correct cross sections for temperature and density effects. Cross sections for temperature and density effects. Cross sections are parameterized by color, control rod state (i.e., in or out), and burnup, allowing fuel depletion to be modeled. Either a macroscopic or microscopic model may be employed

  2. TRAC code development status and plans

    International Nuclear Information System (INIS)

    Spore, J.W.; Liles, D.R.; Nelson, R.A.

    1986-01-01

    This report summarizes the characteristics and current status of the TRAC-PF1/MOD1 computer code. Recent error corrections and user-convenience features are described, and several user enhancements are identified. Current plans for the release of the TRAC-PF1/MOD2 computer code and some preliminary MOD2 results are presented. This new version of the TRAC code implements stability-enhancing two-step numerics into the 3-D vessel, using partial vectorization to obtain a code that has run 400% faster than the MOD1 code

  3. RELAP5/MOD2 calculation of OECD LOFT test LP-FW-01

    International Nuclear Information System (INIS)

    Croxfod, M.G.; Harwood, C.; Hall, P.C.

    1992-04-01

    RELAP5/MOD2 is being used by GDCD for calculation of certain small break loss-of-coolant accidents and pressurized transients in the Sizewell ''B'' PWR. To test the ability of RELAP5/MOD2 to model the primary feed-and-bleed recovery procedure following a complete loss- of-feedwater event, post test calculations have been carried out of OECD LOFT test LP-FW-01. This report describes the comparison between the code calculations and the test data. It is found that although the standard version of RELAP5/MOD2 gives a reasonable prediction of the experimental transient, the long term pressure history is better calculated with a modified code version containing a revised horizontal stratification entrainment model. The latter allows an improved calculation of entrainment of liquid from the hot leg into the surge line. RELAP5/MOD2 is found to give a more accurate simulation of the experimental transient than was achieved in previous UK studies using RETRAN-02/MOD2

  4. 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)

  5. The RETRAN-03 computer code

    International Nuclear Information System (INIS)

    Paulsen, M.P.; McFadden, J.H.; Peterson, C.E.; McClure, J.A.; Gose, G.C.; Jensen, P.J.

    1991-01-01

    The RETRAN-03 code development effort is designed to overcome the major theoretical and practical limitations associated with the RETRAN-02 computer code. The major objectives of the development program are to extend the range of analyses that can be performed with RETRAN, to make the code more dependable and faster running, and to have a more transportable code. The first two objectives are accomplished by developing new models and adding other models to the RETRAN-02 base code. The major model additions for RETRAN-03 are as follows: implicit solution methods for the steady-state and transient forms of the field equations; additional options for the velocity difference equation; a new steady-state initialization option for computer low-power steam generator initial conditions; models for nonequilibrium thermodynamic conditions; and several special-purpose models. The source code and the environmental library for RETRAN-03 are written in standard FORTRAN 77, which allows the last objective to be fulfilled. Some models in RETRAN-02 have been deleted in RETRAN-03. In this paper the changes between RETRAN-02 and RETRAN-03 are reviewed

  6. Computer access security code system

    Science.gov (United States)

    Collins, Earl R., Jr. (Inventor)

    1990-01-01

    A security code system for controlling access to computer and computer-controlled entry situations comprises a plurality of subsets of alpha-numeric characters disposed in random order in matrices of at least two dimensions forming theoretical rectangles, cubes, etc., such that when access is desired, at least one pair of previously unused character subsets not found in the same row or column of the matrix is chosen at random and transmitted by the computer. The proper response to gain access is transmittal of subsets which complete the rectangle, and/or a parallelepiped whose opposite corners were defined by first groups of code. Once used, subsets are not used again to absolutely defeat unauthorized access by eavesdropping, and the like.

  7. Microgravity computing codes. User's guide

    Science.gov (United States)

    1982-01-01

    Codes used in microgravity experiments to compute fluid parameters and to obtain data graphically are introduced. The computer programs are stored on two diskettes, compatible with the floppy disk drives of the Apple 2. Two versions of both disks are available (DOS-2 and DOS-3). The codes are written in BASIC and are structured as interactive programs. Interaction takes place through the keyboard of any Apple 2-48K standard system with single floppy disk drive. The programs are protected against wrong commands given by the operator. The programs are described step by step in the same order as the instructions displayed on the monitor. Most of these instructions are shown, with samples of computation and of graphics.

  8. H0 precessor computer code

    International Nuclear Information System (INIS)

    van Dyck, O.B.; Floyd, R.A.

    1981-05-01

    A spin precessor using H - to H 0 stripping, followed by small precession magnets, has been developed for the LAMPF 800-MeV polarized H - beam. The performance of the system was studied with the computer code documented in this report. The report starts from the fundamental physics of a system of spins with hyperfine coupling in a magnetic field and contains many examples of beam behavior as calculated by the program

  9. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    Energy Technology Data Exchange (ETDEWEB)

    Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  10. Computer code conversion using HISTORIAN

    International Nuclear Information System (INIS)

    Matsumoto, Kiyoshi; Kumakura, Toshimasa.

    1990-09-01

    When a computer program written for a computer A is converted for a computer B, in general, the A version source program is rewritten for B version. However, in this way of program conversion, the following inconvenient problems arise. 1) The original statements to be rewritten for B version are lost. 2) If the original statements of the A version rewritten for B version would remain as comment lines, the B version source program becomes quite large. 3) When update directives of the program are mailed from the organization which developed the program or when some modifications are needed for the program, it is difficult to point out the part to be updated or modified in the B version source program. To solve these problems, the conversion method using the general-purpose software management aid system, HISTORIAN, has been introduced. This conversion method makes a large computer code a easy-to-use program for use to update, modify or improve after the conversion. This report describes the planning and procedures of the conversion method and the MELPROG-PWR/MOD1 code conversion from the CRAY version to the JAERI FACOM version as an example. This report would provide useful information for those who develop or introduce large programs. (author)

  11. Computation of the Genetic Code

    Science.gov (United States)

    Kozlov, Nicolay N.; Kozlova, Olga N.

    2018-03-01

    One of the problems in the development of mathematical theory of the genetic code (summary is presented in [1], the detailed -to [2]) is the problem of the calculation of the genetic code. Similar problems in the world is unknown and could be delivered only in the 21st century. One approach to solving this problem is devoted to this work. For the first time provides a detailed description of the method of calculation of the genetic code, the idea of which was first published earlier [3]), and the choice of one of the most important sets for the calculation was based on an article [4]. Such a set of amino acid corresponds to a complete set of representations of the plurality of overlapping triple gene belonging to the same DNA strand. A separate issue was the initial point, triggering an iterative search process all codes submitted by the initial data. Mathematical analysis has shown that the said set contains some ambiguities, which have been founded because of our proposed compressed representation of the set. As a result, the developed method of calculation was limited to the two main stages of research, where the first stage only the of the area were used in the calculations. The proposed approach will significantly reduce the amount of computations at each step in this complex discrete structure.

  12. RELAP5/MOD2 implementation on various mainframes including the IBM and SX-2 supercomputer

    International Nuclear Information System (INIS)

    DeForest, D.L.; Hassan, Y.A.

    1987-01-01

    The RELAP5/MOD2 (cycle 36.04) code is a one-dimensional, two-fluid, nonequilibrium, nonhomogeneous transient analysis code designed to simulate operational and accident scenarios in pressurized water reactors (PWRs). System models are solved using a semi-implicit finite difference method. The code was developed at EG and G in Idaho Falls under sponsorship of the US Nuclear Regulatory Commission (NRC). The major enhancement from RELAP5/MOD1 is the use of a six-equation, two-fluid nonequilibrium and nonhomogeneous model. Other improvements include the addition of a noncondensible gas component and the revision and addition of drag formulation, wall friction, and wall heat transfer. Several test cases were run to benchmark the IBM and SX-2 installations against the CDC computer and the CRAY-2 and CRAY/XMP. These included the Edward's pipe blow-down and two separate reflood cases developed to simulate the FLECHT-SEASET reflood test 31504 and a postcritical heat flux (CHF) test performed at Lehigh University

  13. Current algorithms used in reactor safety codes and the impact of future computer development on these algorithms

    International Nuclear Information System (INIS)

    Mahaffy, J.H.; Liles, D.R.; Woodruff, S.B.

    1985-01-01

    Computational methods and solution procedures used in the US Nuclear Regulatory Commission's reactor safety systems codes, Transient Reactor Analysis Code (TRAC) and Reactor Leak and Power Safety Excursion Code (RELAP), are reviewed. Methods used in TRAC-PF1/MOD1, including the stability-enhancing two-step (SETS) technique, which permits fast computations by allowing time steps larger than the material Courant stability limit, are described in detail, and the differences from RELAP5/MOD2 are noted. Developments in computing, including parallel and vector processing, and their applicability to nuclear reactor safety codes are described. These developments, coupled with appropriate numerical methods, make detailed faster-than-real-time reactor safety analysis a realistic near-term possibility

  14. Relap5/mod2 post-test calculation of a loss of feedwater experiment at the Pactel test facility

    Energy Technology Data Exchange (ETDEWEB)

    Protze, M. [Siemens-KWU, Erlangen (Germany)

    1995-12-31

    Post-test calculations for verification purposes of the thermal hydraulic code RELAP5/MOD2 are of fundamental importance for the licensing procedure. The RELAP5/MOD2 code has a large international assessment base regarding western PWR. WWER-reactors are russian designed PWRs with some specific differences compared with the western PWR`s, especially the horizontal steam generators. For that reason some post-test calculations have to be performed to verify the RELAP5/MOD2 code for these WWER typical phenomena. The impact of the horizontal steam generators on the accident behaviour during transients or pipe ruptures on the secondary side is significant. The nodalization of the test facility PACTEL was chosen equally to WWER plant nodalization to verify the use of a coarse modelling of the steam generator secondary side for analyses of transient with decreasing water level in the SG secondary side. The calculational results showed a good compliance to the test results, demonstrating the correct use of a coarse nodalization. To sum up, the RELAP5/ MOD2 results met the test results appropriately thereby the RELAP5/ MOD2 code is validated for analyses of transients with decreasing water level in a horizontal steam generator secondary side. (orig.). 4 refs.

  15. Relap5/mod2 post-test calculation of a loss of feedwater experiment at the Pactel test facility

    Energy Technology Data Exchange (ETDEWEB)

    Protze, M [Siemens-KWU, Erlangen (Germany)

    1996-12-31

    Post-test calculations for verification purposes of the thermal hydraulic code RELAP5/MOD2 are of fundamental importance for the licensing procedure. The RELAP5/MOD2 code has a large international assessment base regarding western PWR. WWER-reactors are russian designed PWRs with some specific differences compared with the western PWR`s, especially the horizontal steam generators. For that reason some post-test calculations have to be performed to verify the RELAP5/MOD2 code for these WWER typical phenomena. The impact of the horizontal steam generators on the accident behaviour during transients or pipe ruptures on the secondary side is significant. The nodalization of the test facility PACTEL was chosen equally to WWER plant nodalization to verify the use of a coarse modelling of the steam generator secondary side for analyses of transient with decreasing water level in the SG secondary side. The calculational results showed a good compliance to the test results, demonstrating the correct use of a coarse nodalization. To sum up, the RELAP5/ MOD2 results met the test results appropriately thereby the RELAP5/ MOD2 code is validated for analyses of transients with decreasing water level in a horizontal steam generator secondary side. (orig.). 4 refs.

  16. Assessment of RELAP/MOD2 using large break loss-of-coolant experimental data

    International Nuclear Information System (INIS)

    Kao, L.; Liao, L.Y.; Liang, K.S.; Wang, S.F.; Chen, Y.B.

    1989-01-01

    In this paper assessment of RELAP5/MOD2 using LOFT L2-5 and Semiscale S-06-3 tests are performed to provide information of the code capability and its limitation in analyzing large break LOCA of a nuclear power plant. Experiments L2-5 and S-06-3 are conducted to simulate a hypothetical LOCA which results from a 200% double-ended offset shear break in the cold-leg of a typical pressurized water reactor by utilizing scaling facilities of the LOFT and Semiscale Mod-1 systems, respectively. The RELAP5/MOD2 calculations for both tests begin with break initiation and subsequent blowdown, continue through lower plenum refill, core reflood, and terminate with corewide quench. Major phenomena of both large break loss-of-coolant tests are well predicted by RELAP5/MOD2. The results indicate that the break flow and system pressure are reasonably calculated. The cladding temperature response during blowdown period, which is the major importance to a large break LOCA, calculated by RELAP5/MOD2 shows good agreement with the test data

  17. Transient simulation of ALWR passive safety systems using RELAP5/MOD2

    International Nuclear Information System (INIS)

    Elias, E.; Nekhamkin, Y.; Arshavski, I.

    2004-01-01

    Numerical simulation is presented of some passive safety systems currently incorporated in the design of the next generation advanced light water reactors (ALWRs). The performance and effectiveness of ex-core natural convection cooling and the concept of gravity driven water injection at high pressure are investigated using the RELAP5/MOD2 thermal-hydraulic code. The study identifies areas that should be investigated more fully in future experimental programs related to hypothetical large and small LOCA in ALWRs. (author)

  18. Computer codes for ventilation in nuclear facilities

    International Nuclear Information System (INIS)

    Mulcey, P.

    1987-01-01

    In this paper the authors present some computer codes, developed in the last years, for ventilation and radioprotection. These codes are used for safety analysis in the conception, exploitation and dismantlement of nuclear facilities. The authors present particularly: DACC1 code used for aerosol deposit in sampling circuit of radiation monitors; PIAF code used for modelization of complex ventilation system; CLIMAT 6 code used for optimization of air conditioning system [fr

  19. Evaluation of fuel-temperature feedback mechanisms in TRAC-PF1/MOD2/NESTLE

    International Nuclear Information System (INIS)

    Knepper, Paula L.; Feltus, Madeline; Hochreiter, L.E.; Ivanov, Kostadin

    1999-01-01

    Coupled spatial kinetics and thermal-hydraulics system codes provide a means to model transient nuclear reactor behavior more accurately. Transients marked by strong perturbations, both with thermal-hydraulics and neutronics, such as a control-rod ejection or a main steam-line break, are especially of interest. It is now feasible to model complex reactor behavior with a coupled thermal-hydraulics and spatial kinetics code that provides a means to forecast safety margins. Recently, the Transient Reactor Analysis Code (TRAC)-PF1/MOD2, Version 5.4.25, was coupled with the NESTLE code. This coupled code (TRAC-PF1/MOD2/NESTLE) is used to examine effective fuel-temperature models. The Electric Power Research Institute (EPRI) rod-ejection benchmark was analyzed to evaluate the influence of effective fuel temperature. The rod-ejection transient tests only the fuel-rod, heat-conduction coupling. The coolant thermal-hydraulic coupling is not tested because of the speed of the transient. The neutronics solution changes extremely rapidly, whereas the convective heat transfer at the fuel surface requires more time to influence the coolant temperature of the system. The need to model the response of the system coolant temperature is not crucial in this analysis. The influence of the effective fuel temperature is the key component of this study. Various models were examined using the coupled code to calculate effective fuel temperatures. The influence of different, effective fuel-temperature models on the coupled-code results is studied. Three effective fuel-temperature models are examined: (l) volume average effective fuel temperature, (2) the effective fuel-temperature model suggested by the Office of Economic Cooperation and Development (OECD) rod-ejection benchmark, and (3) the NESTLE effective fuel-temperature model. A discussion is provided describing the effective fuel-temperature models examined in TRAC-PF1/MOD2/NESTLE and the influence of effective fuel temperature in

  20. Reactor safety computer code development at INEL

    International Nuclear Information System (INIS)

    Johnsen, G.W.

    1985-01-01

    This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise

  1. Computer codes for RF cavity design

    International Nuclear Information System (INIS)

    Ko, K.

    1992-08-01

    In RF cavity design, numerical modeling is assuming an increasingly important role with the help of sophisticated computer codes and powerful yet affordable computers. A description of the cavity codes in use in the accelerator community has been given previously. The present paper will address the latest developments and discuss their applications to cavity toning and matching problems

  2. Implatation of MC2 computer code

    International Nuclear Information System (INIS)

    Seehusen, J.; Nair, R.P.K.; Becceneri, J.C.

    1981-01-01

    The implantation of MC2 computer code in the CDC system is presented. The MC2 computer code calculates multigroup cross sections for tipical compositions of fast reactors. The multigroup constants are calculated using solutions of PI or BI approximations for determined buckling value as weighting function. (M.C.K.) [pt

  3. Computer codes for RF cavity design

    International Nuclear Information System (INIS)

    Ko, K.

    1992-01-01

    In RF cavity design, numerical modeling is assuming an increasingly important role with the help of sophisticated computer codes and powerful yet affordable computers. A description of the cavity codes in use in the accelerator community has been given previously. The present paper will address the latest developments and discuss their applications to cavity tuning and matching problems. (Author) 8 refs., 10 figs

  4. Computation of the bounce-average code

    International Nuclear Information System (INIS)

    Cutler, T.A.; Pearlstein, L.D.; Rensink, M.E.

    1977-01-01

    The bounce-average computer code simulates the two-dimensional velocity transport of ions in a mirror machine. The code evaluates and bounce-averages the collision operator and sources along the field line. A self-consistent equilibrium magnetic field is also computed using the long-thin approximation. Optionally included are terms that maintain μ, J invariance as the magnetic field changes in time. The assumptions and analysis that form the foundation of the bounce-average code are described. When references can be cited, the required results are merely stated and explained briefly. A listing of the code is appended

  5. MELMRK 2.0: A description of computer models and results of code testing

    International Nuclear Information System (INIS)

    Wittman, R.S.; Denny, V.; Mertol, A.

    1992-01-01

    An advanced version of the MELMRK computer code has been developed that provides detailed models for conservation of mass, momentum, and thermal energy within relocating streams of molten metallics during meltdown of Savannah River Site (SRS) reactor assemblies. In addition to a mechanistic treatment of transport phenomena within a relocating stream, MELMRK 2.0 retains the MOD1 capability for real-time coupling of the in-depth thermal response of participating assembly heat structure and, further, augments this capability with models for self-heating of relocating melt owing to steam oxidation of metallics and fission product decay power. As was the case for MELMRK 1.0, the MOD2 version offers state-of-the-art numerics for solving coupled sets of nonlinear differential equations. Principal features include application of multi-dimensional Newton-Raphson techniques to accelerate convergence behavior and direct matrix inversion to advance primitive variables from one iterate to the next. Additionally, MELMRK 2.0 provides logical event flags for managing the broad range of code options available for treating such features as (1) coexisting flow regimes, (2) dynamic transitions between flow regimes, and (3) linkages between heatup and relocation code modules. The purpose of this report is to provide a detailed description of the MELMRK 2.0 computer models for melt relocation. Also included are illustrative results for code testing, as well as an integrated calculation for meltdown of a Mark 31a assembly

  6. Prediction of thermal-Hydraulic phenomena in the LBLOCA experiment L2-3 using RELAP5/MOD2

    International Nuclear Information System (INIS)

    Bang, Young Seok; Chung, Bub Dong; Kim, Hho Jung

    1991-01-01

    The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical flow model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5/MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena. (Author)

  7. Modification in the FUDA computer code to predict fuel performance at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Das, M; Arunakumar, B V; Prasad, P N [Nuclear Power Corp., Mumbai (India)

    1997-08-01

    The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig.

  8. Modification in the FUDA computer code to predict fuel performance at high burnup

    International Nuclear Information System (INIS)

    Das, M.; Arunakumar, B.V.; Prasad, P.N.

    1997-01-01

    The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig

  9. Computer Code for Nanostructure Simulation

    Science.gov (United States)

    Filikhin, Igor; Vlahovic, Branislav

    2009-01-01

    Due to their small size, nanostructures can have stress and thermal gradients that are larger than any macroscopic analogue. These gradients can lead to specific regions that are susceptible to failure via processes such as plastic deformation by dislocation emission, chemical debonding, and interfacial alloying. A program has been developed that rigorously simulates and predicts optoelectronic properties of nanostructures of virtually any geometrical complexity and material composition. It can be used in simulations of energy level structure, wave functions, density of states of spatially configured phonon-coupled electrons, excitons in quantum dots, quantum rings, quantum ring complexes, and more. The code can be used to calculate stress distributions and thermal transport properties for a variety of nanostructures and interfaces, transport and scattering at nanoscale interfaces and surfaces under various stress states, and alloy compositional gradients. The code allows users to perform modeling of charge transport processes through quantum-dot (QD) arrays as functions of inter-dot distance, array order versus disorder, QD orientation, shape, size, and chemical composition for applications in photovoltaics and physical properties of QD-based biochemical sensors. The code can be used to study the hot exciton formation/relation dynamics in arrays of QDs of different shapes and sizes at different temperatures. It also can be used to understand the relation among the deposition parameters and inherent stresses, strain deformation, heat flow, and failure of nanostructures.

  10. The computer code SEURBNUK-2

    International Nuclear Information System (INIS)

    Yerkess, A.

    1984-01-01

    SEURBNUK-2 has been designed to model the hydrodynamic development in time of a hypothetical core disrupture accident in a fast breeder reactor. SEURBNUK-2 is a two-dimensional, axisymmetric, eulerian, finite difference containment code. The numerical procedure adopted in SEURBNUK to solve the hydrodynamic equations is based on the semi-implicit ICE method. SEURBNUK has a full thin shell treatment for tanks of arbitrary shape and includes the effects of the compressibility of the fluid. Fluid flow through porous media and porous structures can also be accommodated. An important feature of SEURBNUK is that the thin shell equations are solved quite separately from those of the fluid, and the time step for the fluid flow calculation can be an integer multiple of that for calculating the shell motion. The interaction of the shell with the fluid is then considered as a modification to the coefficients in the implicit pressure equations, the modifications naturally depending on the behaviour of the thin shell section within the fluid cell. The code is limited to dealing with a single fluid, the coolant, whereas the bubble and the cover gas are treated as cavities of uniform pressure calculated via appropriate pressure-volume-energy relationships. This manual describes the input data specifications needed for the execution of SEURBNUK-2 calculations and nine sample problems of varying degrees of complexity highlight the code capabilities. After explaining the output facilities information is included to aid those unfamiliar with SEURBNUK-2 to avoid the common pit-falls experienced by novices

  11. Quantum computation with Turaev-Viro codes

    International Nuclear Information System (INIS)

    Koenig, Robert; Kuperberg, Greg; Reichardt, Ben W.

    2010-01-01

    For a 3-manifold with triangulated boundary, the Turaev-Viro topological invariant can be interpreted as a quantum error-correcting code. The code has local stabilizers, identified by Levin and Wen, on a qudit lattice. Kitaev's toric code arises as a special case. The toric code corresponds to an abelian anyon model, and therefore requires out-of-code operations to obtain universal quantum computation. In contrast, for many categories, such as the Fibonacci category, the Turaev-Viro code realizes a non-abelian anyon model. A universal set of fault-tolerant operations can be implemented by deforming the code with local gates, in order to implement anyon braiding. We identify the anyons in the code space, and present schemes for initialization, computation and measurement. This provides a family of constructions for fault-tolerant quantum computation that are closely related to topological quantum computation, but for which the fault tolerance is implemented in software rather than coming from a physical medium.

  12. Cloud Computing for Complex Performance Codes.

    Energy Technology Data Exchange (ETDEWEB)

    Appel, Gordon John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klein, Brandon Thorin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Miner, John Gifford [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-02-01

    This report describes the use of cloud computing services for running complex public domain performance assessment problems. The work consisted of two phases: Phase 1 was to demonstrate complex codes, on several differently configured servers, could run and compute trivial small scale problems in a commercial cloud infrastructure. Phase 2 focused on proving non-trivial large scale problems could be computed in the commercial cloud environment. The cloud computing effort was successfully applied using codes of interest to the geohydrology and nuclear waste disposal modeling community.

  13. Computer codes for designing proton linear accelerators

    International Nuclear Information System (INIS)

    Kato, Takao

    1992-01-01

    Computer codes for designing proton linear accelerators are discussed from the viewpoint of not only designing but also construction and operation of the linac. The codes are divided into three categories according to their purposes: 1) design code, 2) generation and simulation code, and 3) electric and magnetic fields calculation code. The role of each category is discussed on the basis of experience at KEK (the design of the 40-MeV proton linac and its construction and operation, and the design of the 1-GeV proton linac). We introduce our recent work relevant to three-dimensional calculation and supercomputer calculation: 1) tuning of MAFIA (three-dimensional electric and magnetic fields calculation code) for supercomputer, 2) examples of three-dimensional calculation of accelerating structures by MAFIA, 3) development of a beam transport code including space charge effects. (author)

  14. A plan for the modification and assessment of TRAC-PF1/MOD2 for use in analyzing CANDU 3 transient thermal-hydraulic phenomena

    International Nuclear Information System (INIS)

    Siebe, D.A.; Boyack, B.E.; Giguere, P.T.

    1994-11-01

    This report presents the results of the review and planning done for the United States Nuclear Regulatory Commission to identify the thermal-hydraulic phenomena that could occur in the CANDU 3 reactor design during transient conditions, plan modifications to the TRAC-PF1/MOD2 (TRAC) computer code needed to adequately predict CANDU 3 transient thermal-hydraulic phenomena, and identify an assessment program to verify the ability of TRAC, when modified, to predict these phenomena. This work builds on analyses and recommendations produced by the Idaho National Engineering Laboratory (INEL). To identify the thermal-hydraulic phenomena, a large-break loss-of-coolant accident simulation, performed as part of earlier work by INEL with an Atomic Energy of Canada, Limited (AECL) thermal-hydraulic computer code (CATHENA), was analyzed in detail. Other accident scenarios were examined for additional phenomena. A group of Los Alamos National Laboratory reactor thermal-hydraulics experts ranked the phenomena to produce a preliminary phenomena identification and ranking table (PIRT). The preliminary nature of the PIRT was a result of a lack of direct expertise with the unique processes and phenomena of the CANDU 3. Nonetheless, this PIRT provided an adequate foundation for planning a program of code modifications. We believe that this PIRT captured the most important phenomena and that refinements to the PIRT will mainly produce clarification of the relative importance (ranking) of phenomena. A plan for code modifications was developed based on this PIRT and on information about the modeling methodologies for CANDU-specific phenomena used in AECL codes. AECL thermal-hydraulic test facilities and programs were reviewed and the information used in developing an assessment plan to ensure that TRAC-PF1/MOD2, when modified, will adequately predict CANDU 3 phenomena

  15. Computer codes in particle transport physics

    International Nuclear Information System (INIS)

    Pesic, M.

    2004-01-01

    Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option

  16. Computer code development plant for SMART design

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H.

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  17. Computer code development plant for SMART design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  18. Concentration - dose - risk computer code

    International Nuclear Information System (INIS)

    Frujinoiu, C.; Preda, M.

    1997-01-01

    Generally, the society is less willing in promoting remedial actions in case of low level chronic exposure situations. Radon in dwellings and workplaces is a case connected to chronic exposure. Apart from radon, the solely source on which the international community agreed for setting action levels, there are other numerous sources technically modified by man that can generate chronic exposure. Even if the nuclear installations are the most relevant, we are surrounded by 'man-made radioactivity' such as: mining industry, coal-fired power plants and fertilizer industry. The operating of an installation even within 'normal limits' could generate chronic exposure due to accumulation of the pollutants after a definite time. This asymptotic proclivity to a constant level define a steady-state concentration that represents a characteristic of the source's presence in the environment. The paper presents a methodology and a code package that derives sequentially the steady-state concentration, doses, detriments, as well as the costs of the effects of installation operation in a given environment. (authors)

  19. Quantum computing with Majorana fermion codes

    Science.gov (United States)

    Litinski, Daniel; von Oppen, Felix

    2018-05-01

    We establish a unified framework for Majorana-based fault-tolerant quantum computation with Majorana surface codes and Majorana color codes. All logical Clifford gates are implemented with zero-time overhead. This is done by introducing a protocol for Pauli product measurements with tetrons and hexons which only requires local 4-Majorana parity measurements. An analogous protocol is used in the fault-tolerant setting, where tetrons and hexons are replaced by Majorana surface code patches, and parity measurements are replaced by lattice surgery, still only requiring local few-Majorana parity measurements. To this end, we discuss twist defects in Majorana fermion surface codes and adapt the technique of twist-based lattice surgery to fermionic codes. Moreover, we propose a family of codes that we refer to as Majorana color codes, which are obtained by concatenating Majorana surface codes with small Majorana fermion codes. Majorana surface and color codes can be used to decrease the space overhead and stabilizer weight compared to their bosonic counterparts.

  20. Gender codes why women are leaving computing

    CERN Document Server

    Misa, Thomas J

    2010-01-01

    The computing profession is facing a serious gender crisis. Women are abandoning the computing field at an alarming rate. Fewer are entering the profession than anytime in the past twenty-five years, while too many are leaving the field in mid-career. With a maximum of insight and a minimum of jargon, Gender Codes explains the complex social and cultural processes at work in gender and computing today. Edited by Thomas Misa and featuring a Foreword by Linda Shafer, Chair of the IEEE Computer Society Press, this insightful collection of essays explores the persisting gender imbalance in computing and presents a clear course of action for turning things around.

  1. Turbo Pascal Computer Code for PIXE Analysis

    International Nuclear Information System (INIS)

    Darsono

    2002-01-01

    To optimal utilization of the 150 kV ion accelerator facilities and to govern the analysis technique using ion accelerator, the research and development of low energy PIXE technology has been done. The R and D for hardware of the low energy PIXE installation in P3TM have been carried on since year 2000. To support the R and D of PIXE accelerator facilities in harmonize with the R and D of the PIXE hardware, the development of PIXE software for analysis is also needed. The development of database of PIXE software for analysis using turbo Pascal computer code is reported in this paper. This computer code computes the ionization cross-section, the fluorescence yield, and the stopping power of elements also it computes the coefficient attenuation of X- rays energy. The computer code is named PIXEDASIS and it is part of big computer code planed for PIXE analysis that will be constructed in the near future. PIXEDASIS is designed to be communicative with the user. It has the input from the keyboard. The output shows in the PC monitor, which also can be printed. The performance test of the PIXEDASIS shows that it can be operated well and it can provide data agreement with data form other literatures. (author)

  2. Blow.MOD2: a program for blowdown transient calculations

    International Nuclear Information System (INIS)

    Doval, A.

    1990-01-01

    The BLOW.MOD2 program has been developed to calculate the blowdown phase in a pressurized vessel after a break/valve is opened. It is a one volume model where break height and flow area are specified. Moody critical flow model was adopted under saturation conditions for flow calculation through the break. Heat transfer from structures and internals have been taken into account. Long term depressurization results and a more complex model are compared satisfactorily. (Author)

  3. New coding technique for computer generated holograms.

    Science.gov (United States)

    Haskell, R. E.; Culver, B. C.

    1972-01-01

    A coding technique is developed for recording computer generated holograms on a computer controlled CRT in which each resolution cell contains two beam spots of equal size and equal intensity. This provides a binary hologram in which only the position of the two dots is varied from cell to cell. The amplitude associated with each resolution cell is controlled by selectively diffracting unwanted light into a higher diffraction order. The recording of the holograms is fast and simple.

  4. Assessment of PWR Steam Generator modelling in RELAP5/MOD2

    International Nuclear Information System (INIS)

    Putney, J.M.; Preece, R.J.

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3

  5. LATTICE: an interactive lattice computer code

    International Nuclear Information System (INIS)

    Staples, J.

    1976-10-01

    LATTICE is a computer code which enables an interactive user to calculate the functions of a synchrotron lattice. This program satisfies the requirements at LBL for a simple interactive lattice program by borrowing ideas from both TRANSPORT and SYNCH. A fitting routine is included

  6. Citham-2 computer code-User manual

    International Nuclear Information System (INIS)

    Batista, J.L.

    1984-01-01

    The procedures and the input data for the Citham-2 computer code are described. It is a subroutine that modifies the nuclide concentration taking in account its burn and prepares cross sections library in 2,3 or 4 energy groups, to the used for Citation program. (E.G.) [pt

  7. Computer Security: is your code sane?

    CERN Multimedia

    Stefan Lueders, Computer Security Team

    2015-01-01

    How many of us write code? Software? Programs? Scripts? How many of us are properly trained in this and how well do we do it? Do we write functional, clean and correct code, without flaws, bugs and vulnerabilities*? In other words: are our codes sane?   Figuring out weaknesses is not that easy (see our quiz in an earlier Bulletin article). Therefore, in order to improve the sanity of your code, prevent common pit-falls, and avoid the bugs and vulnerabilities that can crash your code, or – worse – that can be misused and exploited by attackers, the CERN Computer Security team has reviewed its recommendations for checking the security compliance of your code. “Static Code Analysers” are stand-alone programs that can be run on top of your software stack, regardless of whether it uses Java, C/C++, Perl, PHP, Python, etc. These analysers identify weaknesses and inconsistencies including: employing undeclared variables; expressions resu...

  8. Evaluation of the SCANAIR Computer Code

    International Nuclear Information System (INIS)

    Jernkvist, Lars Olof; Massih, Ali

    2001-11-01

    The SCANAIR computer code, version 3.2, has been evaluated from the standpoint of its capability to analyze, simulate and predict nuclear fuel behavior during severe power transients. SCANAIR calculates the thermal and mechanical behavior of a pressurized water reactor (PWR) fuel rod during a postulated reactivity initiated accident (RIA), and our evaluation indicates that SCANAIR is a state of the art computational tool for this purpose. Our evaluation starts by reviewing the basic theoretical models in SCANAIR, namely the governing equations for heat transfer, the mechanical response of fuel and clad, and the fission gas release behavior. The numerical methods used to solve the governing equations are briefly reviewed, and the range of applicability of the models and their limitations are discussed and illustrated with examples. Next, the main features of the SCANAIR user interface are delineated. The code requires an extensive amount of input data, in order to define burnup-dependent initial conditions to the simulated RIA. These data must be provided in a special format by a thermal-mechanical fuel rod analysis code. The user also has to supply the transient power history under RIA as input, which requires a code for neutronics calculation. The programming structure and documentation of the code are also addressed in our evaluation. SCANAIR is programmed in Fortran-77, and makes use of several general Fortran-77 libraries for handling input/output, data storage and graphical presentation of computed results. The documentation of SCANAIR and its helping libraries is generally of good quality. A drawback with SCANAIR in its present form, is that the code and its pre- and post-processors are tied to computers running the Unix or Linux operating systems. As part of our evaluation, we have performed a large number of computations with SCANAIR, some of which are documented in this report. The computations presented here include a hypothetical RIA in a high

  9. Concatenated codes for fault tolerant quantum computing

    Energy Technology Data Exchange (ETDEWEB)

    Knill, E.; Laflamme, R.; Zurek, W.

    1995-05-01

    The application of concatenated codes to fault tolerant quantum computing is discussed. We have previously shown that for quantum memories and quantum communication, a state can be transmitted with error {epsilon} provided each gate has error at most c{epsilon}. We show how this can be used with Shor`s fault tolerant operations to reduce the accuracy requirements when maintaining states not currently participating in the computation. Viewing Shor`s fault tolerant operations as a method for reducing the error of operations, we give a concatenated implementation which promises to propagate the reduction hierarchically. This has the potential of reducing the accuracy requirements in long computations.

  10. An assessment of the annular flow transition criteria and interphase friction models in RELAP5/MOD2

    International Nuclear Information System (INIS)

    Putney, J.M.

    1989-02-01

    An assessment of the annular flow transition criteria and interphase friction models for two-phase flow in tubes used in RELAP5/MOD2 code is described. The assessment examines the theoretical bases for the criteria and models and considers the results of comparisons with experimental data. Several deficiencies in the transition criteria are identified and appropriate improvements proposed. The interphase friction models are found to be adequate for PWR analyses. (author)

  11. Computer codes used in particle accelerator design: First edition

    International Nuclear Information System (INIS)

    1987-01-01

    This paper contains a listing of more than 150 programs that have been used in the design and analysis of accelerators. Given on each citation are person to contact, classification of the computer code, publications describing the code, computer and language runned on, and a short description of the code. Codes are indexed by subject, person to contact, and code acronym

  12. Implementing a modular system of computer codes

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1983-07-01

    A modular computation system has been developed for nuclear reactor core analysis. The codes can be applied repeatedly in blocks without extensive user input data, as needed for reactor history calculations. The primary control options over the calculational paths and task assignments within the codes are blocked separately from other instructions, admitting ready access by user input instruction or directions from automated procedures and promoting flexible and diverse applications at minimum application cost. Data interfacing is done under formal specifications with data files manipulated by an informed manager. This report emphasizes the system aspects and the development of useful capability, hopefully informative and useful to anyone developing a modular code system of much sophistication. Overall, this report in a general way summarizes the many factors and difficulties that are faced in making reactor core calculations, based on the experience of the authors. It provides the background on which work on HTGR reactor physics is being carried out

  13. Present state of the SOURCES computer code

    International Nuclear Information System (INIS)

    Shores, Erik F.

    2002-01-01

    In various stages of development for over two decades, the SOURCES computer code continues to calculate neutron production rates and spectra from four types of problems: homogeneous media, two-region interfaces, three-region interfaces and that of a monoenergetic alpha particle beam incident on a slab of target material. Graduate work at the University of Missouri - Rolla, in addition to user feedback from a tutorial course, provided the impetus for a variety of code improvements. Recently upgraded to version 4B, initial modifications to SOURCES focused on updates to the 'tape5' decay data library. Shortly thereafter, efforts focused on development of a graphical user interface for the code. This paper documents the Los Alamos SOURCES Tape1 Creator and Library Link (LASTCALL) and describes additional library modifications in more detail. Minor improvements and planned enhancements are discussed.

  14. (Nearly) portable PIC code for parallel computers

    International Nuclear Information System (INIS)

    Decyk, V.K.

    1993-01-01

    As part of the Numerical Tokamak Project, the author has developed a (nearly) portable, one dimensional version of the GCPIC algorithm for particle-in-cell codes on parallel computers. This algorithm uses a spatial domain decomposition for the fields, and passes particles from one domain to another as the particles move spatially. With only minor changes, the code has been run in parallel on the Intel Delta, the Cray C-90, the IBM ES/9000 and a cluster of workstations. After a line by line translation into cmfortran, the code was also run on the CM-200. Impressive speeds have been achieved, both on the Intel Delta and the Cray C-90, around 30 nanoseconds per particle per time step. In addition, the author was able to isolate the data management modules, so that the physics modules were not changed much from their sequential version, and the data management modules can be used as open-quotes black boxes.close quotes

  15. Evaluation of the FRAPCON-3 Computer Code

    International Nuclear Information System (INIS)

    Jernkvist, Lars Olof; Massih, Ali

    2002-03-01

    The FRAPCON-3 computer code has been evaluated with respect to its applicability, modeling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under steady-state operational conditions and moderate power excursions. It is applicable to both boiling- and pressurized water reactor fuel rods with UO 2 fuel, ranging up to about 65 MWd/kg U in rod average burnup. The models and numerical methods in FRAPCON-3 are relatively simple, which makes the code transparent and also fairly easy to modify and extend for the user. The fundamental equations for heat transfer, structural analysis and fuel fission gas release are solved in one-dimensional (radial) and stationary (time-independent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow and rod internal gas pressure. The code is fairly easy to use; fuel rod design data and time histories of fuel rod power and coolant inlet conditions are input via a single text file, and the corresponding calculated variation with time of important fuel rod parameters are printed to a single output file in textual form. The results can also be presented in graphical form through an interface to the general graphics program xmgr. FRAPCON-3 also provides the possibility to export calculated results to the transient fuel rod analysis code FRAPTRAN, where the data can be used as burnup-dependent initial conditions to a postulated transient. Most of the source code to FRAPCON-3 is written in Fortran-IV, which is an archaic, non-standard dialect of the Fortran programming language. Since Fortran-IV is not accepted by all compilers for the latest standard of the language, Fortran-95, there is a risk that the source code must be partly rewritten in the future. Documentation of the code comprises (i) a general code description, which briefly presents models

  16. Evaluation of the FRAPCON-3 Computer Code

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala (Sweden)

    2002-03-01

    The FRAPCON-3 computer code has been evaluated with respect to its applicability, modeling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under steady-state operational conditions and moderate power excursions. It is applicable to both boiling- and pressurized water reactor fuel rods with UO{sub 2} fuel, ranging up to about 65 MWd/kg U in rod average burnup. The models and numerical methods in FRAPCON-3 are relatively simple, which makes the code transparent and also fairly easy to modify and extend for the user. The fundamental equations for heat transfer, structural analysis and fuel fission gas release are solved in one-dimensional (radial) and stationary (time-independent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow and rod internal gas pressure. The code is fairly easy to use; fuel rod design data and time histories of fuel rod power and coolant inlet conditions are input via a single text file, and the corresponding calculated variation with time of important fuel rod parameters are printed to a single output file in textual form. The results can also be presented in graphical form through an interface to the general graphics program xmgr. FRAPCON-3 also provides the possibility to export calculated results to the transient fuel rod analysis code FRAPTRAN, where the data can be used as burnup-dependent initial conditions to a postulated transient. Most of the source code to FRAPCON-3 is written in Fortran-IV, which is an archaic, non-standard dialect of the Fortran programming language. Since Fortran-IV is not accepted by all compilers for the latest standard of the language, Fortran-95, there is a risk that the source code must be partly rewritten in the future. Documentation of the code comprises (i) a general code description, which briefly presents models

  17. Computer code to assess accidental pollutant releases

    International Nuclear Information System (INIS)

    Pendergast, M.M.; Huang, J.C.

    1980-07-01

    A computer code was developed to calculate the cumulative frequency distributions of relative concentrations of an air pollutant following an accidental release from a stack or from a building penetration such as a vent. The calculations of relative concentration are based on the Gaussian plume equations. The meteorological data used for the calculation are in the form of joint frequency distributions of wind and atmospheric stability

  18. Poisson/Superfish codes for personal computers

    International Nuclear Information System (INIS)

    Humphries, S.

    1992-01-01

    The Poisson/Superfish codes calculate static E or B fields in two-dimensions and electromagnetic fields in resonant structures. New versions for 386/486 PCs and Macintosh computers have capabilities that exceed the mainframe versions. Notable improvements are interactive graphical post-processors, improved field calculation routines, and a new program for charged particle orbit tracking. (author). 4 refs., 1 tab., figs

  19. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1988-01-01

    This paper gives a collective summary of the studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRANTIC, FTAP, computer code package RALLY, and BOUNDS codes. Two reference study cases were executed by each code. The results obtained logic/probabilistic analysis as well as computation time are compared

  20. Computing Challenges in Coded Mask Imaging

    Science.gov (United States)

    Skinner, Gerald

    2009-01-01

    This slide presaentation reviews the complications and challenges in developing computer systems for Coded Mask Imaging telescopes. The coded mask technique is used when there is no other way to create the telescope, (i.e., when there are wide fields of view, high energies for focusing or low energies for the Compton/Tracker Techniques and very good angular resolution.) The coded mask telescope is described, and the mask is reviewed. The coded Masks for the INTErnational Gamma-Ray Astrophysics Laboratory (INTEGRAL) instruments are shown, and a chart showing the types of position sensitive detectors used for the coded mask telescopes is also reviewed. Slides describe the mechanism of recovering an image from the masked pattern. The correlation with the mask pattern is described. The Matrix approach is reviewed, and other approaches to image reconstruction are described. Included in the presentation is a review of the Energetic X-ray Imaging Survey Telescope (EXIST) / High Energy Telescope (HET), with information about the mission, the operation of the telescope, comparison of the EXIST/HET with the SWIFT/BAT and details of the design of the EXIST/HET.

  1. SALE: Safeguards Analytical Laboratory Evaluation computer code

    International Nuclear Information System (INIS)

    Carroll, D.J.; Bush, W.J.; Dolan, C.A.

    1976-09-01

    The Safeguards Analytical Laboratory Evaluation (SALE) program implements an industry-wide quality control and evaluation system aimed at identifying and reducing analytical chemical measurement errors. Samples of well-characterized materials are distributed to laboratory participants at periodic intervals for determination of uranium or plutonium concentration and isotopic distributions. The results of these determinations are statistically-evaluated, and each participant is informed of the accuracy and precision of his results in a timely manner. The SALE computer code which produces the report is designed to facilitate rapid transmission of this information in order that meaningful quality control will be provided. Various statistical techniques comprise the output of the SALE computer code. Assuming an unbalanced nested design, an analysis of variance is performed in subroutine NEST resulting in a test of significance for time and analyst effects. A trend test is performed in subroutine TREND. Microfilm plots are obtained from subroutine CUMPLT. Within-laboratory standard deviations are calculated in the main program or subroutine VAREST, and between-laboratory standard deviations are calculated in SBLV. Other statistical tests are also performed. Up to 1,500 pieces of data for each nuclear material sampled by 75 (or fewer) laboratories may be analyzed with this code. The input deck necessary to run the program is shown, and input parameters are discussed in detail. Printed output and microfilm plot output are described. Output from a typical SALE run is included as a sample problem

  2. A zero-dimensional EXTRAP computer code

    International Nuclear Information System (INIS)

    Karlsson, P.

    1982-10-01

    A zero-dimensional computer code has been designed for the EXTRAP experiment to predict the density and the temperature and their dependence upon paramenters such as the plasma current and the filling pressure of neutral gas. EXTRAP is a Z-pinch immersed in a vacuum octupole field and could be either linear or toroidal. In this code the density and temperature are assumed to be constant from the axis up to a breaking point from where they decrease linearly in the radial direction out to the plasma radius. All quantities, however, are averaged over the plasma volume thus giving the zero-dimensional character of the code. The particle, momentum and energy one-fluid equations are solved including the effects of the surrounding neutral gas and oxygen impurities. The code shows that the temperature and density are very sensitive to the shape of the plasma, flatter profiles giving higher temperatures and densities. The temperature, however, is not strongly affected for oxygen concentration less than 2% and is well above the radiation barrier even for higher concentrations. (Author)

  3. SKEMA - A computer code to estimate atmospheric dispersion

    International Nuclear Information System (INIS)

    Sacramento, A.M. do.

    1985-01-01

    This computer code is a modified version of DWNWND code, developed in Oak Ridge National Laboratory. The Skema code makes an estimative of concentration in air of a material released in atmosphery, by ponctual source. (C.M.) [pt

  4. The MESORAD dose assessment model: Computer code

    International Nuclear Information System (INIS)

    Ramsdell, J.V.; Athey, G.F.; Bander, T.J.; Scherpelz, R.I.

    1988-10-01

    MESORAD is a dose equivalent model for emergency response applications that is designed to be run on minicomputers. It has been developed by the Pacific Northwest Laboratory for use as part of the Intermediate Dose Assessment System in the US Nuclear Regulatory Commission Operations Center in Washington, DC, and the Emergency Management System in the US Department of Energy Unified Dose Assessment Center in Richland, Washington. This volume describes the MESORAD computer code and contains a listing of the code. The technical basis for MESORAD is described in the first volume of this report (Scherpelz et al. 1986). A third volume of the documentation planned. That volume will contain utility programs and input and output files that can be used to check the implementation of MESORAD. 18 figs., 4 tabs

  5. Neutron spectrum unfolding using computer code SAIPS

    International Nuclear Information System (INIS)

    Karim, S.

    1999-01-01

    The main objective of this project was to study the neutron energy spectrum at rabbit station-1 in Pakistan Research Reactor (PARR-I). To do so, multiple foils activation method was used to get the saturated activities. The computer code SAIPS was used to unfold the neutron spectra from the measured reaction rates. Of the three built in codes in SAIPS, only SANDI and WINDOWS were used. Contribution of thermal part of the spectra was observed to be higher than the fast one. It was found that the WINDOWS gave smooth spectra while SANDII spectra have violet oscillations in the resonance region. The uncertainties in the WINDOWS results are higher than those of SANDII. The results show reasonable agreement with the published results. (author)

  6. Computer code for quantitative ALARA evaluations

    International Nuclear Information System (INIS)

    Voilleque, P.G.

    1984-01-01

    A FORTRAN computer code has been developed to simplify the determination of whether dose reduction actions meet the as low as is reasonably achievable (ALARA) criterion. The calculations are based on the methodology developed for the Atomic Industrial Forum. The code is used for analyses of eight types of dose reduction actions, characterized as follows: reduce dose rate, reduce job frequency, reduce productive working time, reduce crew size, increase administrative dose limit for the task, and increase the workers' time utilization and dose utilization through (a) improved working conditions, (b) basic skill training, or (c) refresher training for special skills. For each type of action, two analysis modes are available. The first is a generic analysis in which the program computes potential benefits (in dollars) for a range of possible improvements, e.g., for a range of lower dose rates. Generic analyses are most useful in the planning stage and for evaluating the general feasibility of alternative approaches. The second is a specific analysis in which the potential annual benefits of a specific level of improvement and the annual implementation cost are compared. The potential benefits reflect savings in operational and societal costs that can be realized if occupational radiation doses are reduced. Because the potential benefits depend upon many variables which characterize the job, the workplace, and the workers, there is no unique relationship between the potential dollar savings and the dose savings. The computer code permits rapid quantitative analyses of alternatives and is a tool that supplements the health physicist's professional judgment. The program output provides a rational basis for decision-making and a record of the assumptions employed

  7. Analog system for computing sparse codes

    Science.gov (United States)

    Rozell, Christopher John; Johnson, Don Herrick; Baraniuk, Richard Gordon; Olshausen, Bruno A.; Ortman, Robert Lowell

    2010-08-24

    A parallel dynamical system for computing sparse representations of data, i.e., where the data can be fully represented in terms of a small number of non-zero code elements, and for reconstructing compressively sensed images. The system is based on the principles of thresholding and local competition that solves a family of sparse approximation problems corresponding to various sparsity metrics. The system utilizes Locally Competitive Algorithms (LCAs), nodes in a population continually compete with neighboring units using (usually one-way) lateral inhibition to calculate coefficients representing an input in an over complete dictionary.

  8. Statistical theory applications and associated computer codes

    International Nuclear Information System (INIS)

    Prince, A.

    1980-01-01

    The general format is along the same lines as that used in the O.M. Session, i.e. an introduction to the nature of the physical problems and methods of solution based on the statistical model of the nucleus. Both binary and higher multiple reactions are considered. The computer codes used in this session are a combination of optical model and statistical theory. As with the O.M. sessions, the preparation of input and analysis of output are thoroughly examined. Again, comparison with experimental data serves to demonstrate the validity of the results and possible areas for improvement. (author)

  9. Nonequilibrium constitutive models for RELAP5/MOD2

    International Nuclear Information System (INIS)

    Lin, J.C.; Trapp, J.A.; Riemke, R.A.; Ransom, V.H.

    1983-01-01

    RELAP5/MOD2 is a new version of RELAP5 containing improved modeling features that provide a generic pressurized-water transient simulation capability. In particular, the nonequilibrium modeling capability has been generalized to include conditions that occur in operational transients including repressurization and emergency-feed-water injection with loss-of-coolant accidents. The improvements include addition of a second energy equation to the hydrodynamic model, addition of nonequilibrium heat-transfer models, and the associated nonequilibrium vapor-generation models. The objective of this paper is to describe these models and to report the developmental assessment results obtained from similar of several separate effects experiments. The assessment shows that RELAP5 calculated results are in good agreement with data and the nonequilibrium phenomena are properly modeled

  10. Computer code validation by high temperature chemistry

    International Nuclear Information System (INIS)

    Alexander, C.A.; Ogden, J.S.

    1988-01-01

    At least five of the computer codes utilized in analysis of severe fuel damage-type events are directly dependent upon or can be verified by high temperature chemistry. These codes are ORIGEN, CORSOR, CORCON, VICTORIA, and VANESA. With the exemption of CORCON and VANESA, it is necessary that verification experiments be performed on real irradiated fuel. For ORIGEN, the familiar knudsen effusion cell is the best choice and a small piece of known mass and known burn-up is selected and volatilized completely into the mass spectrometer. The mass spectrometer is used in the integral mode to integrate the entire signal from preselected radionuclides, and from this integrated signal the total mass of the respective nuclides can be determined. For CORSOR and VICTORIA, experiments with flowing high pressure hydrogen/steam must flow over the irradiated fuel and then enter the mass spectrometer. For these experiments, a high pressure-high temperature molecular beam inlet must be employed. Finally, in support of VANESA-CORCON, the very highest temperature and molten fuels must be contained and analyzed. Results from all types of experiments will be discussed and their applicability to present and future code development will also be covered

  11. MOD-2 wind turbine system concept and preliminary design report. Volume 2: Detailed report

    Science.gov (United States)

    1979-01-01

    The configuration development of the MOD-2 wind turbine system (WTS) is documented. The MOD-2 WTS project is a continuation of DOE programs to develop and achieve early commercialization of wind energy. The MOD-2 is design optimized for commercial production rates which, in multiunit installations, will be integrated into a utility power grid and achieve a cost of electricity at less than four cents per kilowatt hour.

  12. Further development of the computer code ATHLET-CD

    International Nuclear Information System (INIS)

    Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin

    2016-10-01

    developed or improved in the frame of this project: Oxidation in steam and air atmosphere, processes in the lower plenum, melt behaviour, melt relocation into containment after failure of reactor pressure vessel, calculation of the nuclide inventory, release of fission products from the core, transport of fission products in the primary circuit. As a result of these improvements the quality of calculations of experiments as well as postulated reactor transients and accidents has been advanced considerably. Furthermore, the general robustness of the code has been improved. During the reporting period four versions of ATHLET-CD were released and frozen respectively. In October 2012 the version ATHLET-CD Mod 2.2 Cycle C was released as a frozen development version. Main improvements were the module AIDA, which allows the simulation of the late phase effects after relocation of melt into the lower ple-num, the modelling of the nitride formation in case of air ingress and the new version ATHLET 3.0A. In August 2013 the version ATHLET-CD Mod 3.0 Cycle A was released. In comparison to the version 2.2C further extensions and improvements were performed, e. g. general improvements of the code robustness or improvements concerning the simulation of oxidation effects. In contrast to the development version 2.2C the version 3.0A is an official release version, which is available to all licensed users of ATHLET-CD. With the new version the users are able to simulate a complete severe accident sequence from the early phase until the possible failure of the reactor pres-sure vessel after the relocation of melt into the lower plenum. In December 2014 the development version ATHLET-CD Mod 3.0 Cycle B was released. In March 2016 ATHLET-CD 3.1A was frozen. With this version the release of melt from the reactor pressure vessel can be simulated after the failure of the vessel. Furthermore, GRS participated in the international OECD/NEA project ''Benchmark Study of the Accident at the Fukushima

  13. Development of Probabilistic Internal Dosimetry Computer Code

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Siwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Tae-Eun [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Lee, Jai-Ki [Korean Association for Radiation Protection, Seoul (Korea, Republic of)

    2017-02-15

    Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5{sup th}, 5{sup th}, median, 95{sup th}, and 97.5{sup th} percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various

  14. Development of Probabilistic Internal Dosimetry Computer Code

    International Nuclear Information System (INIS)

    Noh, Siwan; Kwon, Tae-Eun; Lee, Jai-Ki

    2017-01-01

    Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5 th , 5 th , median, 95 th , and 97.5 th percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various situations. In cases

  15. 40 CFR 194.23 - Models and computer codes.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Models and computer codes. 194.23... General Requirements § 194.23 Models and computer codes. (a) Any compliance application shall include: (1... obtain stable solutions; (iv) Computer models accurately implement the numerical models; i.e., computer...

  16. Implementation of the thermal-hydraulic transient analysis code RELAP4/MOD5 and MOD6 on the FACOM 230/75 computer system

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Ishigai, Takahiro; Kumakura, Toshimasa; Naraoka, Ken-itsu

    1979-03-01

    Development efforts have continued on the extensively used LOCA analysis code RELAP-4, as seen in its history; that is, from the prototype version MOD2 to the latest one MOD6 which is capable of one-through calculations from blowdown to reflood phase of PWR-LOCA. Many improvements and refinements of the models have enlarged the scopes and extents of phenomena to treat. Correspondingly the size of program has increased version to version, and special programming techniques have continuously been introduced to manage the program within limited capacity of core memory. For example, the Dynamic Storage Allocation of MOD5 and the PRELOAD Preprocessor newly incorporated in MOD6 are those designed for the CDC computer with relatively small core size. Described are these programming techniques in detail and experiences on implementation of the codes on FACOM 230/75, together with some results of confirmatory calculations. (author)

  17. ICAN Computer Code Adapted for Building Materials

    Science.gov (United States)

    Murthy, Pappu L. N.

    1997-01-01

    The NASA Lewis Research Center has been involved in developing composite micromechanics and macromechanics theories over the last three decades. These activities have resulted in several composite mechanics theories and structural analysis codes whose applications range from material behavior design and analysis to structural component response. One of these computer codes, the Integrated Composite Analyzer (ICAN), is designed primarily to address issues related to designing polymer matrix composites and predicting their properties - including hygral, thermal, and mechanical load effects. Recently, under a cost-sharing cooperative agreement with a Fortune 500 corporation, Master Builders Inc., ICAN was adapted to analyze building materials. The high costs and technical difficulties involved with the fabrication of continuous-fiber-reinforced composites sometimes limit their use. Particulate-reinforced composites can be thought of as a viable alternative. They are as easily processed to near-net shape as monolithic materials, yet have the improved stiffness, strength, and fracture toughness that is characteristic of continuous-fiber-reinforced composites. For example, particlereinforced metal-matrix composites show great potential for a variety of automotive applications, such as disk brake rotors, connecting rods, cylinder liners, and other hightemperature applications. Building materials, such as concrete, can be thought of as one of the oldest materials in this category of multiphase, particle-reinforced materials. The adaptation of ICAN to analyze particle-reinforced composite materials involved the development of new micromechanics-based theories. A derivative of the ICAN code, ICAN/PART, was developed and delivered to Master Builders Inc. as a part of the cooperative activity.

  18. A surface code quantum computer in silicon

    Science.gov (United States)

    Hill, Charles D.; Peretz, Eldad; Hile, Samuel J.; House, Matthew G.; Fuechsle, Martin; Rogge, Sven; Simmons, Michelle Y.; Hollenberg, Lloyd C. L.

    2015-01-01

    The exceptionally long quantum coherence times of phosphorus donor nuclear spin qubits in silicon, coupled with the proven scalability of silicon-based nano-electronics, make them attractive candidates for large-scale quantum computing. However, the high threshold of topological quantum error correction can only be captured in a two-dimensional array of qubits operating synchronously and in parallel—posing formidable fabrication and control challenges. We present an architecture that addresses these problems through a novel shared-control paradigm that is particularly suited to the natural uniformity of the phosphorus donor nuclear spin qubit states and electronic confinement. The architecture comprises a two-dimensional lattice of donor qubits sandwiched between two vertically separated control layers forming a mutually perpendicular crisscross gate array. Shared-control lines facilitate loading/unloading of single electrons to specific donors, thereby activating multiple qubits in parallel across the array on which the required operations for surface code quantum error correction are carried out by global spin control. The complexities of independent qubit control, wave function engineering, and ad hoc quantum interconnects are explicitly avoided. With many of the basic elements of fabrication and control based on demonstrated techniques and with simulated quantum operation below the surface code error threshold, the architecture represents a new pathway for large-scale quantum information processing in silicon and potentially in other qubit systems where uniformity can be exploited. PMID:26601310

  19. A surface code quantum computer in silicon.

    Science.gov (United States)

    Hill, Charles D; Peretz, Eldad; Hile, Samuel J; House, Matthew G; Fuechsle, Martin; Rogge, Sven; Simmons, Michelle Y; Hollenberg, Lloyd C L

    2015-10-01

    The exceptionally long quantum coherence times of phosphorus donor nuclear spin qubits in silicon, coupled with the proven scalability of silicon-based nano-electronics, make them attractive candidates for large-scale quantum computing. However, the high threshold of topological quantum error correction can only be captured in a two-dimensional array of qubits operating synchronously and in parallel-posing formidable fabrication and control challenges. We present an architecture that addresses these problems through a novel shared-control paradigm that is particularly suited to the natural uniformity of the phosphorus donor nuclear spin qubit states and electronic confinement. The architecture comprises a two-dimensional lattice of donor qubits sandwiched between two vertically separated control layers forming a mutually perpendicular crisscross gate array. Shared-control lines facilitate loading/unloading of single electrons to specific donors, thereby activating multiple qubits in parallel across the array on which the required operations for surface code quantum error correction are carried out by global spin control. The complexities of independent qubit control, wave function engineering, and ad hoc quantum interconnects are explicitly avoided. With many of the basic elements of fabrication and control based on demonstrated techniques and with simulated quantum operation below the surface code error threshold, the architecture represents a new pathway for large-scale quantum information processing in silicon and potentially in other qubit systems where uniformity can be exploited.

  20. RELAP5/MOD2 analysis of LOFT Experiment L9-3

    International Nuclear Information System (INIS)

    Birchley, J.C.

    1992-04-01

    An analysis has been performed of LOFT Experiment L9-3, a loss-of-feedwater anticipated transient without trip, in order to support the validation of RELAP5/MOD2. Experiment L9-3 exhibited a rapid boildown of the steam generator, following the loss of feed, with the reactor remaining close to its initial power until the steam generator tubes became sufficiently uncovered for primary to secondary heat transfer to be significantly reduced. The ensuing heat up of the primary fluid resulted in a reduction in power induced by the moderator feedback. The primary system pressure increased to the safety relief valve setpoint, before the fall in reactor power allowed the mismatch between primary system heat input and heat removal via the steam generator to be accommodated by cycling of the pilot operated relief valve (PORV). Comparison between calculation and data shows generally good agreement, though with discrepancies in some areas. Weaknesses in the code's treatment of interphase drag and in the representation of the pressuriser spray are indicated, although a shortage of definitive data, particularly in the steam generator, may also be a factor. The overprediction of interphase drag led to a tendency to underpredict the initial inventory in the steam generator and also, perhaps, to overpredict the steam generator heat transfer while the tubes were being uncovered. There is indication that the pressuriser vapour region conditions were close to equilibrium during spray operation. The point kinetics model in RELAP5/MOD2 proved a viable means of representing the power history for this transient

  1. BLOW.MOD2: program for a vessel depressurization calculation with the contribution of structures

    International Nuclear Information System (INIS)

    Doval, A.

    1990-01-01

    The BLOW.MOD2 program developed to calculate pressure vessels' depressurization is presented, considering heat contribution of the structures. The results are opposite to those obtained from other more complex numerical models, being the comparison extremely satisfactory. BLOW.MOD2 is a software of the 'Systems Sub-Branch', INVAP S.E. (Author) [es

  2. Mod-2 wind turbine system concept and preliminary design report. Volume 1: Executive summary

    Science.gov (United States)

    1979-01-01

    The configuration development of the MOD-2 wind turbine system is presented. The MOD-2 is design optimized for commercial production rates which, in multi-unit installations, will be integrated into a utility power grid and achieve a cost of electricity at less than 4 cents per kilowatt hour.

  3. AP600 large-break loss-of-collant-accident developmental assessment plan for TRAC-PF1/MOD2

    International Nuclear Information System (INIS)

    Knight, T.D.

    1996-07-01

    The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems to protect the plant against possible accidents and transients. The design has been submitted to the U.S. NRC for design certification. The NRC has selected the Transient Reactor Analysis Code (TRAC)-PF1/MOD2 for performing large break loss-of coolant-accident (LBLOCA) analysis to support the certification effort. This document defines the tests to be used in the current phase of developmental assessment related to AP600 LBLOCA

  4. Development of computer code in PNC, 3

    International Nuclear Information System (INIS)

    Ohtaki, Akira; Ohira, Hiroaki

    1990-01-01

    Super-COPD, a code which is integrated by calculation modules, has been developed in order to evaluate kinds of dynamics of LMFBR plant by improving COPD. The code involves all models and its advanced models of COPD in module structures. The code makes it possible to simulate the system dynamics of LMFBR plant of any configurations and components. (author)

  5. Hamor-2: a computer code for LWR inventory calculation

    International Nuclear Information System (INIS)

    Guimaraes, L.N.F.; Marzo, M.A.S.

    1985-01-01

    A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt

  6. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author)

  7. Use of computer codes for system reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Sabek, M.; Gaafar, M. (Nuclear Regulatory and Safety Centre, Atomic Energy Authority, Cairo (Egypt)); Poucet, A. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author).

  8. Verification of SACI-2 computer code comparing with experimental results of BIBLIS-A and LOOP-7 computer code

    International Nuclear Information System (INIS)

    Soares, P.A.; Sirimarco, L.F.

    1984-01-01

    SACI-2 is a computer code created to study the dynamic behaviour of a PWR nuclear power plant. To evaluate the quality of its results, SACI-2 was used to recalculate commissioning tests done in BIBLIS-A nuclear power plant and to calculate postulated transients for Angra-2 reactor. The results of SACI-2 computer code from BIBLIS-A showed as much good agreement as those calculated with the KWU Loop 7 computer code for Angra-2. (E.G.) [pt

  9. A study of the dispersed flow interfacial heat transfer model of RELAP5/MOD2.5 and RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Andreani, M. [Swiss Federal Institute of Technology, Zurich (Switzerland); Analytis, G.T.; Aksan, S.N. [Paul Scherrer Institute, Villigen (Switzerland)

    1995-09-01

    The model of interfacial heat transfer for the dispersed flow regime used in the RELAP5 computer codes is investigated in the present paper. Short-transient calculations of two low flooding rate tube reflooding experiments have been performed, where the hydraulic conditions and the heat input to the vapour in the post-dryout region were controlled for the predetermined position of the quench front. Both RELAP5/MOD2.5 and RELAP5/MOD3 substantially underpredicted the exit vapour temperature. The mass flow rate and quality, however, were correct and the heat input to the vapour was larger than the actual one. As the vapour superheat at the tube exit depends on the balance between the heat input from the wall and the heat exchange with the droplets, the discrepancy between the calculated and the measured exit vapour temperature suggested that the inability of both codes to predict the vapour superheat in the dispersed flow region is due to the overprediction of the interfacial heat transfer rate.

  10. Computer codes for neutron data evaluation

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo

    1979-01-01

    Data compilation codes such as NESTOR and REPSTOR, and NDES (Neutron Data Evaluation System) are mainly discussed. NDES is a code for neutron data evaluation using a TSS terminal, TEKTRONIX 4014. Users of NDES can perform plotting of data and calculation with nuclear models under conversational mode. (author)

  11. Calculation study of nonequilibrium post-CHF heat transfer in rod bundle test using modified RELAP5/MOD2

    International Nuclear Information System (INIS)

    Hassan, Y.A.

    1987-01-01

    To date there is only very limited data for non-equilibrium convective film boiling in rod bundle geometries. A recent nine (3 x 3) rod bundle post-critical-flux (CHF) test from the Lehigh University test facility was simulated using RELAP5/MOD2, to assess its capabilities in predicting the overall convective mechanisms in post-CHF heat transfer in rod bundle geometries. The code calculations were compared with experimental data. The code predicted low vapor superheats and void fraction oscillations. A new interfacial heat transfer between the droplet/steam resulted in a reasonable prediction of vapor superheats. A revised dispersed flow film boiling correlation which accounts for the enhancement of steam convective cooling by droplet-induced turbulence was incorporated in the code. Comparison with the data showed a fair agreement

  12. Potential of the MCNP computer code

    International Nuclear Information System (INIS)

    Kyncl, J.

    1995-01-01

    The MCNP code is designed for numerical solution of neutron, photon, and electron transport problems by the Monte Carlo method. The code is based on the linear transport theory of behavior of the differential flux of the particles. The code directly uses data from the cross section point data library for input. Experience is outlined, gained in the application of the code to the calculation of the effective parameters of fuel assemblies and of the entire reactor core, to the determination of the effective parameters of the elementary fuel cell, and to the numerical solution of neutron diffusion and/or transport problems of the fuel assembly. The agreement between the calculated and observed data gives evidence that the MCNP code can be used with advantage for calculations involving WWER type fuel assemblies. (J.B.). 4 figs., 6 refs

  13. Control rod computer code IAMCOS: general theory and numerical methods

    International Nuclear Information System (INIS)

    West, G.

    1982-11-01

    IAMCOS is a computer code for the description of mechanical and thermal behavior of cylindrical control rods for fast breeders. This code version was applied, tested and modified from 1979 to 1981. In this report are described the basic model (02 version), theoretical definitions and computation methods [fr

  14. Implantation of FRAPCON-2 code in HB computer

    International Nuclear Information System (INIS)

    Silva, C.F. da.

    1987-05-01

    The modifications carried out for implanting FRAPCON-2 computer code in the HB DPS-T7 computer are presented. The FRAPCON-2 code calculates thermo-mechanical response during long period of burnup in stationary state for fuel rods of PWR type reactors. (M.C.K.)

  15. Utilization of DRUFAN 01/MOD 02 computer code for the depressurization phase analysis of a postulated loss of coolant accident in Angra 2/3 Nuclear Power Plants

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.; Figueiredo, P.J.M.

    1985-08-01

    The DRUFAN 01/Mod 2 developed by Gesellschaft fur Reaktorsicherheit (GRS) mbh to simulate thermohydraulic behavior of the primary circuit of PWR reactors, during the despressurization phase and initial refilling phase of loss of coolant accidents by great ruptures, is presented. The program simulates the system to be analysed by control volumes-concentrated parameters model - and it is based on numerical solution of conservation equations for mass of water, mass of vapor, quantities of motion and energy, and on the control volume homogeneity hypothesis. The possibilities of thermodynamic disequilibrium, determining mass transfer between liquid and vapor phases assuming that one saturated phase, are considered. The process of computer code implantation in the Honeywell Bull 64 DPS 7 system at CNEN, the modifications done into the program and the application to the despressurization phase analysis of a loss of coolant accident at Angra-2 and Angra-3 reactors are considered. (M.C.K.) [pt

  16. Superimposed Code Theorectic Analysis of DNA Codes and DNA Computing

    Science.gov (United States)

    2010-03-01

    that the hybridization that occurs between a DNA strand and its Watson - Crick complement can be used to perform mathematical computation. This research...ssDNA single stranded DNA WC Watson – Crick A Adenine C Cytosine G Guanine T Thymine ... Watson - Crick (WC) duplex, e.g., TCGCA TCGCA . Note that non-WC duplexes can form and such a formation is called a cross-hybridization. Cross

  17. The SEDA computer code and its utilization for Angra 1

    International Nuclear Information System (INIS)

    Fernandes Filho, T.L.

    1988-11-01

    The implementation of SEDA 2.0 computer code, developed at Ezeiza Atomic Center, Argentine for Angra 1 reactor is described. The SEDA code gives an estimate for radiological consequences of nuclear accidents with release of radiactive materials for the environment. This code is now available for an IBM PC-XT. The computer environment, the files used, data, the programining structure and the models used are presented. The input data and results for two sample case are described. (author) [pt

  18. APC: A new code for Atmospheric Polarization Computations

    International Nuclear Information System (INIS)

    Korkin, Sergey V.; Lyapustin, Alexei I.; Rozanov, Vladimir V.

    2013-01-01

    A new polarized radiative transfer code Atmospheric Polarization Computations (APC) is described. The code is based on separation of the diffuse light field into anisotropic and smooth (regular) parts. The anisotropic part is computed analytically. The smooth regular part is computed numerically using the discrete ordinates method. Vertical stratification of the atmosphere, common types of bidirectional surface reflection and scattering by spherical particles or spheroids are included. A particular consideration is given to computation of the bidirectional polarization distribution function (BPDF) of the waved ocean surface. -- Highlights: •A new code, APC, has been developed. •The code was validated against well-known codes. •The BPDF for an arbitrary Mueller matrix is computed

  19. Computer and compiler effects on code results: status report

    International Nuclear Information System (INIS)

    1996-01-01

    Within the framework of the international effort on the assessment of computer codes, which are designed to describe the overall reactor coolant system (RCS) thermalhydraulic response, core damage progression, and fission product release and transport during severe accidents, there has been a continuous debate as to whether the code results are influenced by different code users or by different computers or compilers. The first aspect, the 'Code User Effect', has been investigated already. In this paper the other aspects will be discussed and proposals are given how to make large system codes insensitive to different computers and compilers. Hardware errors and memory problems are not considered in this report. The codes investigated herein are integrated code systems (e. g. ESTER, MELCOR) and thermalhydraulic system codes with extensions for severe accident simulation (e. g. SCDAP/RELAP, ICARE/CATHARE, ATHLET-CD), and codes to simulate fission product transport (e. g. TRAPMELT, SOPHAEROS). Since all of these codes are programmed in Fortran 77, the discussion herein is based on this programming language although some remarks are made about Fortran 90. Some observations about different code results by using different computers are reported and possible reasons for this unexpected behaviour are listed. Then methods are discussed how to avoid portability problems

  20. Analysis of parallel computing performance of the code MCNP

    International Nuclear Information System (INIS)

    Wang Lei; Wang Kan; Yu Ganglin

    2006-01-01

    Parallel computing can reduce the running time of the code MCNP effectively. With the MPI message transmitting software, MCNP5 can achieve its parallel computing on PC cluster with Windows operating system. Parallel computing performance of MCNP is influenced by factors such as the type, the complexity level and the parameter configuration of the computing problem. This paper analyzes the parallel computing performance of MCNP regarding with these factors and gives measures to improve the MCNP parallel computing performance. (authors)

  1. A Comparative Study of RCS Computation Codes

    National Research Council Canada - National Science Library

    Tong, Chia T; Wah, Ang T; Hwee, Lim K; Philip, Ou S; Heng, Yar K; Rowse, David; Amos, Matthew; Keen, Alan; Pegg, Neil; Thain, Andrew

    2005-01-01

    .... The first test object is a (fictitious) generic missile. It provides a test problem for benchmarking the performance of CEM codes on geometries containing real world deficiencies, such as thin bodies and sharp corners...

  2. TC-13 Mod 0 and Mod 2 Steam Catapult Test Site

    Data.gov (United States)

    Federal Laboratory Consortium — Located on 11,000 feet of test runway, the TC-13 Mod 0 and Mod 2 Steam Catapult Test Site has in-ground catapults identical to those aboard carriers. This test site...

  3. Computer-assisted Particle-in-Cell code development

    International Nuclear Information System (INIS)

    Kawata, S.; Boonmee, C.; Teramoto, T.; Drska, L.; Limpouch, J.; Liska, R.; Sinor, M.

    1997-12-01

    This report presents a new approach for an electromagnetic Particle-in-Cell (PIC) code development by a computer: in general PIC codes have a common structure, and consist of a particle pusher, a field solver, charge and current density collections, and a field interpolation. Because of the common feature, the main part of the PIC code can be mechanically developed on a computer. In this report we use the packages FIDE and GENTRAN of the REDUCE computer algebra system for discretizations of field equations and a particle equation, and for an automatic generation of Fortran codes. The approach proposed is successfully applied to the development of 1.5-dimensional PIC code. By using the generated PIC code the Weibel instability in a plasma is simulated. The obtained growth rate agrees well with the theoretical value. (author)

  4. RELAP5/MOD2 benchmarking study: Critical heat flux under low-flow conditions

    International Nuclear Information System (INIS)

    Ruggles, E.; Williams, P.T.

    1990-01-01

    Experimental studies by Mishima and Ishii performed at Argonne National Laboratory and subsequent experimental studies performed by Mishima and Nishihara have investigated the critical heat flux (CHF) for low-pressure low-mass flux situations where low-quality burnout may occur. These flow situations are relevant to long-term decay heat removal after a loss of forced flow. The transition from burnout at high quality to burnout at low quality causes very low burnout heat flux values. Mishima and Ishii postulated a model for the low-quality burnout based on flow regime transition from churn turbulent to annular flow. This model was validated by both flow visualization and burnout measurements. Griffith et al. also studied CHF in low mass flux, low-pressure situations and correlated data for upflows, counter-current flows, and downflows with the local fluid conditions. A RELAP5/MOD2 CHF benchmarking study was carried out investigating the performance of the code for low-flow conditions. Data from the experimental study by Mishima and Ishii were the basis for the benchmark comparisons

  5. Reducing Computational Overhead of Network Coding with Intrinsic Information Conveying

    DEFF Research Database (Denmark)

    Heide, Janus; Zhang, Qi; Pedersen, Morten V.

    is RLNC (Random Linear Network Coding) and the goal is to reduce the amount of coding operations both at the coding and decoding node, and at the same time remove the need for dedicated signaling messages. In a traditional RLNC system, coding operation takes up significant computational resources and adds...... the coding operations must be performed in a particular way, which we introduce. Finally we evaluate the suggested system and find that the amount of coding can be significantly reduced both at nodes that recode and decode.......This paper investigated the possibility of intrinsic information conveying in network coding systems. The information is embedded into the coding vector by constructing the vector based on a set of predefined rules. This information can subsequently be retrieved by any receiver. The starting point...

  6. PORPST: A statistical postprocessor for the PORMC computer code

    International Nuclear Information System (INIS)

    Eslinger, P.W.; Didier, B.T.

    1991-06-01

    This report describes the theory underlying the PORPST code and gives details for using the code. The PORPST code is designed to do statistical postprocessing on files written by the PORMC computer code. The data written by PORMC are summarized in terms of means, variances, standard deviations, or statistical distributions. In addition, the PORPST code provides for plotting of the results, either internal to the code or through use of the CONTOUR3 postprocessor. Section 2.0 discusses the mathematical basis of the code, and Section 3.0 discusses the code structure. Section 4.0 describes the free-format point command language. Section 5.0 describes in detail the commands to run the program. Section 6.0 provides an example program run, and Section 7.0 provides the references. 11 refs., 1 fig., 17 tabs

  7. SIMCRI: a simple computer code for calculating nuclear criticality parameters

    International Nuclear Information System (INIS)

    Nakamaru, Shou-ichi; Sugawara, Nobuhiko; Naito, Yoshitaka; Katakura, Jun-ichi; Okuno, Hiroshi.

    1986-03-01

    This is a user's manual for a simple criticality calculation code SIMCRI. The code has been developed to facilitate criticality calculation on a single unit of nuclear fuel. SIMCRI makes an extensive survey with a little computing time. Cross section library MGCL for SIMCRI is the same one for the Monte Carlo criticality code KENOIV; it is, therefore, easy to compare the results of the two codes. SIMCRI solves eigenvalue problems and fixed source problems based on the one space point B 1 equation. The results include infinite and effective multiplication factor, critical buckling, migration area, diffusion coefficient and so on. SIMCRI is comprised in the criticality safety evaluation code system JACS. (author)

  8. BEACON/MOD2A analysis of the Arkansas-1 reactor cavity during a hypothetical hot leg break

    International Nuclear Information System (INIS)

    Ramsthaler, J.A.

    1979-01-01

    As part of the evaluation of the new MOD2A version of the BEACON code, the Arkansas-1 reactor cavity was modeled during a hypothetical loss-of-coolant accident. Results of the BEACON analysis were compared with results obtained previously with the COMPARE containment code. Studies were also made investigating some of the BEACON interphasic, timestep control, and wall heat transfer options to assure that these models were working properly and to observe their effects on the results. Descriptions of the Arkansas-1 reactor cavity, initial assumptions during the hypothetical LOCA, and methods of modeling with BEACON are presented. Some of the problems encountered in accurately modeling the penetrations surrounding the hot and cold leg pipes are also discussed

  9. High burnup models in computer code fair

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)

    1997-08-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.

  10. High burnup models in computer code fair

    International Nuclear Information System (INIS)

    Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.

    1997-01-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

  11. Post-test analysis of LOBI BT-01 using RELAP5/MOD2 and RELAP5/MOD3

    International Nuclear Information System (INIS)

    Holmes, B.J.

    1991-08-01

    LOBI is a high pressure, electrically heated integral system test facility simulating a KWU 1300 MW PWR scaled 1:712 by volume, although full scale has been maintained in the vertical direction. This report describes the results of an analysis of test BT-01, which simulates a 10% steam line break. The bulk of the analysis was performed using the Project Version of RELAP5/MOD2, with additional calculations using RELAP5/MOD3 for comparison. The codes provided generally good agreement with data. In particular, the break flows were well modelled, although the mass flow data proved to be unreliable, and this conclusion had to be derived from interpreting other signals. RELAP over-predicted primary/secondary heat transfer in the broken loop, however, leading to a more rapid cool-down of the primary circuit. Furthermore, the primary side pressure response was critically dependent upon the pressuriser behaviour, and the correct timing of the uncovery of the surge line. Inter-phase drag was not well predicted in the broken loop steam generator intermals, although some improvement was seen in the RELAP5/MOD3 predictions. MOD3 gave a reduction in primary/secondary heat transfer during the test pre-conditioning phase, resulting in a lower secondary side pressure at the start of the transient compared with MOD2. (author)

  12. Continuous Materiality: Through a Hierarchy of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jichen Zhu

    2008-01-01

    Full Text Available The legacy of Cartesian dualism inherent in linguistic theory deeply influences current views on the relation between natural language, computer code, and the physical world. However, the oversimplified distinction between mind and body falls short of capturing the complex interaction between the material and the immaterial. In this paper, we posit a hierarchy of codes to delineate a wide spectrum of continuous materiality. Our research suggests that diagrams in architecture provide a valuable analog for approaching computer code in emergent digital systems. After commenting on ways that Cartesian dualism continues to haunt discussions of code, we turn our attention to diagrams and design morphology. Finally we notice the implications a material understanding of code bears for further research on the relation between human cognition and digital code. Our discussion concludes by noticing several areas that we have projected for ongoing research.

  13. Experiment data report for semiscale Mod-2A primary feed and bleed experiment series (Tests S-SR-1 and S-SR-2)

    International Nuclear Information System (INIS)

    Fogdall, S.P.

    1982-10-01

    This report presents test data recorded for Tests S-SR-1 and S-SR-2 of the Semiscale Mod-2A Primary Feed and Bleed Tests. These tests are part of a series of Semiscale tests that investigate the thermal-hydraulic phenomena resulting from a hypothesized loss-of-coolant accident (LOCA) or abnormal operating transient. These tests provide experimental data for assessing the analytical capability of computer codes used in LOCA and operational transient analysis. The primary objectives of Tests S-SR-1 and -2 were to provide data on primary system recovery through the use of primary feed and bleed cooling, with no heat transfer to the secondaries. Data was obtained using high- and low-head pump curves for the safety injection (SI) pumps. This report presents the uninterpreted data from Tests S-SR-1 and -2 for analysis. The data, presented as graphs in engineering units, have been analyzed only to the extent necessary to ensure that they are reasonable and consistent

  14. Code 672 observational science branch computer networks

    Science.gov (United States)

    Hancock, D. W.; Shirk, H. G.

    1988-01-01

    In general, networking increases productivity due to the speed of transmission, easy access to remote computers, ability to share files, and increased availability of peripherals. Two different networks within the Observational Science Branch are described in detail.

  15. Two-dimensional color-code quantum computation

    International Nuclear Information System (INIS)

    Fowler, Austin G.

    2011-01-01

    We describe in detail how to perform universal fault-tolerant quantum computation on a two-dimensional color code, making use of only nearest neighbor interactions. Three defects (holes) in the code are used to represent logical qubits. Triple-defect logical qubits are deformed into isolated triangular sections of color code to enable transversal implementation of all single logical qubit Clifford group gates. Controlled-NOT (CNOT) is implemented between pairs of triple-defect logical qubits via braiding.

  16. Theoretical calculation possibilities of the computer code HAMMER

    International Nuclear Information System (INIS)

    Onusic Junior, J.

    1978-06-01

    With the aim to know the theoretical calculation possibilities of the computer code HAMMER, developed at Savanah River Laboratory, a analysis of the crytical cells assembly of the kind utilized in PWR reactors is made. (L.F.S.) [pt

  17. Computer code qualification program for the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Popov, N.K.; Wren, D.J.; Snell, V.G.; White, A.J.; Boczar, P.G.

    2003-01-01

    Atomic Energy of Canada Ltd (AECL) has developed and implemented a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, and assessment of their applicability in the safety analyses of the ACR design. An outline of the incremental validation program, and an overview of the experimental program in support of the code validation are also presented. An outline of the SQA program used to qualify these computer codes is also briefly presented. To provide context to the differences in the SQA with respect to current CANDUs, the paper also provides an overview of the ACR design features that have an impact on the computer code qualification. (author)

  18. Lattice Boltzmann method fundamentals and engineering applications with computer codes

    CERN Document Server

    Mohamad, A A

    2014-01-01

    Introducing the Lattice Boltzmann Method in a readable manner, this book provides detailed examples with complete computer codes. It avoids the most complicated mathematics and physics without scarifying the basic fundamentals of the method.

  19. Computer codes for level 1 probabilistic safety assessment

    International Nuclear Information System (INIS)

    1990-06-01

    Probabilistic Safety Assessment (PSA) entails several laborious tasks suitable for computer codes assistance. This guide identifies these tasks, presents guidelines for selecting and utilizing computer codes in the conduct of the PSA tasks and for the use of PSA results in safety management and provides information on available codes suggested or applied in performing PSA in nuclear power plants. The guidance is intended for use by nuclear power plant system engineers, safety and operating personnel, and regulators. Large efforts are made today to provide PC-based software systems and PSA processed information in a way to enable their use as a safety management tool by the nuclear power plant overall management. Guidelines on the characteristics of software needed for management to prepare a software that meets their specific needs are also provided. Most of these computer codes are also applicable for PSA of other industrial facilities. The scope of this document is limited to computer codes used for the treatment of internal events. It does not address other codes available mainly for the analysis of external events (e.g. seismic analysis) flood and fire analysis. Codes discussed in the document are those used for probabilistic rather than for phenomenological modelling. It should be also appreciated that these guidelines are not intended to lead the user to selection of one specific code. They provide simply criteria for the selection. Refs and tabs

  20. GASFLOW computer code (physical models and input data)

    International Nuclear Information System (INIS)

    Muehlbauer, Petr

    2007-11-01

    The GASFLOW computer code was developed jointly by the Los Alamos National Laboratory, USA, and Forschungszentrum Karlsruhe, Germany. The code is primarily intended for calculations of the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and in other facilities. The physical models and the input data are described, and a commented simple calculation is presented

  1. User manual of FRAPCON-I computer code

    International Nuclear Information System (INIS)

    Chia, C.T.

    1985-11-01

    The manual for using the FRAPCON-I code implanted by Reactor Department of Brazilian-CNEN to convert IBM FORTRAN in FORTRAN 77 of Honeywell Bull computer is presented. The FRAPCON-I code describes the behaviour of fuel rods of PWR type reactors at stationary state during long periods of burnup. (M.C.K.)

  2. Study of nuclear computer code maintenance and management system

    International Nuclear Information System (INIS)

    Ryu, Chang Mo; Kim, Yeon Seung; Eom, Heung Seop; Lee, Jong Bok; Kim, Ho Joon; Choi, Young Gil; Kim, Ko Ryeo

    1989-01-01

    Software maintenance is one of the most important problems since late 1970's.We wish to develop a nuclear computer code system to maintenance and manage KAERI's nuclear software. As a part of this system, we have developed three code management programs for use on CYBER and PC systems. They are used in systematic management of computer code in KAERI. The first program is embodied on the CYBER system to rapidly provide information on nuclear codes to the users. The second and the third programs were embodied on the PC system for the code manager and for the management of data in korean language, respectively. In the requirement analysis, we defined each code, magnetic tape, manual and abstract information data. In the conceptual design, we designed retrieval, update, and output functions. In the implementation design, we described the technical considerations of database programs, utilities, and directions for the use of databases. As a result of this research, we compiled the status of nuclear computer codes which belonged KAERI until September, 1988. Thus, by using these three database programs, we could provide the nuclear computer code information to the users more rapidly. (Author)

  3. Nonuniform code concatenation for universal fault-tolerant quantum computing

    Science.gov (United States)

    Nikahd, Eesa; Sedighi, Mehdi; Saheb Zamani, Morteza

    2017-09-01

    Using transversal gates is a straightforward and efficient technique for fault-tolerant quantum computing. Since transversal gates alone cannot be computationally universal, they must be combined with other approaches such as magic state distillation, code switching, or code concatenation to achieve universality. In this paper we propose an alternative approach for universal fault-tolerant quantum computing, mainly based on the code concatenation approach proposed in [T. Jochym-O'Connor and R. Laflamme, Phys. Rev. Lett. 112, 010505 (2014), 10.1103/PhysRevLett.112.010505], but in a nonuniform fashion. The proposed approach is described based on nonuniform concatenation of the 7-qubit Steane code with the 15-qubit Reed-Muller code, as well as the 5-qubit code with the 15-qubit Reed-Muller code, which lead to two 49-qubit and 47-qubit codes, respectively. These codes can correct any arbitrary single physical error with the ability to perform a universal set of fault-tolerant gates, without using magic state distillation.

  4. APC: A New Code for Atmospheric Polarization Computations

    Science.gov (United States)

    Korkin, Sergey V.; Lyapustin, Alexei I.; Rozanov, Vladimir V.

    2014-01-01

    A new polarized radiative transfer code Atmospheric Polarization Computations (APC) is described. The code is based on separation of the diffuse light field into anisotropic and smooth (regular) parts. The anisotropic part is computed analytically. The smooth regular part is computed numerically using the discrete ordinates method. Vertical stratification of the atmosphere, common types of bidirectional surface reflection and scattering by spherical particles or spheroids are included. A particular consideration is given to computation of the bidirectional polarization distribution function (BPDF) of the waved ocean surface.

  5. Hauser*5, a computer code to calculate nuclear cross sections

    International Nuclear Information System (INIS)

    Mann, F.M.

    1979-07-01

    HAUSER*5 is a computer code that uses the statistical (Hauser-Feshbach) model, the pre-equilibrium model, and a statistical model of direct reactions to predict nuclear cross sections. The code is unrestricted as to particle type, includes fission and capture, makes width-fluctuation corrections, and performs three-body calculations - all in minimum computer time. Transmission coefficients can be generated internally or supplied externally. This report describes equations used, necessary input, and resulting output. 2 figures, 4 tables

  6. Computer codes developed in FRG to analyse hypothetical meltdown accidents

    International Nuclear Information System (INIS)

    Hassmann, K.; Hosemann, J.P.; Koerber, H.; Reineke, H.

    1978-01-01

    It is the purpose of this paper to give the status of all significant computer codes developed in the core melt-down project which is incorporated in the light water reactor safety research program of the Federal Ministry of Research and Technology. For standard pressurized water reactors, results of some computer codes will be presented, describing the course and the duration of the hypothetical core meltdown accident. (author)

  7. The computer code EURDYN-1M (release 2). User's manual

    International Nuclear Information System (INIS)

    1982-01-01

    EURDYN-1M is a finite element computer code developed at J.R.C. Ispra to compute the response of two-dimensional coupled fluid-structure configurations to transient dynamic loading for reactor safety studies. This report gives instructions for preparing input data to EURDYN-1M, release 2, and describes a test problem in order to illustrate both the input and the output of the code

  8. Heat Transfer treatment in computer codes for safety analysis

    International Nuclear Information System (INIS)

    Jerele, A.; Gregoric, M.

    1984-01-01

    Increased number of operating nuclear power plants has stressed importance of nuclear safety evaluation. For this reason, accordingly to regulatory commission request, safety analyses with computer codes are preformed. In this paper part of this thermohydraulic models dealing with wall-to-fluid heat transfer correlations in computer codes TRAC=PF1, RELAP4/MOD5, RELAP5/MOD1 and COBRA-IV is discussed. (author)

  9. Two-phase computer codes for zero-gravity applications

    International Nuclear Information System (INIS)

    Krotiuk, W.J.

    1986-10-01

    This paper discusses the problems existing in the development of computer codes which can analyze the thermal-hydraulic behavior of two-phase fluids especially in low gravity nuclear reactors. The important phenomenon affecting fluid flow and heat transfer in reduced gravity is discussed. The applicability of using existing computer codes for space applications is assessed. Recommendations regarding the use of existing earth based fluid flow and heat transfer correlations are made and deficiencies in these correlations are identified

  10. Los Alamos radiation transport code system on desktop computing platforms

    International Nuclear Information System (INIS)

    Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.; West, J.T.

    1990-01-01

    The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. The current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines

  11. A restructuring of CF package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, K. R.; Kim, D. H.; Cho, S. W.

    2004-01-01

    CF package, which evaluates user-specified 'control functions' and applies them to define or control various aspects of computation, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the CF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory, difficulty is more over because its data is location information of other package's data due to characteristics of CF package. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  12. Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)

    International Nuclear Information System (INIS)

    Maiorino, J.R.

    1990-01-01

    The user manual of the ANISN computer code describing input and output subroutines is presented. ANISN code was developed to solve one-dimensional transport equation for neutron or gamma rays in slab, sphere or cylinder geometry with general anisotropic scattering. The solution technique is the discrete ordinate method. (M.C.K.)

  13. Computer codes incorporating pre-equilibrium decay

    International Nuclear Information System (INIS)

    Prince, A.

    1980-01-01

    After establishing the need to describe the high-energy particle spectrum which is evident in the experimental data, the various models used in the interpretation are presented. This includes the following: a) Cascade Model; b) Fermi-Gas Relaxation Model; c) Exciton Model; d) Hybrid and Geometry-Dependent Model. The codes description and preparation of input data for STAPRE was presented (Dr. Strohmaier). A simulated output was employed for a given input and comparison with experimental data substantiated the rather sophisticated treatment. (author)

  14. Adaptation of HAMMER computer code to CYBER 170/750 computer

    International Nuclear Information System (INIS)

    Pinheiro, A.M.B.S.; Nair, R.P.K.

    1982-01-01

    The adaptation of HAMMER computer code to CYBER 170/750 computer is presented. The HAMMER code calculates cell parameters by multigroup transport theory and reactor parameters by few group diffusion theory. The auxiliary programs, the carried out modifications and the use of HAMMER system adapted to CYBER 170/750 computer are described. (M.C.K.) [pt

  15. The archaeology of computer codes - illustrated on the basis of the code SABINE

    International Nuclear Information System (INIS)

    Sdouz, G.

    1987-02-01

    Computer codes used by the physics group of the Institute for Reactor Safety are stored on back-up-tapes. However during the last years both the computer and the system have been changed. For new tasks these programmes have to be available. A new procedure is necessary to find and to activate a stored programme. This procedure is illustrated on the basis of the code SABINE. (Author)

  16. Case studies in Gaussian process modelling of computer codes

    International Nuclear Information System (INIS)

    Kennedy, Marc C.; Anderson, Clive W.; Conti, Stefano; O'Hagan, Anthony

    2006-01-01

    In this paper we present a number of recent applications in which an emulator of a computer code is created using a Gaussian process model. Tools are then applied to the emulator to perform sensitivity analysis and uncertainty analysis. Sensitivity analysis is used both as an aid to model improvement and as a guide to how much the output uncertainty might be reduced by learning about specific inputs. Uncertainty analysis allows us to reflect output uncertainty due to unknown input parameters, when the finished code is used for prediction. The computer codes themselves are currently being developed within the UK Centre for Terrestrial Carbon Dynamics

  17. Low Computational Complexity Network Coding For Mobile Networks

    DEFF Research Database (Denmark)

    Heide, Janus

    2012-01-01

    Network Coding (NC) is a technique that can provide benefits in many types of networks, some examples from wireless networks are: In relay networks, either the physical or the data link layer, to reduce the number of transmissions. In reliable multicast, to reduce the amount of signaling and enable......-flow coding technique. One of the key challenges of this technique is its inherent computational complexity which can lead to high computational load and energy consumption in particular on the mobile platforms that are the target platform in this work. To increase the coding throughput several...

  18. Statistical screening of input variables in a complex computer code

    International Nuclear Information System (INIS)

    Krieger, T.J.

    1982-01-01

    A method is presented for ''statistical screening'' of input variables in a complex computer code. The object is to determine the ''effective'' or important input variables by estimating the relative magnitudes of their associated sensitivity coefficients. This is accomplished by performing a numerical experiment consisting of a relatively small number of computer runs with the code followed by a statistical analysis of the results. A formula for estimating the sensitivity coefficients is derived. Reference is made to an earlier work in which the method was applied to a complex reactor code with good results

  19. Computer codes for beam dynamics analysis of cyclotronlike accelerators

    Science.gov (United States)

    Smirnov, V.

    2017-12-01

    Computer codes suitable for the study of beam dynamics in cyclotronlike (classical and isochronous cyclotrons, synchrocyclotrons, and fixed field alternating gradient) accelerators are reviewed. Computer modeling of cyclotron segments, such as the central zone, acceleration region, and extraction system is considered. The author does not claim to give a full and detailed description of the methods and algorithms used in the codes. Special attention is paid to the codes already proven and confirmed at the existing accelerating facilities. The description of the programs prepared in the worldwide known accelerator centers is provided. The basic features of the programs available to users and limitations of their applicability are described.

  20. Multitasking the code ARC3D. [for computational fluid dynamics

    Science.gov (United States)

    Barton, John T.; Hsiung, Christopher C.

    1986-01-01

    The CRAY multitasking system was developed in order to utilize all four processors and sharply reduce the wall clock run time. This paper describes the techniques used to modify the computational fluid dynamics code ARC3D for this run and analyzes the achieved speedup. The ARC3D code solves either the Euler or thin-layer N-S equations using an implicit approximate factorization scheme. Results indicate that multitask processing can be used to achieve wall clock speedup factors of over three times, depending on the nature of the program code being used. Multitasking appears to be particularly advantageous for large-memory problems running on multiple CPU computers.

  1. Code system to compute radiation dose in human phantoms

    International Nuclear Information System (INIS)

    Ryman, J.C.; Cristy, M.; Eckerman, K.F.; Davis, J.L.; Tang, J.S.; Kerr, G.D.

    1986-01-01

    Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods

  2. Thermohydraulic analysis of nuclear power plant accidents by computer codes

    International Nuclear Information System (INIS)

    Petelin, S.; Stritar, A.; Istenic, R.; Gregoric, M.; Jerele, A.; Mavko, B.

    1982-01-01

    RELAP4/MOD6, BRUCH-D-06, CONTEMPT-LT-28, RELAP5/MOD1 and COBRA-4-1 codes were successful y implemented at the CYBER 172 computer in Ljubljana. Input models of NPP Krsko for the first three codes were prepared. Because of the high computer cost only one analysis of double ended guillotine break of the cold leg of NPP Krsko by RELAP4 code has been done. BRUCH code is easier and cheaper for use. Several analysis have been done. Sensitivity study was performed with CONTEMPT-LT-28 for double ended pump suction break. These codes are intended to be used as a basis for independent safety analyses. (author)

  3. MISER-I: a computer code for JOYO fuel management

    International Nuclear Information System (INIS)

    Yamashita, Yoshioki

    1976-06-01

    A computer code ''MISER-I'' is for a nuclear fuel management of Japan Experimental Fast Breeder Reactor JOYO. The nuclear fuel management in JOYO can be regarded as a fuel assembly management because a handling unit of fuel in JOYO plant is a fuel subassembly (core and blanket subassembly), and so the recording of material balance in computer code is made with each subassembly. The input information into computer code is given with each subassembly for a transfer operation, or with one reactor cycle and every one month for a burn-up in reactor core. The output information of MISER-I code is the fuel assembly storage record, fuel storage weight record in each material balance subarea at any specified day, and fuel subassembly transfer history record. Change of nuclear fuel composition and weight due to a burn-up is calculated with JOYO-Monitoring Code by off-line computation system. MISER-I code is written in FORTRAN-IV language for FACOM 230-48 computer. (auth.)

  4. Development Of A Navier-Stokes Computer Code

    Science.gov (United States)

    Yoon, Seokkwan; Kwak, Dochan

    1993-01-01

    Report discusses aspects of development of CENS3D computer code, solving three-dimensional Navier-Stokes equations of compressible, viscous, unsteady flow. Implements implicit finite-difference or finite-volume numerical-integration scheme, called "lower-upper symmetric-Gauss-Seidel" (LU-SGS), offering potential for very low computer time per iteration and for fast convergence.

  5. Nuclear data to support computer code validation

    International Nuclear Information System (INIS)

    Fisher, S.E.; Broadhead, B.L.; DeHart, M.D.; Primm, R.T. III

    1997-04-01

    The rate of plutonium disposition will be a key parameter in determining the degree of success of the Fissile Materials Disposition Program. Estimates of the disposition rate are dependent on neutronics calculations. To ensure that these calculations are accurate, the codes and data should be validated against applicable experimental measurements. Further, before mixed-oxide (MOX) fuel can be fabricated and loaded into a reactor, the fuel vendors, fabricators, fuel transporters, reactor owners and operators, regulatory authorities, and the Department of Energy (DOE) must accept the validity of design calculations. This report presents sources of neutronics measurements that have potential application for validating reactor physics (predicting the power distribution in the reactor core), predicting the spent fuel isotopic content, predicting the decay heat generation rate, certifying criticality safety of fuel cycle facilities, and ensuring adequate radiation protection at the fuel cycle facilities and the reactor. The U.S. in-reactor experience with MOX fuel is first presented, followed by information related to other aspects of the MOX fuel performance information that is valuable to this program, but the data base remains largely proprietary. Thus, this information is not reported here. It is expected that the selected consortium will make the necessary arrangements to procure or have access to the requisite information

  6. Assessment study of RELAP5/MOD2, CYCLE 36. 04 based on spray start-up test for DOEL-4

    Energy Technology Data Exchange (ETDEWEB)

    Moeyaert, P.; Stubbe, E.

    1989-07-01

    This report presents an assessment study for the code RELAP-5 MOD-2 based on a pressurizer spray start-up test of the Doel-4 power plant. Doel-4 is a three loop WESTINGHOUSE PWR plant ordered by the EBES utility with a nominal power rating of 1000 MWe and equipped with preheater type E steam generators. A large series of commissioning tests are normally performed on new plants, of which the so called pressurizer spray and heater test (SU-PR-01) was performed on February 2nd 1985. TRACTEBEL, being the Architect-Engineer for this plant was closely involved with all start-up tests and was responsible for the final approval of the tests.

  7. A restructuring of RN1 package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, D. H.; Kim, K. R.

    2003-01-01

    RN1 package, which is one of two fission product-related packages in MELCOR, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN1 package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1 package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  8. A restructuring of COR package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S.H.; Kim, K.R.; Kim, D.H.

    2004-01-01

    The COR package, which calculates the thermal response of the core and the lower plenum internal structures and models the relocation of the core and lower plenum structural materials, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the COR package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as a waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the COR package addressed in this paper includes a module development, subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerated the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

  9. A restructuring of RN2 package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, D. H.

    2003-01-01

    RN2 package, which is one of two fission product-related package in MELCOR, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN2 package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN2 package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The validation has been done by comparing the results of the modified code with those from the existing code. As the trends are the similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  10. Automated uncertainty analysis methods in the FRAP computer codes

    International Nuclear Information System (INIS)

    Peck, S.O.

    1980-01-01

    A user oriented, automated uncertainty analysis capability has been incorporated in the Fuel Rod Analysis Program (FRAP) computer codes. The FRAP codes have been developed for the analysis of Light Water Reactor fuel rod behavior during steady state (FRAPCON) and transient (FRAP-T) conditions as part of the United States Nuclear Regulatory Commission's Water Reactor Safety Research Program. The objective of uncertainty analysis of these codes is to obtain estimates of the uncertainty in computed outputs of the codes is to obtain estimates of the uncertainty in computed outputs of the codes as a function of known uncertainties in input variables. This paper presents the methods used to generate an uncertainty analysis of a large computer code, discusses the assumptions that are made, and shows techniques for testing them. An uncertainty analysis of FRAP-T calculated fuel rod behavior during a hypothetical loss-of-coolant transient is presented as an example and carried through the discussion to illustrate the various concepts

  11. HUDU: The Hanford Unified Dose Utility computer code

    International Nuclear Information System (INIS)

    Scherpelz, R.I.

    1991-02-01

    The Hanford Unified Dose Utility (HUDU) computer program was developed to provide rapid initial assessment of radiological emergency situations. The HUDU code uses a straight-line Gaussian atmospheric dispersion model to estimate the transport of radionuclides released from an accident site. For dose points on the plume centerline, it calculates internal doses due to inhalation and external doses due to exposure to the plume. The program incorporates a number of features unique to the Hanford Site (operated by the US Department of Energy), including a library of source terms derived from various facilities' safety analysis reports. The HUDU code was designed to run on an IBM-PC or compatible personal computer. The user interface was designed for fast and easy operation with minimal user training. The theoretical basis and mathematical models used in the HUDU computer code are described, as are the computer code itself and the data libraries used. Detailed instructions for operating the code are also included. Appendices to the report contain descriptions of the program modules, listings of HUDU's data library, and descriptions of the verification tests that were run as part of the code development. 14 refs., 19 figs., 2 tabs

  12. Computer Security: better code, fewer problems

    CERN Multimedia

    Stefan Lueders, Computer Security Team

    2016-01-01

    The origin of many security incidents is negligence or unintentional mistakes made by web developers or programmers. In the rush to complete the work, due to skewed priorities, or just to ignorance, basic security principles can be omitted or forgotten.   The resulting vulnerabilities lie dormant until the evil side spots them and decides to hit hard. Computer security incidents in the past have put CERN’s reputation at risk due to websites being defaced with negative messages about the Organization, hash files of passwords being extracted, restricted data exposed… And it all started with a little bit of negligence! If you check out the Top 10 web development blunders, you will see that the most prevalent mistakes are: Not filtering input, e.g. accepting “<“ or “>” in input fields even if only a number is expected.  Not validating that input: you expect a birth date? So why accept letters? &...

  13. Development of computer code in PNC, 8

    International Nuclear Information System (INIS)

    Ohhira, Mitsuru

    1990-01-01

    Private buildings applied base isolation system, are on the practical stage now. So, under Construction and Maintenance Management Office, we are doing an application study of base isolation system to nuclear fuel facilities. On the process of this study, we have developed Dynamic Analysis Program-Base Isolation System (DAP-BS) which is able to run a 32-bit personal computer. Using this program, we can analyze a 3-dimensional structure, and evaluate the various properties of base isolation parts that are divided into maximum 16 blocks. And from the results of some simulation analyses, we thought that DAP-BS had good reliability and marketability. So, we put DAP-BS on the market. (author)

  14. A compendium of computer codes in fault tree analysis

    International Nuclear Information System (INIS)

    Lydell, B.

    1981-03-01

    In the past ten years principles and methods for a unified system reliability and safety analysis have been developed. Fault tree techniques serve as a central feature of unified system analysis, and there exists a specific discipline within system reliability concerned with the theoretical aspects of fault tree evaluation. Ever since the fault tree concept was established, computer codes have been developed for qualitative and quantitative analyses. In particular the presentation of the kinetic tree theory and the PREP-KITT code package has influenced the present use of fault trees and the development of new computer codes. This report is a compilation of some of the better known fault tree codes in use in system reliability. Numerous codes are available and new codes are continuously being developed. The report is designed to address the specific characteristics of each code listed. A review of the theoretical aspects of fault tree evaluation is presented in an introductory chapter, the purpose of which is to give a framework for the validity of the different codes. (Auth.)

  15. A three-dimensional magnetostatics computer code for insertion devices

    International Nuclear Information System (INIS)

    Chubar, O.; Elleaume, P.; Chavanne, J.

    1998-01-01

    RADIA is a three-dimensional magnetostatics computer code optimized for the design of undulators and wigglers. It solves boundary magnetostatics problems with magnetized and current-carrying volumes using the boundary integral approach. The magnetized volumes can be arbitrary polyhedrons with non-linear (iron) or linear anisotropic (permanent magnet) characteristics. The current-carrying elements can be straight or curved blocks with rectangular cross sections. Boundary conditions are simulated by the technique of mirroring. Analytical formulae used for the computation of the field produced by a magnetized volume of a polyhedron shape are detailed. The RADIA code is written in object-oriented C++ and interfaced to Mathematica (Mathematica is a registered trademark of Wolfram Research, Inc.). The code outperforms currently available finite-element packages with respect to the CPU time of the solver and accuracy of the field integral estimations. An application of the code to the case of a wedge-pole undulator is presented

  16. Holonomic surface codes for fault-tolerant quantum computation

    Science.gov (United States)

    Zhang, Jiang; Devitt, Simon J.; You, J. Q.; Nori, Franco

    2018-02-01

    Surface codes can protect quantum information stored in qubits from local errors as long as the per-operation error rate is below a certain threshold. Here we propose holonomic surface codes by harnessing the quantum holonomy of the system. In our scheme, the holonomic gates are built via auxiliary qubits rather than the auxiliary levels in multilevel systems used in conventional holonomic quantum computation. The key advantage of our approach is that the auxiliary qubits are in their ground state before and after each gate operation, so they are not involved in the operation cycles of surface codes. This provides an advantageous way to implement surface codes for fault-tolerant quantum computation.

  17. A restructuring of TF package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S. H.; Song, Y. M.; Kim, D. H.

    2002-01-01

    TF package which defines some interpolation and extrapolation condition through user defined table has been restructured in MIDAS computer code. To do this, data transferring methods of current MELCOR code are modified and adopted into TF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of the meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of TF package addressed in this paper does module development and subroutine modification, and treats MELGEN which is making restart file as well as MELCOR which is processing calculation. The validation has been done by comparing the results of the modified code with those from the existing code, and it is confirmed that the results are the same. It hints that the similar approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  18. Computer codes for problems of isotope and radiation research

    International Nuclear Information System (INIS)

    Remer, M.

    1986-12-01

    A survey is given of computer codes for problems in isotope and radiation research. Altogether 44 codes are described as titles with abstracts. 17 of them are in the INIS scope and are processed individually. The subjects are indicated in the chapter headings: 1) analysis of tracer experiments, 2) spectrum calculations, 3) calculations of ion and electron trajectories, 4) evaluation of gamma irradiation plants, and 5) general software

  19. Sample test cases using the environmental computer code NECTAR

    International Nuclear Information System (INIS)

    Ponting, A.C.

    1984-06-01

    This note demonstrates a few of the many different ways in which the environmental computer code NECTAR may be used. Four sample test cases are presented and described to show how NECTAR input data are structured. Edited output is also presented to illustrate the format of the results. Two test cases demonstrate how NECTAR may be used to study radio-isotopes not explicitly included in the code. (U.K.)

  20. Computer code for calculating personnel doses due to tritium exposures

    International Nuclear Information System (INIS)

    Graham, C.L.; Parlagreco, J.R.

    1977-01-01

    This report describes a computer code written in LLL modified Fortran IV that can be used on a CDC 7600 for calculating personnel doses due to internal exposures to tritium. The code is capable of handling various exposure situations and is also capable of detecting a large variety of data input errors that would lead to errors in the dose assessment. The critical organ is the body water

  1. RADTRAN: a computer code to analyze transportation of radioactive material

    International Nuclear Information System (INIS)

    Taylor, J.M.; Daniel, S.L.

    1977-04-01

    A computer code is presented which predicts the environmental impact of any specific scheme of radioactive material transportation. Results are presented in terms of annual latent cancer fatalities and annual early fatility probability resulting from exposure, during normal transportation or transport accidents. The code is developed in a generalized format to permit wide application including normal transportation analysis; consideration of alternatives; and detailed consideration of specific sectors of industry

  2. COMPBRN III: a computer code for modeling compartment fires

    International Nuclear Information System (INIS)

    Ho, V.; Siu, N.; Apostolakis, G.; Flanagan, G.F.

    1986-07-01

    The computer code COMPBRN III deterministically models the behavior of compartment fires. This code is an improvement of the original COMPBRN codes. It employs a different air entrainment model and numerical scheme to estimate properties of the ceiling hot gas layer model. Moreover, COMPBRN III incorporates a number of improvements in shape factor calculations and error checking, which distinguish it from the COMPBRN II code. This report presents the ceiling hot gas layer model employed by COMPBRN III as well as several other modifications. Information necessary to run COMPBRN III, including descriptions of required input and resulting output, are also presented. Simulation of experiments and a sample problem are included to demonstrate the usage of the code. 37 figs., 46 refs

  3. Survey of computer codes applicable to waste facility performance evaluations

    International Nuclear Information System (INIS)

    Alsharif, M.; Pung, D.L.; Rivera, A.L.; Dole, L.R.

    1988-01-01

    This study is an effort to review existing information that is useful to develop an integrated model for predicting the performance of a radioactive waste facility. A summary description of 162 computer codes is given. The identified computer programs address the performance of waste packages, waste transport and equilibrium geochemistry, hydrological processes in unsaturated and saturated zones, and general waste facility performance assessment. Some programs also deal with thermal analysis, structural analysis, and special purposes. A number of these computer programs are being used by the US Department of Energy, the US Nuclear Regulatory Commission, and their contractors to analyze various aspects of waste package performance. Fifty-five of these codes were identified as being potentially useful on the analysis of low-level radioactive waste facilities located above the water table. The code summaries include authors, identification data, model types, and pertinent references. 14 refs., 5 tabs

  4. FLASH: A finite element computer code for variably saturated flow

    International Nuclear Information System (INIS)

    Baca, R.G.; Magnuson, S.O.

    1992-05-01

    A numerical model was developed for use in performance assessment studies at the INEL. The numerical model, referred to as the FLASH computer code, is designed to simulate two-dimensional fluid flow in fractured-porous media. The code is specifically designed to model variably saturated flow in an arid site vadose zone and saturated flow in an unconfined aquifer. In addition, the code also has the capability to simulate heat conduction in the vadose zone. This report presents the following: description of the conceptual frame-work and mathematical theory; derivations of the finite element techniques and algorithms; computational examples that illustrate the capability of the code; and input instructions for the general use of the code. The FLASH computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of Energy Order 5820.2A

  5. CRACKEL: a computer code for CFR fuel management calculations

    International Nuclear Information System (INIS)

    Burstall, R.F.; Ball, M.A.; Thornton, D.E.J.

    1975-12-01

    The CRACKLE computer code is designed to perform rapid fuel management surveys of CFR systems. The code calculates overall features such as reactivity, power distributions and breeding gain, and also calculates for each sub-assembly plutonium content and power output. A number of alternative options are built into the code, in order to permit different fuel management strategies to be calculated, and to perform more detailed calculations when necessary. A brief description is given of the methods of calculation, and the input facilities of CRACKLE, with examples. (author)

  6. Establishment of computer code system for nuclear reactor design - analysis

    International Nuclear Information System (INIS)

    Subki, I.R.; Santoso, B.; Syaukat, A.; Lee, S.M.

    1996-01-01

    Establishment of computer code system for nuclear reactor design analysis is given in this paper. This establishment is an effort to provide the capability in running various codes from nuclear data to reactor design and promote the capability for nuclear reactor design analysis particularly from neutronics and safety points. This establishment is also an effort to enhance the coordination of nuclear codes application and development existing in various research centre in Indonesia. Very prospective results have been obtained with the help of IAEA technical assistance. (author). 6 refs, 1 fig., 1 tab

  7. Quality assurance aspects of the computer code CODAR2

    International Nuclear Information System (INIS)

    Maul, P.R.

    1986-03-01

    The computer code CODAR2 was developed originally for use in connection with the Sizewell Public Inquiry to evaluate the radiological impact of routine discharges to the sea from the proposed PWR. It has subsequently bee used to evaluate discharges from Heysham 2. The code was frozen in September 1983, and this note gives details of its verification, validation and evaluation. Areas where either improved modelling methods or more up-to-date information relevant to CODAR2 data bases have subsequently become available are indicated; these will be incorporated in any future versions of the code. (author)

  8. Compendium of computer codes for the safety analysis of LMFBR's

    International Nuclear Information System (INIS)

    1975-06-01

    A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

  9. Independent peer review of nuclear safety computer codes

    International Nuclear Information System (INIS)

    Boyack, B.E.; Jenks, R.P.

    1993-01-01

    A structured, independent computer code peer-review process has been developed to assist the US Nuclear Regulatory Commission (NRC) and the US Department of Energy in their nuclear safety missions. This paper describes a structured process of independent code peer review, benefits associated with a code-independent peer review, as well as the authors' recent peer-review experience. The NRC adheres to the principle that safety of plant design, construction, and operation are the responsibility of the licensee. Nevertheless, NRC staff must have the ability to independently assess plant designs and safety analyses submitted by license applicants. According to Ref. 1, open-quotes this requires that a sound understanding be obtained of the important physical phenomena that may occur during transients in operating power plants.close quotes The NRC concluded that computer codes are the principal products to open-quotes understand and predict plant response to deviations from normal operating conditionsclose quotes and has developed several codes for that purpose. However, codes cannot be used blindly; they must be assessed and found adequate for the purposes they are intended. A key part of the qualification process can be accomplished through code peer reviews; this approach has been adopted by the NRC

  10. User's manual for the NEFTRAN II computer code

    International Nuclear Information System (INIS)

    Olague, N.E.; Campbell, J.E.; Leigh, C.D.; Longsine, D.E.

    1991-02-01

    This document describes the NEFTRAN II (NEtwork Flow and TRANsport in Time-Dependent Velocity Fields) computer code and is intended to provide the reader with sufficient information to use the code. NEFTRAN II was developed as part of a performance assessment methodology for storage of high-level nuclear waste in unsaturated, welded tuff. NEFTRAN II is a successor to the NEFTRAN and NWFT/DVM computer codes and contains several new capabilities. These capabilities include: (1) the ability to input pore velocities directly to the transport model and bypass the network fluid flow model, (2) the ability to transport radionuclides in time-dependent velocity fields, (3) the ability to account for the effect of time-dependent saturation changes on the retardation factor, and (4) the ability to account for time-dependent flow rates through the source regime. In addition to these changes, the input to NEFTRAN II has been modified to be more convenient for the user. This document is divided into four main sections consisting of (1) a description of all the models contained in the code, (2) a description of the program and subprograms in the code, (3) a data input guide and (4) verification and sample problems. Although NEFTRAN II is the fourth generation code, this document is a complete description of the code and reference to past user's manuals should not be necessary. 19 refs., 33 figs., 25 tabs

  11. Radiological impact assessment in Malaysia using RESRAD computer code

    International Nuclear Information System (INIS)

    Syed Hakimi Sakuma Syed Ahmad; Khairuddin Mohamad Kontol; Razali Hamzah

    1999-01-01

    Radiological Impact Assessment (RIA) can be conducted in Malaysia by using the RESRAD computer code developed by Argonne National Laboratory, U.S.A. The code can do analysis to derive site specific guidelines for allowable residual concentrations of radionuclides in soil. Concepts of the RIA in the context of waste management concern in Malaysia, some regulatory information and assess status of data collection are shown. Appropriate use scenarios and site specific parameters are used as much as possible so as to be realistic so that will reasonably ensure that individual dose limits and or constraints will be achieved. Case study have been conducted to fulfil Atomic Energy Licensing Board (AELB) requirements where for disposal purpose the operator must be required to carry out. a radiological impact assessment to all proposed disposals. This is to demonstrate that no member of public will be exposed to more than 1 mSv/year from all activities. Results obtained from analyses show the RESRAD computer code is able to calculate doses, risks, and guideline values. Sensitivity analysis by the computer code shows that the parameters used as input are justified so as to improve confidence to the public and the AELB the results of the analysis. The computer code can also be used as an initial assessment to conduct screening assessment in order to determine a proper disposal site. (Author)

  12. Abstracts of digital computer code packages. Assembled by the Radiation Shielding Information Center. [Radiation transport codes

    Energy Technology Data Exchange (ETDEWEB)

    McGill, B.; Maskewitz, B.F.; Anthony, C.M.; Comolander, H.E.; Hendrickson, H.R.

    1976-01-01

    The term ''code package'' is used to describe a miscellaneous grouping of materials which, when interpreted in connection with a digital computer, enables the scientist--user to solve technical problems in the area for which the material was designed. In general, a ''code package'' consists of written material--reports, instructions, flow charts, listings of data, and other useful material and IBM card decks (or, more often, a reel of magnetic tape) on which the source decks, sample problem input (including libraries of data) and the BCD/EBCDIC output listing from the sample problem are written. In addition to the main code, and any available auxiliary routines are also included. The abstract format was chosen to give to a potential code user several criteria for deciding whether or not he wishes to request the code package. (RWR)

  13. SURE: a system of computer codes for performing sensitivity/uncertainty analyses with the RELAP code

    International Nuclear Information System (INIS)

    Bjerke, M.A.

    1983-02-01

    A package of computer codes has been developed to perform a nonlinear uncertainty analysis on transient thermal-hydraulic systems which are modeled with the RELAP computer code. Using an uncertainty around the analyses of experiments in the PWR-BDHT Separate Effects Program at Oak Ridge National Laboratory. The use of FORTRAN programs running interactively on the PDP-10 computer has made the system very easy to use and provided great flexibility in the choice of processing paths. Several experiments simulating a loss-of-coolant accident in a nuclear reactor have been successfully analyzed. It has been shown that the system can be automated easily to further simplify its use and that the conversion of the entire system to a base code other than RELAP is possible

  14. SWIMS: a small-angle multiple scattering computer code

    International Nuclear Information System (INIS)

    Sayer, R.O.

    1976-07-01

    SWIMS (Sigmund and WInterbon Multiple Scattering) is a computer code for calculation of the angular dispersion of ion beams that undergo small-angle, incoherent multiple scattering by gaseous or solid media. The code uses the tabulated angular distributions of Sigmund and Winterbon for a Thomas-Fermi screened Coulomb potential. The fraction of the incident beam scattered into a cone defined by the polar angle α is computed as a function of α for reduced thicknesses over the range 0.01 less than or equal to tau less than or equal to 10.0. 1 figure, 2 tables

  15. The FOCON96 1.0 computer code

    International Nuclear Information System (INIS)

    Merle-Szeremeta, A.; Thomassin, A.

    1999-01-01

    The Institute of Protection and Nuclear Safety (I.P.S.N.) has developed a computer code, FOCON96 1.0 to calculate the dosimetric consequences of atmospheric radioactive releases from nuclear installations after several years of usual operation. This communication describes the principal characteristics of FOCON96 1.0 and its functionalities. The principal elements of a comparison between FOCON96 1.0 and PC-CREAM ( European computer code developed by the N.R.P.B. and answering the same criteria) are given here. (N.C.)

  16. Computer code MLCOSP for multiple-correlation and spectrum analysis with a hybrid computer

    International Nuclear Information System (INIS)

    Oguma, Ritsuo; Fujii, Yoshio; Usui, Hozumi; Watanabe, Koichi

    1975-10-01

    Usage of the computer code MLCOSP(Multiple Correlation and Spectrum) developed is described for a hybrid computer installed in JAERI Functions of the hybrid computer and its terminal devices are utilized ingeniously in the code to reduce complexity of the data handling which occurrs in analysis of the multivariable experimental data and to perform the analysis in perspective. Features of the code are as follows; Experimental data can be fed to the digital computer through the analog part of the hybrid computer by connecting with a data recorder. The computed results are displayed in figures, and hardcopies are taken when necessary. Series-messages to the code are shown on the terminal, so man-machine communication is possible. And further the data can be put in through a keyboard, so case study according to the results of analysis is possible. (auth.)

  17. High performance computer code for molecular dynamics simulations

    International Nuclear Information System (INIS)

    Levay, I.; Toekesi, K.

    2007-01-01

    Complete text of publication follows. Molecular Dynamics (MD) simulation is a widely used technique for modeling complicated physical phenomena. Since 2005 we are developing a MD simulations code for PC computers. The computer code is written in C++ object oriented programming language. The aim of our work is twofold: a) to develop a fast computer code for the study of random walk of guest atoms in Be crystal, b) 3 dimensional (3D) visualization of the particles motion. In this case we mimic the motion of the guest atoms in the crystal (diffusion-type motion), and the motion of atoms in the crystallattice (crystal deformation). Nowadays, it is common to use Graphics Devices in intensive computational problems. There are several ways to use this extreme processing performance, but never before was so easy to programming these devices as now. The CUDA (Compute Unified Device) Architecture introduced by nVidia Corporation in 2007 is a very useful for every processor hungry application. A Unified-architecture GPU include 96-128, or more stream processors, so the raw calculation performance is 576(!) GFLOPS. It is ten times faster, than the fastest dual Core CPU [Fig.1]. Our improved MD simulation software uses this new technology, which speed up our software and the code run 10 times faster in the critical calculation code segment. Although the GPU is a very powerful tool, it has a strongly paralleled structure. It means, that we have to create an algorithm, which works on several processors without deadlock. Our code currently uses 256 threads, shared and constant on-chip memory, instead of global memory, which is 100 times slower than others. It is possible to implement the total algorithm on GPU, therefore we do not need to download and upload the data in every iteration. On behalf of maximal throughput, every thread run with the same instructions

  18. Prodeto, a computer code for probabilistic fatigue design

    Energy Technology Data Exchange (ETDEWEB)

    Braam, H [ECN-Solar and Wind Energy, Petten (Netherlands); Christensen, C J; Thoegersen, M L [Risoe National Lab., Roskilde (Denmark); Ronold, K O [Det Norske Veritas, Hoevik (Norway)

    1999-03-01

    A computer code for structural relibility analyses of wind turbine rotor blades subjected to fatigue loading is presented. With pre-processors that can transform measured and theoretically predicted load series to load range distributions by rain-flow counting and with a family of generic distribution models for parametric representation of these distribution this computer program is available for carying through probabilistic fatigue analyses of rotor blades. (au)

  19. 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

  20. Computer simulation of variform fuel assemblies using Dragon code

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun; Yao Dong

    2005-01-01

    The DRAGON is a cell code that developed for the CANDU reactor by the Ecole Polytechnique de Montreal of CANADA. Although, the DRAGON is mainly used to simulate the CANDU super-cell fuel assembly, it has an ability to simulate other geometries of the fuel assembly. However, only NEACRP benchmark problem of the BWR lattice cell was analyzed until now except for the CANDU reactor. We also need to develop the code to simulate the variform fuel assemblies, especially, for design of the advanced reactor. We validated that the cell code DRAGON is useful for simulating various kinds of the fuel assembly by analyzing the rod-shape fuel assembly of the PWR and the MTR plate-shape fuel assembly. Some other kinds of geometry of geometry were computed. Computational results show that the DRAGON is able to analyze variform fuel assembly problems and the precision is high. (authors)

  1. Development of Reference Data Set (RDS) for LOBI-MOD2 Integral Test Facility- IAEA Fellowship Training at Nuclear Research Group of San Piero A Grado (GRNSPG), University of PISA, Italy

    International Nuclear Information System (INIS)

    Mohd Rizal Mamat

    2013-01-01

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process in order to ensure compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. In order to ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardised and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa which describes the whole processes or steps involved in the preparation of complete database for system thermal-hydraulic code applications for facilities or plants. Under this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization for system thermal-hydraulics code simulation has been proposed. This paper describes the experience of having undergone 2 months of IAEA Fellowship training at Nuclear Research Group of San Piero A Grado (GRNSPG) in University of PISA, Italy and the application of RDS and its effectiveness. Two RDS documents have been developed for an Integral Test Facility of LOBI-MOD2 facility and Test A1-83, 10% small cold leg break LOCA (Loss of Coolant Accident). (author)

  2. Plagiarism Detection Algorithm for Source Code in Computer Science Education

    Science.gov (United States)

    Liu, Xin; Xu, Chan; Ouyang, Boyu

    2015-01-01

    Nowadays, computer programming is getting more necessary in the course of program design in college education. However, the trick of plagiarizing plus a little modification exists among some students' home works. It's not easy for teachers to judge if there's plagiarizing in source code or not. Traditional detection algorithms cannot fit this…

  3. Protect Heterogeneous Environment Distributed Computing from Malicious Code Assignment

    Directory of Open Access Journals (Sweden)

    V. S. Gorbatov

    2011-09-01

    Full Text Available The paper describes the practical implementation of the protection system of heterogeneous environment distributed computing from malicious code for the assignment. A choice of technologies, development of data structures, performance evaluation of the implemented system security are conducted.

  4. Atmospheric dispersion of radioactive releases: Computer code DIASPORA

    International Nuclear Information System (INIS)

    Synodinou, B.M.; Bartzis, J.M.

    1982-05-01

    The computer code DIASPORA is presented. Air and ground concentrations of an airborne radioactive material released from an elevated continuous point source are calculated using Gaussian plume models. Dry and wet deposition as well as plume rise effects are taken into consideration. (author)

  5. 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

  6. Computer code for double beta decay QRPA based calculations

    Energy Technology Data Exchange (ETDEWEB)

    Barbero, C. A.; Mariano, A. [Departamento de Física, Facultad de Ciencias Exactas, Universidad Nacional de La Plata, La Plata, Argentina and Instituto de Física La Plata, CONICET, La Plata (Argentina); Krmpotić, F. [Instituto de Física La Plata, CONICET, La Plata, Argentina and Instituto de Física Teórica, Universidade Estadual Paulista, São Paulo (Brazil); Samana, A. R.; Ferreira, V. dos Santos [Departamento de Ciências Exatas e Tecnológicas, Universidade Estadual de Santa Cruz, BA (Brazil); Bertulani, C. A. [Department of Physics, Texas A and M University-Commerce, Commerce, TX (United States)

    2014-11-11

    The computer code developed by our group some years ago for the evaluation of nuclear matrix elements, within the QRPA and PQRPA nuclear structure models, involved in neutrino-nucleus reactions, muon capture and β{sup ±} processes, is extended to include also the nuclear double beta decay.

  7. Connecting Neural Coding to Number Cognition: A Computational Account

    Science.gov (United States)

    Prather, Richard W.

    2012-01-01

    The current study presents a series of computational simulations that demonstrate how the neural coding of numerical magnitude may influence number cognition and development. This includes behavioral phenomena cataloged in cognitive literature such as the development of numerical estimation and operational momentum. Though neural research has…

  8. Linking CATHENA with other computer codes through a remote process

    Energy Technology Data Exchange (ETDEWEB)

    Vasic, A.; Hanna, B.N.; Waddington, G.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Girard, R. [Hydro-Quebec, Montreal, Quebec (Canada)

    2005-07-01

    'Full text:' CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a computer code developed by Atomic Energy of Canada Limited (AECL). The code uses a transient, one-dimensional, two-fluid representation of two-phase flow in piping networks. CATHENA is used primarily for the analysis of postulated upset conditions in CANDU reactors; however, the code has found a wider range of applications. In the past, the CATHENA thermalhydraulics code included other specialized codes, i.e. ELOCA and the Point LEPreau CONtrol system (LEPCON) as callable subroutine libraries. The combined program was compiled and linked as a separately named code. This code organizational process is not suitable for independent development, maintenance, validation and version tracking of separate computer codes. The alternative solution to provide code development independence is to link CATHENA to other computer codes through a Parallel Virtual Machine (PVM) interface process. PVM is a public domain software package, developed by Oak Ridge National Laboratory and enables a heterogeneous collection of computers connected by a network to be used as a single large parallel machine. The PVM approach has been well accepted by the global computing community and has been used successfully for solving large-scale problems in science, industry, and business. Once development of the appropriate interface for linking independent codes through PVM is completed, future versions of component codes can be developed, distributed separately and coupled as needed by the user. This paper describes the coupling of CATHENA to the ELOCA-IST and the TROLG2 codes through a PVM remote process as an illustration of possible code connections. ELOCA (Element Loss Of Cooling Analysis) is the Industry Standard Toolset (IST) code developed by AECL to simulate the thermo-mechanical response of CANDU fuel elements to transient thermalhydraulics boundary conditions. A separate ELOCA driver program

  9. Linking CATHENA with other computer codes through a remote process

    International Nuclear Information System (INIS)

    Vasic, A.; Hanna, B.N.; Waddington, G.M.; Sabourin, G.; Girard, R.

    2005-01-01

    'Full text:' CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a computer code developed by Atomic Energy of Canada Limited (AECL). The code uses a transient, one-dimensional, two-fluid representation of two-phase flow in piping networks. CATHENA is used primarily for the analysis of postulated upset conditions in CANDU reactors; however, the code has found a wider range of applications. In the past, the CATHENA thermalhydraulics code included other specialized codes, i.e. ELOCA and the Point LEPreau CONtrol system (LEPCON) as callable subroutine libraries. The combined program was compiled and linked as a separately named code. This code organizational process is not suitable for independent development, maintenance, validation and version tracking of separate computer codes. The alternative solution to provide code development independence is to link CATHENA to other computer codes through a Parallel Virtual Machine (PVM) interface process. PVM is a public domain software package, developed by Oak Ridge National Laboratory and enables a heterogeneous collection of computers connected by a network to be used as a single large parallel machine. The PVM approach has been well accepted by the global computing community and has been used successfully for solving large-scale problems in science, industry, and business. Once development of the appropriate interface for linking independent codes through PVM is completed, future versions of component codes can be developed, distributed separately and coupled as needed by the user. This paper describes the coupling of CATHENA to the ELOCA-IST and the TROLG2 codes through a PVM remote process as an illustration of possible code connections. ELOCA (Element Loss Of Cooling Analysis) is the Industry Standard Toolset (IST) code developed by AECL to simulate the thermo-mechanical response of CANDU fuel elements to transient thermalhydraulics boundary conditions. A separate ELOCA driver program starts, ends

  10. Methods and computer codes for probabilistic sensitivity and uncertainty analysis

    International Nuclear Information System (INIS)

    Vaurio, J.K.

    1985-01-01

    This paper describes the methods and applications experience with two computer codes that are now available from the National Energy Software Center at Argonne National Laboratory. The purpose of the SCREEN code is to identify a group of most important input variables of a code that has many (tens, hundreds) input variables with uncertainties, and do this without relying on judgment or exhaustive sensitivity studies. Purpose of the PROSA-2 code is to propagate uncertainties and calculate the distributions of interesting output variable(s) of a safety analysis code using response surface techniques, based on the same runs used for screening. Several applications are discussed, but the codes are generic, not tailored to any specific safety application code. They are compatible in terms of input/output requirements but also independent of each other, e.g., PROSA-2 can be used without first using SCREEN if a set of important input variables has first been selected by other methods. Also, although SCREEN can select cases to be run (by random sampling), a user can select cases by other methods if he so prefers, and still use the rest of SCREEN for identifying important input variables

  11. Development and application of computational aerothermodynamics flowfield computer codes

    Science.gov (United States)

    Venkatapathy, Ethiraj

    1993-01-01

    Computations are presented for one-dimensional, strong shock waves that are typical of those that form in front of a reentering spacecraft. The fluid mechanics and thermochemistry are modeled using two different approaches. The first employs traditional continuum techniques in solving the Navier-Stokes equations. The second-approach employs a particle simulation technique (the direct simulation Monte Carlo method, DSMC). The thermochemical models employed in these two techniques are quite different. The present investigation presents an evaluation of thermochemical models for nitrogen under hypersonic flow conditions. Four separate cases are considered. The cases are governed, respectively, by the following: vibrational relaxation; weak dissociation; strong dissociation; and weak ionization. In near-continuum, hypersonic flow, the nonequilibrium thermochemical models employed in continuum and particle simulations produce nearly identical solutions. Further, the two approaches are evaluated successfully against available experimental data for weakly and strongly dissociating flows.

  12. 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

  13. Computed radiography simulation using the Monte Carlo code MCNPX

    International Nuclear Information System (INIS)

    Correa, S.C.A.; Souza, E.M.; Silva, A.X.; Lopes, R.T.

    2009-01-01

    Simulating x-ray images has been of great interest in recent years as it makes possible an analysis of how x-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data. (author)

  14. Computed radiography simulation using the Monte Carlo code MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Correa, S.C.A. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Centro Universitario Estadual da Zona Oeste (CCMAT)/UEZO, Av. Manuel Caldeira de Alvarenga, 1203, Campo Grande, 23070-200, Rio de Janeiro, RJ (Brazil); Souza, E.M. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Silva, A.X., E-mail: ademir@con.ufrj.b [PEN/COPPE-DNC/Poli CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Cassiano, D.H. [Instituto de Radioprotecao e Dosimetria/CNEN Av. Salvador Allende, s/n, Recreio, 22780-160, Rio de Janeiro, RJ (Brazil); Lopes, R.T. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil)

    2010-09-15

    Simulating X-ray images has been of great interest in recent years as it makes possible an analysis of how X-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data.

  15. Numerical computation of molecular integrals via optimized (vectorized) FORTRAN code

    International Nuclear Information System (INIS)

    Scott, T.C.; Grant, I.P.; Saunders, V.R.

    1997-01-01

    The calculation of molecular properties based on quantum mechanics is an area of fundamental research whose horizons have always been determined by the power of state-of-the-art computers. A computational bottleneck is the numerical calculation of the required molecular integrals to sufficient precision. Herein, we present a method for the rapid numerical evaluation of molecular integrals using optimized FORTRAN code generated by Maple. The method is based on the exploitation of common intermediates and the optimization can be adjusted to both serial and vectorized computations. (orig.)

  16. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  17. Fuel rod computations. The COMETHE code in its CEA version

    International Nuclear Information System (INIS)

    Lenepveu, Dominique.

    1976-01-01

    The COMETHE code (COde d'evolution MEcanique et THermique) is intended for computing the irradiation behavior of water reactor fuel pins. It is concerned with steadily operated cylindrical pins, containing fuel pellet stacks (UO 2 or PuO 2 ). The pin consists in five different axial zones: two expansion chambers, two blankets, and a central core that may be divided into several stacks parted by plugs. As far as computation is concerned, the pin is divided into slices (maximum 15) in turn divided into rings (maximum 50). Information are obtained for each slice: the radial temperature distribution, heat transfer coefficients, thermal flux at the pin surface, changes in geometry according to temperature conditions, and specific burn-up. The physical models involved take account for: heat transfer, fission gas release, fuel expansion, and creep of the can. Results computed with COMETHE are compared with those from ELP and EPEL irradiation experiments [fr

  18. A DOE Computer Code Toolbox: Issues and Opportunities

    International Nuclear Information System (INIS)

    Vincent, A.M. III

    2001-01-01

    The initial activities of a Department of Energy (DOE) Safety Analysis Software Group to establish a Safety Analysis Toolbox of computer models are discussed. The toolbox shall be a DOE Complex repository of verified and validated computer models that are configuration-controlled and made available for specific accident analysis applications. The toolbox concept was recommended by the Defense Nuclear Facilities Safety Board staff as a mechanism to partially address Software Quality Assurance issues. Toolbox candidate codes have been identified through review of a DOE Survey of Software practices and processes, and through consideration of earlier findings of the Accident Phenomenology and Consequence Evaluation program sponsored by the DOE National Nuclear Security Agency/Office of Defense Programs. Planning is described to collect these high-use codes, apply tailored SQA specific to the individual codes, and implement the software toolbox concept. While issues exist such as resource allocation and the interface among code developers, code users, and toolbox maintainers, significant benefits can be achieved through a centralized toolbox and subsequent standardized applications

  19. Additional extensions to the NASCAP computer code, volume 3

    Science.gov (United States)

    Mandell, M. J.; Cooke, D. L.

    1981-01-01

    The ION computer code is designed to calculate charge exchange ion densities, electric potentials, plasma temperatures, and current densities external to a neutralized ion engine in R-Z geometry. The present version assumes the beam ion current and density to be known and specified, and the neutralizing electrons to originate from a hot-wire ring surrounding the beam orifice. The plasma is treated as being resistive, with an electron relaxation time comparable to the plasma frequency. Together with the thermal and electrical boundary conditions described below and other straightforward engine parameters, these assumptions suffice to determine the required quantities. The ION code, written in ASCII FORTRAN for UNIVAC 1100 series computers, is designed to be run interactively, although it can also be run in batch mode. The input is free-format, and the output is mainly graphical, using the machine-independent graphics developed for the NASCAP code. The executive routine calls the code's major subroutines in user-specified order, and the code allows great latitude for restart and parameter change.

  20. COMPUTATION FORMAT computer codes X4TOC4 and PLOTC4. Implementing and Testing on a Personal Computer

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1987-05-01

    This document describes the contents of the diskette containing the COMPUTATION FORMAT codes X4TOC4 and PLOTC4 by D.E. Cullen, and example data for use in implementing and testing these codes on a Personal Computer of the type IBM-PC/AT. Upon request the codes are available from the IAEA Nuclear Data Section, free of charge, on a single diskette. (author)

  1. Benchmarking severe accident computer codes for heavy water reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J.H. [International Atomic Energy Agency, Vienna (Austria)

    2010-07-01

    Consideration of severe accidents at a nuclear power plant (NPP) is an essential component of the defence in depth approach used in nuclear safety. Severe accident analysis involves 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. International cooperative research programmes are established by the IAEA in areas that are of common interest to a number of Member States. These co-operative efforts are carried out through coordinated research projects (CRPs), typically 3 to 6 years in duration, and often involving experimental activities. Such CRPs allow a sharing of efforts on an international basis, foster team-building and benefit from the experience and expertise of researchers from all participating institutes. The IAEA is organizing a CRP on benchmarking severe accident computer codes for heavy water reactor (HWR) applications. The CRP scope includes defining the severe accident sequence and conducting benchmark analyses for HWRs, evaluating the capabilities of existing computer codes to predict important severe accident phenomena, and suggesting necessary code improvements and/or new experiments to reduce uncertainties. The CRP has been planned on the advice and with the support of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for HWRs. (author)

  2. LMFBR models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.

    1981-10-01

    Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th- 238 U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given

  3. Integrated computer codes for nuclear power plant severe accident analysis

    International Nuclear Information System (INIS)

    Jordanov, I.; Khristov, Y.

    1995-01-01

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs

  4. Integrated computer codes for nuclear power plant severe accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jordanov, I; Khristov, Y [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs.

  5. A computer code for Tokamak reactor concepts evaluation

    International Nuclear Information System (INIS)

    Rosatelli, F.; Raia, G.

    1985-01-01

    A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

  6. RADTRAN 5: A computer code for transportation risk analysis

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Kanipe, F.L.

    1991-01-01

    RADTRAN 5 is a computer code developed at Sandia National Laboratories (SNL) in Albuquerque, NM, to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI Standard FORTRAN 77 and contains significant advances in the methodology for route-specific analysis first developed by SNL for RADTRAN 4 (Neuhauser and Kanipe, 1992). Like the previous RADTRAN codes, RADTRAN 5 contains two major modules for incident-free and accident risk amlysis, respectively. All commercially important transportation modes may be analyzed with RADTRAN 5: highway by combination truck; highway by light-duty vehicle; rail; barge; ocean-going ship; cargo air; and passenger air

  7. A computer code for fault tree calculations: PATREC

    International Nuclear Information System (INIS)

    Blin, A.; Carnino, A.; Koen, B.V.; Duchemin, B.; Lanore, J.M.; Kalli, H.

    1978-01-01

    A computer code for evaluating the reliability of complex system by fault tree is described in this paper. It uses pattern recognition approach and programming techniques from IBM PL1 language. It can take account of many of the present day problems: multi-dependencies treatment, dispersion in the reliability data parameters, influence of common mode failures. The code is running currently since two years now in Commissariat a l'Energie Atomique Saclay center and shall be used in a future extension for automatic fault trees construction

  8. An improved thermal model for the computer code NAIAD

    International Nuclear Information System (INIS)

    Rainbow, M.T.

    1982-12-01

    An improved thermal model, based on the concept of heat slabs, has been incorporated as an option into the thermal hydraulic computer code NAIAD. The heat slabs are one-dimensional thermal conduction models with temperature independent thermal properties which may be internal and/or external to the fluid. Thermal energy may be added to or removed from the fluid via heat slabs and passed across the external boundary of external heat slabs at a rate which is a linear function of the external surface temperatures. The code input for the new option has been restructured to simplify data preparation. A full description of current input requirements is presented

  9. Integrated severe accident containment analysis with the CONTAIN computer code

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.

    1985-12-01

    Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite

  10. War of ontology worlds: mathematics, computer code, or Esperanto?

    Science.gov (United States)

    Rzhetsky, Andrey; Evans, James A

    2011-09-01

    The use of structured knowledge representations-ontologies and terminologies-has become standard in biomedicine. Definitions of ontologies vary widely, as do the values and philosophies that underlie them. In seeking to make these views explicit, we conducted and summarized interviews with a dozen leading ontologists. Their views clustered into three broad perspectives that we summarize as mathematics, computer code, and Esperanto. Ontology as mathematics puts the ultimate premium on rigor and logic, symmetry and consistency of representation across scientific subfields, and the inclusion of only established, non-contradictory knowledge. Ontology as computer code focuses on utility and cultivates diversity, fitting ontologies to their purpose. Like computer languages C++, Prolog, and HTML, the code perspective holds that diverse applications warrant custom designed ontologies. Ontology as Esperanto focuses on facilitating cross-disciplinary communication, knowledge cross-referencing, and computation across datasets from diverse communities. We show how these views align with classical divides in science and suggest how a synthesis of their concerns could strengthen the next generation of biomedical ontologies.

  11. Computer codes for evaluation of control room habitability (HABIT)

    International Nuclear Information System (INIS)

    Stage, S.A.

    1996-06-01

    This report describes the Computer Codes for Evaluation of Control Room Habitability (HABIT). HABIT is a package of computer codes designed to be used for the evaluation of control room habitability in the event of an accidental release of toxic chemicals or radioactive materials. Given information about the design of a nuclear power plant, a scenario for the release of toxic chemicals or radionuclides, and information about the air flows and protection systems of the control room, HABIT can be used to estimate the chemical exposure or radiological dose to control room personnel. HABIT is an integrated package of several programs that previously needed to be run separately and required considerable user intervention. This report discusses the theoretical basis and physical assumptions made by each of the modules in HABIT and gives detailed information about the data entry windows. Sample runs are given for each of the modules. A brief section of programming notes is included. A set of computer disks will accompany this report if the report is ordered from the Energy Science and Technology Software Center. The disks contain the files needed to run HABIT on a personal computer running DOS. Source codes for the various HABIT routines are on the disks. Also included are input and output files for three demonstration runs

  12. Theoretical Atomic Physics code development IV: LINES, A code for computing atomic line spectra

    International Nuclear Information System (INIS)

    Abdallah, J. Jr.; Clark, R.E.H.

    1988-12-01

    A new computer program, LINES, has been developed for simulating atomic line emission and absorption spectra using the accurate fine structure energy levels and transition strengths calculated by the (CATS) Cowan Atomic Structure code. Population distributions for the ion stages are obtained in LINES by using the Local Thermodynamic Equilibrium (LTE) model. LINES is also useful for displaying the pertinent atomic data generated by CATS. This report describes the use of LINES. Both CATS and LINES are part of the Theoretical Atomic PhysicS (TAPS) code development effort at Los Alamos. 11 refs., 9 figs., 1 tab

  13. Experience with the WIMS computer code at Skoda Plzen

    International Nuclear Information System (INIS)

    Vacek, J.; Mikolas, P.

    1991-01-01

    Validation of the program for neutronics analysis is described. Computational results are compared with results of experiments on critical assemblies and with results of other codes for different types of lattices. Included are the results for lattices containing Gd as burnable absorber. With minor exceptions, the results of benchmarking were quite satisfactory and justified the inclusion of WIMS in the production system of codes for WWER analysis. The first practical application was the adjustment of the WWER-440 few-group diffusion constants library of the three-dimensional diffusion code MOBY-DICK, which led to a remarkable improvement of results for operational states. Then a new library for the analysis of WWER-440 start-up was generated and tested and at present a new library for the analysis of WWER-440 operational states is being tested. Preparation of the library for WWER-1000 is in progress. (author). 19 refs

  14. Benchmarking of computer codes and approaches for modeling exposure scenarios

    International Nuclear Information System (INIS)

    Seitz, R.R.; Rittmann, P.D.; Wood, M.I.; Cook, J.R.

    1994-08-01

    The US Department of Energy Headquarters established a performance assessment task team (PATT) to integrate the activities of DOE sites that are preparing performance assessments for the disposal of newly generated low-level waste. The PATT chartered a subteam with the task of comparing computer codes and exposure scenarios used for dose calculations in performance assessments. This report documents the efforts of the subteam. Computer codes considered in the comparison include GENII, PATHRAE-EPA, MICROSHIELD, and ISOSHLD. Calculations were also conducted using spreadsheets to provide a comparison at the most fundamental level. Calculations and modeling approaches are compared for unit radionuclide concentrations in water and soil for the ingestion, inhalation, and external dose pathways. Over 30 tables comparing inputs and results are provided

  15. Microdosimetry computation code of internal sources - MICRODOSE 1

    International Nuclear Information System (INIS)

    Li Weibo; Zheng Wenzhong; Ye Changqing

    1995-01-01

    This paper describes a microdosimetry computation code, MICRODOSE 1, on the basis of the following described methods: (1) the method of calculating f 1 (z) for charged particle in the unit density tissues; (2) the method of calculating f(z) for a point source; (3) the method of applying the Fourier transform theory to the calculation of the compound Poisson process; (4) the method of using fast Fourier transform technique to determine f(z) and, giving some computed examples based on the code, MICRODOSE 1, including alpha particles emitted from 239 Pu in the alveolar lung tissues and from radon progeny RaA and RAC in the human respiratory tract. (author). 13 refs., 6 figs

  16. Vessel coolant mass depletion during a 5% SBLOCA in the Semiscale Mod-2C facility

    International Nuclear Information System (INIS)

    Shaw, R.A.; Loomis, G.G.

    1985-01-01

    Experimental results are presented from two 5% small-break loss-of-coolant accident (SBLOCA) simulations in the Semiscale Mod-2C facility. In performing the simulated 5% SBLOCAs, boundary conditions scaled from a pressurized water reactor (PWR) were used. The experiment was run with initial conditions typical of a PWR (15.6 MPa pressure and 35 K core differential temperature). The Mod-2C facility represents the state-of-the-art in small facilities scaled from PWRs. Phenomena which occurred during the transient included: primary fluid saturation (change from subcooled to saturated blowdown), break uncovery (a centerline break was simulated), condensation-induced liquid hold-up in the steam generator primary tubes, pump suction liquid seal formation and core level depression with resulting core rod temperature excursion, pump suction liquid seal clearance, loop fluid mass redistribution, and gradual core rewet. The influence of core bypass flow is also discussed. 11 refs., 13 figs

  17. Validation and testing of the VAM2D computer code

    International Nuclear Information System (INIS)

    Kool, J.B.; Wu, Y.S.

    1991-10-01

    This document describes two modeling studies conducted by HydroGeoLogic, Inc. for the US NRC under contract no. NRC-04089-090, entitled, ''Validation and Testing of the VAM2D Computer Code.'' VAM2D is a two-dimensional, variably saturated flow and transport code, with applications for performance assessment of nuclear waste disposal. The computer code itself is documented in a separate NUREG document (NUREG/CR-5352, 1989). The studies presented in this report involve application of the VAM2D code to two diverse subsurface modeling problems. The first one involves modeling of infiltration and redistribution of water and solutes in an initially dry, heterogeneous field soil. This application involves detailed modeling over a relatively short, 9-month time period. The second problem pertains to the application of VAM2D to the modeling of a waste disposal facility in a fractured clay, over much larger space and time scales and with particular emphasis on the applicability and reliability of using equivalent porous medium approach for simulating flow and transport in fractured geologic media. Reflecting the separate and distinct nature of the two problems studied, this report is organized in two separate parts. 61 refs., 31 figs., 9 tabs

  18. FRANTIC: a computer code for time dependent unavailability analysis

    International Nuclear Information System (INIS)

    Vesely, W.E.; Goldberg, F.F.

    1977-03-01

    The FRANTIC computer code evaluates the time dependent and average unavailability for any general system model. The code is written in FORTRAN IV for the IBM 370 computer. Non-repairable components, monitored components, and periodically tested components are handled. One unique feature of FRANTIC is the detailed, time dependent modeling of periodic testing which includes the effects of test downtimes, test overrides, detection inefficiencies, and test-caused failures. The exponential distribution is used for the component failure times and periodic equations are developed for the testing and repair contributions. Human errors and common mode failures can be included by assigning an appropriate constant probability for the contributors. The output from FRANTIC consists of tables and plots of the system unavailability along with a breakdown of the unavailability contributions. Sensitivity studies can be simply performed and a wide range of tables and plots can be obtained for reporting purposes. The FRANTIC code represents a first step in the development of an approach that can be of direct value in future system evaluations. Modifications resulting from use of the code, along with the development of reliability data based on operating reactor experience, can be expected to provide increased confidence in its use and potential application to the licensing process

  19. User's manual for computer code RIBD-II, a fission product inventory code

    International Nuclear Information System (INIS)

    Marr, D.R.

    1975-01-01

    The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)

  20. Available computer codes and data for radiation transport analysis

    International Nuclear Information System (INIS)

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

    1975-01-01

    The Radiation Shielding Information Center (RSIC), sponsored and supported by the Energy Research and Development Administration (ERDA) and the Defense Nuclear Agency (DNA), is a technical institute serving the radiation transport and shielding community. It acquires, selects, stores, retrieves, evaluates, analyzes, synthesizes, and disseminates information on shielding and ionizing radiation transport. The major activities include: (1) operating a computer-based information system and answering inquiries on radiation analysis, (2) collecting, checking out, packaging, and distributing large computer codes, and evaluated and processed data libraries. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

  1. Computer code for qualitative analysis of gamma-ray spectra

    International Nuclear Information System (INIS)

    Yule, H.P.

    1979-01-01

    Computer code QLN1 provides complete analysis of gamma-ray spectra observed with Ge(Li) detectors and is used at both the National Bureau of Standards and the Environmental Protection Agency. It locates peaks, resolves multiplets, identifies component radioisotopes, and computes quantitative results. The qualitative-analysis (or component identification) algorithms feature thorough, self-correcting steps which provide accurate isotope identification in spite of errors in peak centroids, energy calibration, and other typical problems. The qualitative-analysis algorithm is described in this paper

  2. Verification of structural analysis computer codes in nuclear engineering

    International Nuclear Information System (INIS)

    Zebeljan, Dj.; Cizelj, L.

    1990-01-01

    Sources of potential errors, which can take place during use of finite element method based computer programs, are described in the paper. The magnitude of errors was defined as acceptance criteria for those programs. Error sources are described as they are treated by 'National Agency for Finite Element Methods and Standards (NAFEMS)'. Specific verification examples are used from literature of Nuclear Regulatory Commission (NRC). Example of verification is made on PAFEC-FE computer code for seismic response analyses of piping systems by response spectrum method. (author)

  3. A computer code package for electron transport Monte Carlo simulation

    International Nuclear Information System (INIS)

    Popescu, Lucretiu M.

    1999-01-01

    A computer code package was developed for solving various electron transport problems by Monte Carlo simulation. It is based on condensed history Monte Carlo algorithm. In order to get reliable results over wide ranges of electron energies and target atomic numbers, specific techniques of electron transport were implemented such as: Moliere multiscatter angular distributions, Blunck-Leisegang multiscatter energy distribution, sampling of electron-electron and Bremsstrahlung individual interactions. Path-length and lateral displacement corrections algorithms and the module for computing collision, radiative and total restricted stopping powers and ranges of electrons are also included. Comparisons of simulation results with experimental measurements are finally presented. (author)

  4. Methods for the development of large computer codes under LTSS

    International Nuclear Information System (INIS)

    Sicilian, J.M.

    1977-06-01

    TRAC is a large computer code being developed by Group Q-6 for the analysis of the transient thermal hydraulic behavior of light-water nuclear reactors. A system designed to assist the development of TRAC is described. The system consists of a central HYDRA dataset, R6LIB, containing files used in the development of TRAC, and a file maintenance program, HORSE, which facilitates the use of this dataset

  5. WAMCUT, a computer code for fault tree evaluation. Final report

    International Nuclear Information System (INIS)

    Erdmann, R.C.

    1978-06-01

    WAMCUT is a code in the WAM family which produces the minimum cut sets (MCS) for a given fault tree. The MCS are useful as they provide a qualitative evaluation of a system, as well as providing a means of determining the probability distribution function for the top of the tree. The program is very efficient and will produce all the MCS in a very short computer time span. 22 figures, 4 tables

  6. Parallel computing by Monte Carlo codes MVP/GMVP

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Nakagawa, Masayuki; Mori, Takamasa

    2001-01-01

    General-purpose Monte Carlo codes MVP/GMVP are well-vectorized and thus enable us to perform high-speed Monte Carlo calculations. In order to achieve more speedups, we parallelized the codes on the different types of parallel computing platforms or by using a standard parallelization library MPI. The platforms used for benchmark calculations are a distributed-memory vector-parallel computer Fujitsu VPP500, a distributed-memory massively parallel computer Intel paragon and a distributed-memory scalar-parallel computer Hitachi SR2201, IBM SP2. As mentioned generally, linear speedup could be obtained for large-scale problems but parallelization efficiency decreased as the batch size per a processing element(PE) was smaller. It was also found that the statistical uncertainty for assembly powers was less than 0.1% by the PWR full-core calculation with more than 10 million histories and it took about 1.5 hours by massively parallel computing. (author)

  7. Computer codes for the analysis of flask impact problems

    International Nuclear Information System (INIS)

    Neilson, A.J.

    1984-09-01

    This review identifies typical features of the design of transportation flasks and considers some of the analytical tools required for the analysis of impact events. Because of the complexity of the physical problem, it is unlikely that a single code will adequately deal with all the aspects of the impact incident. Candidate codes are identified on the basis of current understanding of their strengths and limitations. It is concluded that the HONDO-II, DYNA3D AND ABAQUS codes which ar already mounted on UKAEA computers will be suitable tools for use in the analysis of experiments conducted in the proposed AEEW programme and of general flask impact problems. Initial attention should be directed at the DYNA3D and ABAQUS codes with HONDO-II being reserved for situations where the three-dimensional elements of DYNA3D may provide uneconomic simulations in planar or axisymmetric geometries. Attention is drawn to the importance of access to suitable mesh generators to create the nodal coordinate and element topology data required by these structural analysis codes. (author)

  8. Computer codes for the operational control of the research reactors

    International Nuclear Information System (INIS)

    Kalker, K.J.; Nabbi, R.; Bormann, H.J.

    1986-01-01

    Four small computer codes developed by ZFR are presented, which have been used for several years during operation of the research reactors FRJ-1, FRJ-2, AVR (all in Juelich) and DR-2 (Riso, Denmark). Because of interest coming from the other reactor stations the codes are documented within the frame work of the IAEA Research Contract No. 3634/FG. The zero-dimensional burnup program CREMAT is used for reactor cores in which flux measurements at each individual fuel element are carried out during operation. The program yields burnup data for each fuel element and for the whole core. On the basis of these data, fuel reloading is prepared for the next operational period under consideration of the permitted minimum shut down reactivity of the system. The program BURNY calculates burnup for fuel elements inaccessible for flux measurements, but for which 'position weighting factors' have been measured/calculated during zero power operation of the core, and which are assumed to be constant in all operational situations. The code CURIAX calculates post-irradiation data for discharged fuel elements needed in their manipulation and transport. These three programs have been written for highly enriched fuel and take into account U-235 only. The modification of CREMAT for LEU Cores and its combiantion with ORIGEN is in preparation. KINIK is an inverse kinetic code and widely used for absorber rod calibration at the abovementioned research reactors. It includes a special polynomial subroutine which can easily be used in other codes. (orig.) [de

  9. Use of computer codes to improve nuclear power plant operation

    International Nuclear Information System (INIS)

    Misak, J.; Polak, V.; Filo, J.; Gatas, J.

    1985-01-01

    For safety and economic reasons, the scope for carrying out experiments on operational nuclear power plants (NPPs) is very limited and any changes in technical equipment and operating parameters or conditions have to be supported by theoretical calculations. In the Nuclear Power Plant Scientific Research Institute (NIIAEhS), computer codes are systematically used to analyse actual operating events, assess safety aspects of changes in equipment and operating conditions, optimize the conditions, preparation and analysis of NPP startup trials and review and amend operating instructions. In addition, calculation codes are gradually being introduced into power plant computer systems to perform real time processing of the parameters being measured. The paper describes a number of specific examples of the use of calculation codes for the thermohydraulic analysis of operating and accident conditions aimed at improving the operation of WWER-440 units at the Jaslovske Bohunice V-1 and V-2 nuclear power plants. These examples confirm that computer calculations are an effective way of solving operating problems and of further increasing the level of safety and economic efficiency of NPP operation. (author)

  10. TRAC-PF1/MOD1 independent assessment: Semiscale Mod-2A intermediate break test S-IB-3

    International Nuclear Information System (INIS)

    Kmetyk, L.N.

    1986-02-01

    The TRAC-PF1/MOD1 independent assessment project at Sandia National Laboratories is part of an overall effort funded by the NRC to determine the ability of various system codes to predict the detailed thermal/hydraulic response of light water reactors during accident and off-normal conditions. The TRAC code is being assessed at SNLA against test data from various integral and separate effects test facilities. As part of this assessment matrix, an intermediate break test (S-IB-3), performed at the Semiscale Mod-2A facility, has been analyzed. Using an input model with a 3-D VESSEL component, the vessel and downcomer inventories during 3-IB-3 were generally well predicted, but the core heatup was underpredicted compared to data. An equivalent calculation with an all 1-D input model ran about twice as fast as our basecase analysis using a 3-D VESSEL in the input model, but the results of the two calculations diverged significantly for many parameters of interest, with the 3-D VESSEL model results in better agreement with data. 22 refs., 100 figs

  11. ABINIT: a computer code for matter; Abinit: un code au service de la matiere

    Energy Technology Data Exchange (ETDEWEB)

    Amadon, B.; Bottin, F.; Bouchet, J.; Dewaele, A.; Jollet, F.; Jomard, G.; Loubeyre, P.; Mazevet, S.; Recoules, V.; Torrent, M.; Zerah, G. [CEA Bruyeres-le-Chatel, 91 (France)

    2008-07-01

    The PAW (Projector Augmented Wave) method has been implemented in the ABINIT Code that computes electronic structures in atoms. This method relies on the simultaneous use of a set of auxiliary functions (in plane waves) and a sphere around each atom. This method allows the computation of systems including many atoms and gives the expression of energy, forces, stress... in terms of the auxiliary function only. We have generated atomic data for iron at very high pressure (over 200 GPa). We get a bcc-hcp transition around 10 GPa and the magnetic order disappears around 50 GPa. This method has been validated on a series of metals. The development of the PAW method has required a great effort for the massive parallelization of the ABINIT code. (A.C.)

  12. 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)

  13. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  14. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

  15. Interface between computational fluid dynamics (CFD) and plant analysis computer codes

    International Nuclear Information System (INIS)

    Coffield, R.D.; Dunckhorst, F.F.; Tomlinson, E.T.; Welch, J.W.

    1993-01-01

    Computational fluid dynamics (CFD) can provide valuable input to the development of advanced plant analysis computer codes. The types of interfacing discussed in this paper will directly contribute to modeling and accuracy improvements throughout the plant system and should result in significant reduction of design conservatisms that have been applied to such analyses in the past

  16. WSRC approach to validation of criticality safety computer codes

    International Nuclear Information System (INIS)

    Finch, D.R.; Mincey, J.F.

    1991-01-01

    Recent hardware and operating system changes at Westinghouse Savannah River Site (WSRC) have necessitated review of the validation for JOSHUA criticality safety computer codes. As part of the planning for this effort, a policy for validation of JOSHUA and other criticality safety codes has been developed. This policy will be illustrated with the steps being taken at WSRC. The objective in validating a specific computational method is to reliably correlate its calculated neutron multiplication factor (K eff ) with known values over a well-defined set of neutronic conditions. Said another way, such correlations should be: (1) repeatable; (2) demonstrated with defined confidence; and (3) identify the range of neutronic conditions (area of applicability) for which the correlations are valid. The general approach to validation of computational methods at WSRC must encompass a large number of diverse types of fissile material processes in different operations. Special problems are presented in validating computational methods when very few experiments are available (such as for enriched uranium systems with principal second isotope 236 U). To cover all process conditions at WSRC, a broad validation approach has been used. Broad validation is based upon calculation of many experiments to span all possible ranges of reflection, nuclide concentrations, moderation ratios, etc. Narrow validation, in comparison, relies on calculations of a few experiments very near anticipated worst-case process conditions. The methods and problems of broad validation are discussed

  17. Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

    International Nuclear Information System (INIS)

    Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

    2002-01-01

    A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

  18. Probabilistic evaluations for CANTUP computer code analysis improvement

    International Nuclear Information System (INIS)

    Florea, S.; Pavelescu, M.

    2004-01-01

    Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for

  19. Improvement of level-1 PSA computer code package

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Park, C. K.; Kim, K. Y.; Han, S. H.; Jung, W. D.; Chang, S. C.; Yang, J. E.; Sung, T. Y.; Kang, D. I.; Park, J. H.; Lee, Y. H.; Kim, S. H.; Hwang, M. J.; Choi, S. Y.

    1997-07-01

    This year the fifth (final) year of the phase-I of the Government-sponsored Mid- and Long-term Nuclear Power Technology Development Project. The scope of this subproject titled on `The improvement of level-1 PSA Computer Codes` is divided into two main activities : (1) improvement of level-1 PSA methodology, (2) development of applications methodology of PSA techniques to operations and maintenance of nuclear power plant. Level-1 PSA code KIRAP is converted to PC-Windows environment. For the improvement of efficiency in performing PSA, the fast cutset generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. Using about 30 foreign generic data sources, generic component reliability database (GDB) are developed considering dependency among source data. A computer program which handles dependency among data sources are also developed based on three stage bayesian updating technique. Common cause failure (CCF) analysis methods are reviewed and CCF database are established. Impact vectors can be estimated from this CCF database. A computer code, called MPRIDP, which handles CCF database are also developed. A CCF analysis reflecting plant-specific defensive strategy against CCF event is also performed. A risk monitor computer program, called Risk Monster, are being developed for the application to the operation and maintenance of nuclear power plant. The PSA application technique is applied to review the feasibility study of on-line maintenance and to the prioritization of in-service test (IST) of motor-operated valves (MOV). Finally, the root cause analysis (RCA) and reliability-centered maintenance (RCM) technologies are adopted and applied to the improvement of reliability of emergency diesel generators (EDG) of nuclear power plant. To help RCA and RCM analyses, two software programs are developed, which are EPIS and RAM Pro. (author). 129 refs., 20 tabs., 60 figs.

  20. Improvement of level-1 PSA computer code package

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Park, C. K.; Kim, K. Y.; Han, S. H.; Jung, W. D.; Chang, S. C.; Yang, J. E.; Sung, T. Y.; Kang, D. I.; Park, J. H.; Lee, Y. H.; Kim, S. H.; Hwang, M. J.; Choi, S. Y.

    1997-07-01

    This year the fifth (final) year of the phase-I of the Government-sponsored Mid- and Long-term Nuclear Power Technology Development Project. The scope of this subproject titled on 'The improvement of level-1 PSA Computer Codes' is divided into two main activities : 1) improvement of level-1 PSA methodology, 2) development of applications methodology of PSA techniques to operations and maintenance of nuclear power plant. Level-1 PSA code KIRAP is converted to PC-Windows environment. For the improvement of efficiency in performing PSA, the fast cutset generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. Using about 30 foreign generic data sources, generic component reliability database (GDB) are developed considering dependency among source data. A computer program which handles dependency among data sources are also developed based on three stage bayesian updating technique. Common cause failure (CCF) analysis methods are reviewed and CCF database are established. Impact vectors can be estimated from this CCF database. A computer code, called MPRIDP, which handles CCF database are also developed. A CCF analysis reflecting plant-specific defensive strategy against CCF event is also performed. A risk monitor computer program, called Risk Monster, are being developed for the application to the operation and maintenance of nuclear power plant. The PSA application technique is applied to review the feasibility study of on-line maintenance and to the prioritization of in-service test (IST) of motor-operated valves (MOV). Finally, the root cause analysis (RCA) and reliability-centered maintenance (RCM) technologies are adopted and applied to the improvement of reliability of emergency diesel generators (EDG) of nuclear power plant. To help RCA and RCM analyses, two software programs are developed, which are EPIS and RAM Pro. (author). 129 refs., 20 tabs., 60 figs

  1. Computer codes and methods for simulating accelerator driven systems

    International Nuclear Information System (INIS)

    Sartori, E.; Byung Chan Na

    2003-01-01

    A large set of computer codes and associated data libraries have been developed by nuclear research and industry over the past half century. A large number of them are in the public domain and can be obtained under agreed conditions from different Information Centres. The areas covered comprise: basic nuclear data and models, reactor spectra and cell calculations, static and dynamic reactor analysis, criticality, radiation shielding, dosimetry and material damage, fuel behaviour, safety and hazard analysis, heat conduction and fluid flow in reactor systems, spent fuel and waste management (handling, transportation, and storage), economics of fuel cycles, impact on the environment of nuclear activities etc. These codes and models have been developed mostly for critical systems used for research or power generation and other technological applications. Many of them have not been designed for accelerator driven systems (ADS), but with competent use, they can be used for studying such systems or can form the basis for adapting existing methods to the specific needs of ADS's. The present paper describes the types of methods, codes and associated data available and their role in the applications. It provides Web addresses for facilitating searches for such tools. Some indications are given on the effect of non appropriate or 'blind' use of existing tools to ADS. Reference is made to available experimental data that can be used for validating the methods use. Finally, some international activities linked to the different computational aspects are described briefly. (author)

  2. MQRAD, a computer code for synchrotron radiation from quadrupole magnets

    International Nuclear Information System (INIS)

    Morimoto, Teruhisa.

    1984-01-01

    The computer code, MQRAD, is developed for the calculation of the synchrotron radiation from the particles passing through quadrupole magnets at the straight section of the electron-positron colliding machine. This code computes the distributions of photon numbers and photon energies at any given points on the beam orbit. In this code, elements such as the quadrupole magnets and the drift spaces can be divided into many sub-elements in order to obtain the results with good accuracy. The synchrotron radiation produced by inserted quadrupole magnets at the interaction region of the electron-positron collider is one of the main background sources to the detector. The masking system against the synchrotron radiation at TRISTAN is very important because of the relatively high beam energy and the long straight section, which are 30 GeV and 100 meters, respectively. MQRAD has been used to design the masking system of the TOPAZ detector and the result is presented here as an example. (author)

  3. Phenomenological optical potentials and optical model computer codes

    International Nuclear Information System (INIS)

    Prince, A.

    1980-01-01

    An introduction to the Optical Model is presented. Starting with the purpose and nature of the physical problems to be analyzed, a general formulation and the various phenomenological methods of solution are discussed. This includes the calculation of observables based on assumed potentials such as local and non-local and their forms, e.g. Woods-Saxon, folded model etc. Also discussed are the various calculational methods and model codes employed to describe nuclear reactions in the spherical and deformed regions (e.g. coupled-channel analysis). An examination of the numerical solutions and minimization techniques associated with the various codes, is briefly touched upon. Several computer programs are described for carrying out the calculations. The preparation of input, (formats and options), determination of model parameters and analysis of output are described. The class is given a series of problems to carry out using the available computer. Interpretation and evaluation of the samples includes the effect of varying parameters, and comparison of calculations with the experimental data. Also included is an intercomparison of the results from the various model codes, along with their advantages and limitations. (author)

  4. Monocrystal sputtering by the computer simulation code ACOCT

    International Nuclear Information System (INIS)

    Yamamura, Yasunori; Takeuchi, Wataru.

    1987-09-01

    A new computer code ACOCT has been developed in order to simulate the atomic collisions in the crystalline target within the binary collision approximation. The present code is more convenient as compared with the MARLOWE code, and takes the higher-order simultaneous collisions into account. To cheke the validity of the ACOCT program, we have calculated sputtering yields for various ion-target combinations and compared with the MARLOWE results. It is found that the calculated yields by the ACOCT program are in good agreements with those by the MARLOWE code. The ejection patterns of sputtered atoms were also calculated for the major surfaces of fcc, bcc, diamond and hcp structures, and we have got reasonable agreements with experimental results. In order to know the effects of the simultaneous collision in the slowing down process the sputtering yields and the projected ranges are calculated, changeing the parameter of the criterion for the simultaneous collision, and the effect of the simultaneous collision is found to depend on the crystal orientation. (author)

  5. STADIC: a computer code for combining probability distributions

    International Nuclear Information System (INIS)

    Cairns, J.J.; Fleming, K.N.

    1977-03-01

    The STADIC computer code uses a Monte Carlo simulation technique for combining probability distributions. The specific function for combination of the input distribution is defined by the user by introducing the appropriate FORTRAN statements to the appropriate subroutine. The code generates a Monte Carlo sampling from each of the input distributions and combines these according to the user-supplied function to provide, in essence, a random sampling of the combined distribution. When the desired number of samples is obtained, the output routine calculates the mean, standard deviation, and confidence limits for the resultant distribution. This method of combining probability distributions is particularly useful in cases where analytical approaches are either too difficult or undefined

  6. RADTRAN 5 - A computer code for transportation risk analysis

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Kanipe, F.L.

    1993-01-01

    The RADTRAN 5 computer code has been developed to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI standard FORTRAN 77; the code contains significant advances in the methodology first pioneered with the LINK option of RADTRAN 4. A major application of the LINK methodology is route-specific analysis. Another application is comparisons of attributes along the same route segments. Nonradiological risk factors have been incorporated to allow users to estimate nonradiological fatalities and injuries that might occur during the transportation event(s) being analyzed. These fatalities include prompt accidental fatalities from mechanical causes. Values of these risk factors for the United States have been made available in the code as optional defaults. Several new health effects models have been published in the wake of the Hiroshima-Nagasaki dosimetry reassessment, and this has emphasized the need for flexibility in the RADTRAN approach to health-effects calculations. Therefore, the basic set of health-effects conversion equations in RADTRAN have been made user-definable. All parameter values can be changed by the user, but a complete set of default values are available for both the new International Commission on Radiation Protection model (ICRP Publication 60) and the recent model of the U.S. National Research Council's Committee on the Biological Effects of Radiation (BEIR V). The meteorological input data tables have been modified to permit optional entry of maximum downwind distances for each dose isopleth. The expected dose to an individual in each isodose area is also calculated and printed automatically. Examples are given that illustrate the power and flexibility of the RADTRAN 5 computer code. (J.P.N.)

  7. Compilation of the abstracts of nuclear computer codes available at CPD/IPEN

    International Nuclear Information System (INIS)

    Granzotto, A.; Gouveia, A.S. de; Lourencao, E.M.

    1981-06-01

    A compilation of all computer codes available at IPEN in S.Paulo are presented. These computer codes are classified according to Argonne National Laboratory - and Energy Nuclear Agency schedule. (E.G.) [pt

  8. CARP: a computer code and albedo data library for use by BREESE, the MORSE albedo package

    International Nuclear Information System (INIS)

    Emmett, M.B.; Rhoades, W.A.

    1978-10-01

    The CARP computer code was written to allow processing of DOT angular flux tapes to produce albedo data for use in the MORSE computer code. An albedo data library was produced containing several materials. 3 tables

  9. Nuclear model codes available at the Nuclear Energy Agency Computer Program Library (NEA-CPL)

    International Nuclear Information System (INIS)

    Sartori, E.; Garcia Viedma, L. de

    1976-01-01

    This paper briefly outlines the objectives of the NEA-CPL and its activities in the field of Nuclear Model Computer Codes. A short description of the computer codes available from the CPL in this field is also presented. (author)

  10. Computer codes in nuclear safety, radiation transport and dosimetry

    International Nuclear Information System (INIS)

    Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M.

    2006-01-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations

  11. Development validation and use of computer codes for inelastic analysis

    International Nuclear Information System (INIS)

    Jobson, D.A.

    1983-01-01

    A finite element scheme is a system which provides routines so carry out the operations which are common to all finite element programs. The list of items that can be provided as standard by the finite element scheme is surprisingly large and the list provided by the UNCLE finite element scheme is unusually comprehensive. This presentation covers the following: construction of the program, setting up a finite element mesh, generation of coordinates, incorporating boundary and load conditions. Program validation was done by creep calculations performed using CAUSE code. Program use is illustrated by calculating a typical inelastic analysis problem. This includes computer model of the PFR intermediate heat exchanger

  12. 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

  13. Some neutronics and thermal-hydraulics codes for reactor analysis using personal computers

    International Nuclear Information System (INIS)

    Woodruff, W.L.

    1990-01-01

    Some neutronics and thermal-hydraulics codes formerly available only for main frame computers may now be run on personal computers. Brief descriptions of the codes are provided. Running times for some of the codes are compared for an assortment of personal and main frame computers. With some limitations in detail, personal computer versions of the codes can be used to solve many problems of interest in reactor analyses at very modest costs. 11 refs., 4 tabs

  14. Analysis of the Length of Braille Texts in English Braille American Edition, the Nemeth Code, and Computer Braille Code versus the Unified English Braille Code

    Science.gov (United States)

    Knowlton, Marie; Wetzel, Robin

    2006-01-01

    This study compared the length of text in English Braille American Edition, the Nemeth code, and the computer braille code with the Unified English Braille Code (UEBC)--also known as Unified English Braille (UEB). The findings indicate that differences in the length of text are dependent on the type of material that is transcribed and the grade…

  15. Measurements by activation foils and comparative computations by MCNP code

    International Nuclear Information System (INIS)

    Kyncl, J.

    2008-01-01

    Systematic study of the radioactive waste minimisation problem is subject of the SPHINX project. Its idea is that burning or transmutation of the waste inventory problematic part will be realized in a nuclear reactor the fuel of which is in the form of liquid fluorides. In frame of the project, several experiments have been performed with so-called inserted experimental channel. The channel was filled up by the fluorides mixture, surrounded by six fuel assemblies with moderator and placed into LR-0 reactor vessel. This formation was brought to critical state and measurement with activation foil detectors were carried out at selected positions of the inserted channel. Main aim of the measurements was to determine reaction rates for the detectors mentioned. For experiment evaluation, comparative computations were accomplished by code MCNP4a. The results obtained show that very often, computed values of reaction rates differ substantially from the values that were obtained from the experiment. This contribution deals with analysis of the reasons of these differences from the point of view of computations by Monte Carlo method. The analysis of concrete cases shows that the inaccuracy of reaction rate computed is caused mostly by three circumstances:-space region that is occupied by detector is relatively very small;- microscopic effective cross-section R(E) of the reaction changes strongly with energy just in the energy interval that gives the greatest contribution to the reaction; - in the energy interval that gives the greatest contribution to reaction rate, the error of the computed neutron flux is great. These circumstances evoke that the computation of reaction rate with casual accuracy submits extreme demands on computing time. (Author)

  16. Computer code for shielding calculations of x-rays rooms

    International Nuclear Information System (INIS)

    Affonso, R.R.W.; Borges, D. da S.; Lava, D.D.; Moreira, M. de L.; Guimarães, A.C.F.

    2015-01-01

    The building an effective barrier against ionizing radiation present in radiographic rooms requires consideration of many variables. The methodology used for thickness specification of primary and secondary, barrier of a traditional radiographic room, considers the following factors: Use Factor, Occupational Factor, distance between the source and the wall, Workload, Kerma in the air and distance between the patient and the source. With these data it was possible to develop a computer code, which aims to identify and use variables in functions obtained through graphics regressions provided by NCRP-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) report, for shielding calculation of room walls, and the walls of the dark room and adjacent areas. With the implemented methodology, it was made a code validation by comparison of results with a study case provided by the report. The obtained values for thickness comprise different materials such as concrete, lead and glass. After validation it was made a case study of an arbitrary radiographic room.The development of the code resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in september/2011. (authors)

  17. PERCON: A flexible computer code for detailed thermal performance studies

    International Nuclear Information System (INIS)

    Boardman, F.B.; Collier, W.D.

    1975-07-01

    PERCON is a computer code which evaluates temperatures in three dimensions for a block containing heat sources and having coolant flow in one dimension. The solution is obtained at successive planes perpendicular to the coolant flow and the progression from one plane to the next occurs by the heat to the coolant determining convective boundary conditions at the next plane after due allowance being made for any lateral mixing or mass transfer between coolants. It is also possible to calculate the diametral change along a radius as a function of irradiation shrinkage and thermal expansion. This is used in a 'through life' calculation which evalates interaction pressure in tubular fuel elements. Physical property data used by the code may be specified as functions of temperature. The coolant flow may be specified, or alternatively derived by the program to satisfy either a specified overall pressure drop or mixed mean temperature rise. The pressure drop through each coolant is calculated and the flow modified, followed by a repeat of the temperature calculation, until the pressure imbalance between chosen flow channels at chosen axial positions is less than the specified convergence limit. A detailed description of the facilities in the code is given and some cases which have been studied are discussed. (U.K.)

  18. 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

  19. RELAP5/MOD2 blind calculation of GERDA small break test and data comparison

    International Nuclear Information System (INIS)

    Ogden, D.M.; Steiner, J.L.; Waterman, M.E.

    1985-01-01

    The Idaho National Engineering Laboratory (INEL), in support of the USNRC, has developed a RELAP5/MOD2 model of the GERDA facility to be used for analysis of the GERDA data, particularly relative to the phenomena of natural circulation and the boiler condenser mode of heat transfer. A blind calculation of GERDA Test 1605AA and a preliminary comparison with experimental data has been performed. The GERDA facility is a single loop integral facility with an electrically heated core. A general arrangement diagram of the facility is shown. The GERDA facility was designed for the performance of both separate effects and overall systems tests

  20. Parameters that affect parallel processing for computational electromagnetic simulation codes on high performance computing clusters

    Science.gov (United States)

    Moon, Hongsik

    What is the impact of multicore and associated advanced technologies on computational software for science? Most researchers and students have multicore laptops or desktops for their research and they need computing power to run computational software packages. Computing power was initially derived from Central Processing Unit (CPU) clock speed. That changed when increases in clock speed became constrained by power requirements. Chip manufacturers turned to multicore CPU architectures and associated technological advancements to create the CPUs for the future. Most software applications benefited by the increased computing power the same way that increases in clock speed helped applications run faster. However, for Computational ElectroMagnetics (CEM) software developers, this change was not an obvious benefit - it appeared to be a detriment. Developers were challenged to find a way to correctly utilize the advancements in hardware so that their codes could benefit. The solution was parallelization and this dissertation details the investigation to address these challenges. Prior to multicore CPUs, advanced computer technologies were compared with the performance using benchmark software and the metric was FLoting-point Operations Per Seconds (FLOPS) which indicates system performance for scientific applications that make heavy use of floating-point calculations. Is FLOPS an effective metric for parallelized CEM simulation tools on new multicore system? Parallel CEM software needs to be benchmarked not only by FLOPS but also by the performance of other parameters related to type and utilization of the hardware, such as CPU, Random Access Memory (RAM), hard disk, network, etc. The codes need to be optimized for more than just FLOPs and new parameters must be included in benchmarking. In this dissertation, the parallel CEM software named High Order Basis Based Integral Equation Solver (HOBBIES) is introduced. This code was developed to address the needs of the

  1. Further development of the computer code ATHLET-CD; Weiterentwicklung des Rechenprogramms ATHLET-CD. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin

    2016-10-15

    developed or improved in the frame of this project: Oxidation in steam and air atmosphere, processes in the lower plenum, melt behaviour, melt relocation into containment after failure of reactor pressure vessel, calculation of the nuclide inventory, release of fission products from the core, transport of fission products in the primary circuit. As a result of these improvements the quality of calculations of experiments as well as postulated reactor transients and accidents has been advanced considerably. Furthermore, the general robustness of the code has been improved. During the reporting period four versions of ATHLET-CD were released and frozen respectively. In October 2012 the version ATHLET-CD Mod 2.2 Cycle C was released as a frozen development version. Main improvements were the module AIDA, which allows the simulation of the late phase effects after relocation of melt into the lower ple-num, the modelling of the nitride formation in case of air ingress and the new version ATHLET 3.0A. In August 2013 the version ATHLET-CD Mod 3.0 Cycle A was released. In comparison to the version 2.2C further extensions and improvements were performed, e. g. general improvements of the code robustness or improvements concerning the simulation of oxidation effects. In contrast to the development version 2.2C the version 3.0A is an official release version, which is available to all licensed users of ATHLET-CD. With the new version the users are able to simulate a complete severe accident sequence from the early phase until the possible failure of the reactor pres-sure vessel after the relocation of melt into the lower plenum. In December 2014 the development version ATHLET-CD Mod 3.0 Cycle B was released. In March 2016 ATHLET-CD 3.1A was frozen. With this version the release of melt from the reactor pressure vessel can be simulated after the failure of the vessel. Furthermore, GRS participated in the international OECD/NEA project ''Benchmark Study of the Accident at the

  2. SHEAT for PC. A computer code for probabilistic seismic hazard analysis for personal computer, user's manual

    International Nuclear Information System (INIS)

    Yamada, Hiroyuki; Tsutsumi, Hideaki; Ebisawa, Katsumi; Suzuki, Masahide

    2002-03-01

    The SHEAT code developed at Japan Atomic Energy Research Institute is for probabilistic seismic hazard analysis which is one of the tasks needed for seismic Probabilistic Safety Assessment (PSA) of a nuclear power plant. At first, SHEAT was developed as the large sized computer version. In addition, a personal computer version was provided to improve operation efficiency and generality of this code in 2001. It is possible to perform the earthquake hazard analysis, display and the print functions with the Graphical User Interface. With the SHEAT for PC code, seismic hazard which is defined as an annual exceedance frequency of occurrence of earthquake ground motions at various levels of intensity at a given site is calculated by the following two steps as is done with the large sized computer. One is the modeling of earthquake generation around a site. Future earthquake generation (locations, magnitudes and frequencies of postulated earthquake) is modeled based on the historical earthquake records, active fault data and expert judgment. Another is the calculation of probabilistic seismic hazard at the site. An earthquake ground motion is calculated for each postulated earthquake using an attenuation model taking into account its standard deviation. Then the seismic hazard at the site is calculated by summing the frequencies of ground motions by all the earthquakes. This document is the user's manual of the SHEAT for PC code. It includes: (1) Outline of the code, which include overall concept, logical process, code structure, data file used and special characteristics of code, (2) Functions of subprogram and analytical models in them, (3) Guidance of input and output data, (4) Sample run result, and (5) Operational manual. (author)

  3. GAM-HEAT -- a computer code to compute heat transfer in complex enclosures

    International Nuclear Information System (INIS)

    Cooper, R.E.; Taylor, J.R.; Kielpinski, A.L.; Steimke, J.L.

    1991-02-01

    The GAM-HEAT code was developed for heat transfer analyses associated with postulated Double Ended Guillotine Break Loss Of Coolant Accidents (DEGB LOCA) resulting in a drained reactor vessel. In these analyses the gamma radiation resulting from fission product decay constitutes the primary source of energy as a function of time. This energy is deposited into the various reactor components and is re- radiated as thermal energy. The code accounts for all radiant heat exchanges within and leaving the reactor enclosure. The SRS reactors constitute complex radiant exchange enclosures since there are many assemblies of various types within the primary enclosure and most of the assemblies themselves constitute enclosures. GAM-HEAT accounts for this complexity by processing externally generated view factors and connectivity matrices, and also accounts for convective, conductive, and advective heat exchanges. The code is applicable for many situations involving heat exchange between surfaces within a radiatively passive medium. The GAM-HEAT code has been exercised extensively for computing transient temperatures in SRS reactors with specific charges and control components. Results from these computations have been used to establish the need for and to evaluate hardware modifications designed to mitigate results of postulated accident scenarios, and to assist in the specification of safe reactor operating power limits. The code utilizes temperature dependence on material properties. The efficiency of the code has been enhanced by the use of an iterative equation solver. Verification of the code to date consists of comparisons with parallel efforts at Los Alamos National Laboratory and with similar efforts at Westinghouse Science and Technology Center in Pittsburgh, PA, and benchmarked using problems with known analytical or iterated solutions. All comparisons and tests yield results that indicate the GAM-HEAT code performs as intended

  4. A computer code to simulate X-ray imaging techniques

    International Nuclear Information System (INIS)

    Duvauchelle, Philippe; Freud, Nicolas; Kaftandjian, Valerie; Babot, Daniel

    2000-01-01

    A computer code was developed to simulate the operation of radiographic, radioscopic or tomographic devices. The simulation is based on ray-tracing techniques and on the X-ray attenuation law. The use of computer-aided drawing (CAD) models enables simulations to be carried out with complex three-dimensional (3D) objects and the geometry of every component of the imaging chain, from the source to the detector, can be defined. Geometric unsharpness, for example, can be easily taken into account, even in complex configurations. Automatic translations or rotations of the object can be performed to simulate radioscopic or tomographic image acquisition. Simulations can be carried out with monochromatic or polychromatic beam spectra. This feature enables, for example, the beam hardening phenomenon to be dealt with or dual energy imaging techniques to be studied. The simulation principle is completely deterministic and consequently the computed images present no photon noise. Nevertheless, the variance of the signal associated with each pixel of the detector can be determined, which enables contrast-to-noise ratio (CNR) maps to be computed, in order to predict quantitatively the detectability of defects in the inspected object. The CNR is a relevant indicator for optimizing the experimental parameters. This paper provides several examples of simulated images that illustrate some of the rich possibilities offered by our software. Depending on the simulation type, the computation time order of magnitude can vary from 0.1 s (simple radiographic projection) up to several hours (3D tomography) on a PC, with a 400 MHz microprocessor. Our simulation tool proves to be useful in developing new specific applications, in choosing the most suitable components when designing a new testing chain, and in saving time by reducing the number of experimental tests

  5. A computer code to simulate X-ray imaging techniques

    Energy Technology Data Exchange (ETDEWEB)

    Duvauchelle, Philippe E-mail: philippe.duvauchelle@insa-lyon.fr; Freud, Nicolas; Kaftandjian, Valerie; Babot, Daniel

    2000-09-01

    A computer code was developed to simulate the operation of radiographic, radioscopic or tomographic devices. The simulation is based on ray-tracing techniques and on the X-ray attenuation law. The use of computer-aided drawing (CAD) models enables simulations to be carried out with complex three-dimensional (3D) objects and the geometry of every component of the imaging chain, from the source to the detector, can be defined. Geometric unsharpness, for example, can be easily taken into account, even in complex configurations. Automatic translations or rotations of the object can be performed to simulate radioscopic or tomographic image acquisition. Simulations can be carried out with monochromatic or polychromatic beam spectra. This feature enables, for example, the beam hardening phenomenon to be dealt with or dual energy imaging techniques to be studied. The simulation principle is completely deterministic and consequently the computed images present no photon noise. Nevertheless, the variance of the signal associated with each pixel of the detector can be determined, which enables contrast-to-noise ratio (CNR) maps to be computed, in order to predict quantitatively the detectability of defects in the inspected object. The CNR is a relevant indicator for optimizing the experimental parameters. This paper provides several examples of simulated images that illustrate some of the rich possibilities offered by our software. Depending on the simulation type, the computation time order of magnitude can vary from 0.1 s (simple radiographic projection) up to several hours (3D tomography) on a PC, with a 400 MHz microprocessor. Our simulation tool proves to be useful in developing new specific applications, in choosing the most suitable components when designing a new testing chain, and in saving time by reducing the number of experimental tests.

  6. The implementation of CP1 computer code in the Honeywell Bull computer in Brazilian Nuclear Energy Commission (CNEN)

    International Nuclear Information System (INIS)

    Couto, R.T.

    1987-01-01

    The implementation of the CP1 computer code in the Honeywell Bull computer in Brazilian Nuclear Energy Comission is presented. CP1 is a computer code used to solve the equations of punctual kinetic with Doppler feed back from the system temperature variation based on the Newton refrigeration equation (E.G.) [pt

  7. Interface design of VSOP'94 computer code for safety analysis

    International Nuclear Information System (INIS)

    Natsir, Khairina; Andiwijayakusuma, D.; Wahanani, Nursinta Adi; Yazid, Putranto Ilham

    2014-01-01

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects

  8. Interface design of VSOP'94 computer code for safety analysis

    Science.gov (United States)

    Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

    2014-09-01

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

  9. A computer code for analysis of severe accidents in LWRs

    International Nuclear Information System (INIS)

    2001-01-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  10. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  11. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  12. Computer codes for simulation of Angra 1 reactor steam generator

    International Nuclear Information System (INIS)

    Pinto, A.C.

    1978-01-01

    A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

  13. Computer code for simulating pressurized water reactor core

    International Nuclear Information System (INIS)

    Serrano, A.M.B.

    1978-01-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)

  14. Application of the RESRAD computer code to VAMP scenario S

    International Nuclear Information System (INIS)

    Gnanapragasam, E.K.; Yu, C.

    1997-03-01

    The RESRAD computer code developed at Argonne National Laboratory was among 11 models from 11 countries participating in the international Scenario S validation of radiological assessment models with Chernobyl fallout data from southern Finland. The validation test was conducted by the Multiple Pathways Assessment Working Group of the Validation of Environmental Model Predictions (VAMP) program coordinated by the International Atomic Energy Agency. RESRAD was enhanced to provide an output of contaminant concentrations in environmental media and in food products to compare with measured data from southern Finland. Probability distributions for inputs that were judged to be most uncertain were obtained from the literature and from information provided in the scenario description prepared by the Finnish Centre for Radiation and Nuclear Safety. The deterministic version of RESRAD was run repeatedly to generate probability distributions for the required predictions. These predictions were used later to verify the probabilistic RESRAD code. The RESRAD predictions of radionuclide concentrations are compared with measured concentrations in selected food products. The radiological doses predicted by RESRAD are also compared with those estimated by the Finnish Centre for Radiation and Nuclear Safety

  15. A computer code SPHINCS for sodium fire safety evaluation

    International Nuclear Information System (INIS)

    Yamaguchi, Akira

    2000-01-01

    A computer code SPHINCS solves coupled phenomena of thermal-hydraulics and sodium fire based on a multi-zone model. It deals with arbitrary number of rooms each of which is connected mutually by doorway and penetrations. With regard to the combustion phenomena, flame sheet model and liquid droplet combustion model are used for pool and spray fire, respectively, with the chemical equilibrium model using Gibbs free energy minimization method. The chemical reaction and mass and heat transfer are solved interactively. A specific feature of SPHINCS is detailed representation of thermal-hydraulics of a sodium pool and a steel liner, which is placed on the floor to prevent sodium-concrete contact. The author analyzed a series of pool combustion experiments, in which gas and liner temperatures are measured in detail. It has been found that good agreement is obtained and the SPHINCS has been validated with regard to the pool combustion phenomena. Further research needs are identified for the pool spreading modeling considering thermal deformation of liner and measurement of pool fluidity property of a mixture of liquid sodium and reaction products. SPHINCS code is to be used mainly in the safety evaluation of the consequence of sodium fire accident of liquid metal cooled fast reactor. (author)

  16. Development Of The Computer Code For Comparative Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Purwadi, Mohammad Dhandhang

    2001-01-01

    The qualitative and quantitative chemical analysis with Neutron Activation Analysis (NAA) is an importance utilization of a nuclear research reactor, and this should be accelerated and promoted in application and its development to raise the utilization of the reactor. The application of Comparative NAA technique in GA Siwabessy Multi Purpose Reactor (RSG-GAS) needs special (not commercially available yet) soft wares for analyzing the spectrum of multiple elements in the analysis at once. The application carried out using a single spectrum software analyzer, and comparing each result manually. This method really degrades the quality of the analysis significantly. To solve the problem, a computer code was designed and developed for comparative NAA. Spectrum analysis in the code is carried out using a non-linear fitting method. Before the spectrum analyzed, it was passed to the numerical filter which improves the signal to noise ratio to do the deconvolution operation. The software was developed using the G language and named as PASAN-K The testing result of the developed software was benchmark with the IAEA spectrum and well operated with less than 10 % deviation

  17. Computer code for the atomistic simulation of lattice defects and dynamics. [COMENT code

    Energy Technology Data Exchange (ETDEWEB)

    Schiffgens, J.O.; Graves, N.J.; Oster, C.A.

    1980-04-01

    This document has been prepared to satisfy the need for a detailed, up-to-date description of a computer code that can be used to simulate phenomena on an atomistic level. COMENT was written in FORTRAN IV and COMPASS (CDC assembly language) to solve the classical equations of motion for a large number of atoms interacting according to a given force law, and to perform the desired ancillary analysis of the resulting data. COMENT is a dual-purpose intended to describe static defect configurations as well as the detailed motion of atoms in a crystal lattice. It can be used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect migration, and defect stability.

  18. Users guide for NRC145-2 accident assessment computer code

    International Nuclear Information System (INIS)

    Pendergast, M.M.

    1982-08-01

    An accident assessment computer code has been developed for use at the Savannah River Plant. This computer code is based upon NRC Regulatory Guide 1.145 which provides guidence for accident assessements for power reactors. The code contains many options so that the user may utilize the code for many different assessments. For example the code can be used for non-nuclear assessments such as Sulpher Dioxide which may be required by the EPA. A discription of the code is contained in DP-1646. This document is a compilation of step-by-step instructions on how to use the code on the SRP IBM 3308 computer. This document consists of a number of tables which contain copies of computer listings. Some of the computer listings are copies of input; other listings give examples of computer output

  19. Standardization of computer programs - basis of the Czechoslovak library of nuclear codes

    International Nuclear Information System (INIS)

    Gregor, M.

    1987-01-01

    A standardized form of computer code documentation has been established in the CSSR in the field of reactor safety. Structure and content of the documentation are described and codes already subject to this process are mentioned. The formation of a Czechoslovak nuclear code library and facilitated discussion of safety reports containing results of standardized codes are aimed at

  20. Results of comparative RBMK neutron computation using VNIIEF codes (cell computation, 3D statics, 3D kinetics). Final report

    Energy Technology Data Exchange (ETDEWEB)

    Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A. [and others

    1995-12-31

    In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.

  1. Analysis of the UPTF Separate Effects Test 11 (steam-water counter-current flow in the broken loop hot leg) using RELAP5/MOD2

    International Nuclear Information System (INIS)

    Dillistone, M.J.

    1989-08-01

    RELAP5/MOD2 predictions of countercurrent flow limitation in the UPTF hot leg separate effects Test (test 11) are compared with the experimental data. The code underestimates, by a factor of more than three, the gas flow necessary to prevent liquid runback from the steam generator, and this is shown to be due to an oversimplified flow-regime map which does not allow the possibility of stratified flow in the hot leg riser. The predicted countercurrent flow is also shown to depend, wrongly, on the depth of liquid in the steam generator plenum. The same test is also modelled using a version of the code in which stratified flow in the riser is made possible. The gas flow needed to prevent liquid runback is then predicted quite well, but at all lower gas flows the code predicts that the flow is completely unrestricted - i.e. liquid flows between full flow and zero flow are not predicted. This is shown to happen because the code cannot calculate correctly the liquid level in the hot leg, mainly because of a numerical effect of upwind donoring in the momentum flux terms of the code's basic equations. It is also shown that the code cannot model the considerable effect of the ECCS injection pipe (which runs inside the hot leg) on the liquid level. (author)

  2. Final technical position on documentation of computer codes for high-level waste management

    International Nuclear Information System (INIS)

    Silling, S.A.

    1983-06-01

    Guidance is given for the content of documentation of computer codes which are used in support of a license application for high-level waste disposal. The guidelines cover theoretical basis, programming, and instructions for use of the code

  3. RADTRAN II: revised computer code to analyze transportation of radioactive material

    International Nuclear Information System (INIS)

    Taylor, J.M.; Daniel, S.L.

    1982-10-01

    A revised and updated version of the RADTRAN computer code is presented. This code has the capability to predict the radiological impacts associated with specific schemes of radioactive material shipments and mode specific transport variables

  4. CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1989-02-01

    A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  5. Sensitivity and uncertainty studies of the CRAC2 computer code

    International Nuclear Information System (INIS)

    Kocher, D.C.; Ward, R.C.; Killough, G.G.; Dunning, D.E. Jr.; Hicks, B.B.; Hosker, R.P. Jr.; Ku, J.Y.; Rao, K.S.

    1985-05-01

    This report presents a study of the sensitivity of early fatalities, early injuries, latent cancer fatalities, and economic costs for hypothetical nuclear reactor accidents as predicted by the CRAC2 computer code (CRAC = Calculation of Reactor Accident Consequences) to uncertainties in selected models and parameters used in the code. The sources of uncertainty that were investigated in the CRAC2 sensitivity studies include (1) the model for plume rise, (2) the model for wet deposition, (3) the procedure for meteorological bin-sampling involving the selection of weather sequences that contain rain, (4) the dose conversion factors for inhalation as they are affected by uncertainties in the physical and chemical form of the released radionuclides, (5) the weathering half-time for external ground-surface exposure, and (6) the transfer coefficients for estimating exposures via terrestrial foodchain pathways. The sensitivity studies were performed for selected radionuclide releases, hourly meteorological data, land-use data, a fixed non-uniform population distribution, a single evacuation model, and various release heights and sensible heat rates. Two important general conclusions from the sensitivity and uncertainty studies are as follows: (1) The large effects on predicted early fatalities and early injuries that were observed in some of the sensitivity studies apparently are due in part to the presence of thresholds in the dose-response models. Thus, the observed sensitivities depend in part on the magnitude of the radionuclide releases. (2) Some of the effects on predicted early fatalities and early injuries that were observed in the sensitivity studies were comparable to effects that were due only to the selection of different sets of weather sequences in bin-sampling runs. 47 figs., 50 tabs

  6. COMTA - a computer code for fuel mechanical and thermal analysis

    International Nuclear Information System (INIS)

    Basu, S.; Sawhney, S.S.; Anand, A.K.; Anantharaman, K.; Mehta, S.K.

    1979-01-01

    COMTA is a generalized computer code for integrity analysis of the free standing fuel cladding, with natural UO 2 or mixed oxide fuel pellets. Thermal and Mechanical analysis is done simultaneously for any power history of the fuel pin. For analysis, the fuel cladding is assumed to be axisymmetric and is subjected to axisymmetric load due to contact pressure, gas pressure, coolant pressure and thermal loads. Axial variation of load is neglected and creep and plasticity are assumed to occur at constant volume. The pellet is assumed to be made of concentric annuli. The fission gas release integral is dependent on the temperature and the power produced in each annulus. To calculate the temperature distribution in the fuel pin, the variation of bulk coolant temperature is given as an input to the code. Gap conductance is calculated at every time step, considering fuel densification, fuel relocation and gap closure, filler gas dilution by released fission gas, gap closure by expansion and irradiation swelling. Overall gap conductance is contributed by heat transfer due to the three modes; conduction convection and radiation as per modified Ross and Stoute model. Equilibrium equations, compatibility equations, stress strain relationships (including thermal strains and permanent strains due to creep and plasticity) are used to obtain triaxial stresses and strains. Thermal strain is assumed to be zero at hot zero power conditions. The boundary conditions are obtained for radial stresses at outside and inside surfaces by making these equal to coolant pressure and internal pressure respectively. A multi-mechanism creep model which accounts for thermal and irradiation creep is used to calculate the overall creep rate. Effective plastic strain is a function of effective stress and material constants. (orig.)

  7. Verification study of the FORE-2M nuclear/thermal-hydraulilc analysis computer code

    International Nuclear Information System (INIS)

    Coffield, R.D.; Tang, Y.S.; Markley, R.A.

    1982-01-01

    The verification of the LMFBR core transient performance code, FORE-2M, was performed in two steps. Different components of the computation (individual models) were verified by comparing with analytical solutions and with results obtained from other conventionally accepted computer codes (e.g., TRUMP, LIFE, etc.). For verification of the integral computation method of the code, experimental data in TREAT, SEFOR and natural circulation experiments in EBR-II were compared with the code calculations. Good agreement was obtained for both of these steps. Confirmation of the code verification for undercooling transients is provided by comparisons with the recent FFTF natural circulation experiments. (orig.)

  8. Computer codes for shaping the magnetic field of the JINR phasotron

    International Nuclear Information System (INIS)

    Zaplatin, N.L.; Morozov, N.A.

    1983-01-01

    The computer codes providing for the shaping the magnetic field of the JINR high current phasotron are presented. Using these codes the control for the magnetic field mapping was realized in on- or off-line regimes. Then these field parameters were calculated and ferromagnetic correcting elements and trim coils setting were chosen. Some computer codes were realised for the magnetic field horizontal component measurements. The data are presented on some codes possibilities. The codes were used on the EC-1010 and the CDC-6500 computers

  9. ANTEO: An optimised PC computer code for the steady state thermal hydraulic analysis of rod bundles

    International Nuclear Information System (INIS)

    Cevolani, S.

    1996-07-01

    The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of a such code was made possible by two facts: first, the increase the computing power of the desk machines; secondly, the fact several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes

  10. FISPIN - a computer code for nuclide inventory calculations

    International Nuclear Information System (INIS)

    Burstall, R.F.

    1979-10-01

    The code is used for assessment of three groups of nuclides, the actinides, the fission products, and structural materials. The methods of calculation are described, together with the input and output of the code and examples of both. Recommendations are given for the best use of the code. (author)

  11. Internal radiation dose calculations with the INREM II computer code

    International Nuclear Information System (INIS)

    Dunning, D.E. Jr.; Killough, G.G.

    1978-01-01

    A computer code, INREM II, was developed to calculate the internal radiation dose equivalent to organs of man which results from the intake of a radionuclide by inhalation or ingestion. Deposition and removal of radioactivity from the respiratory tract is represented by the Internal Commission on Radiological Protection Task Group Lung Model. A four-segment catenary model of the gastrointestinal tract is used to estimate movement of radioactive material that is ingested, or swallowed after being cleared from the respiratory tract. Retention of radioactivity in other organs is specified by linear combinations of decaying exponential functions. The formation and decay of radioactive daughters is treated explicitly, with each radionuclide in the decay chain having its own uptake and retention parameters, as supplied by the user. The dose equivalent to a target organ is computed as the sum of contributions from each source organ in which radioactivity is assumed to be situated. This calculation utilizes a matrix of dosimetric S-factors (rem/μCi-day) supplied by the user for the particular choice of source and target organs. Output permits the evaluation of components of dose from cross-irradiations when penetrating radiations are present. INREM II has been utilized with current radioactive decay data and metabolic models to produce extensive tabulations of dose conversion factors for a reference adult for approximately 150 radionuclides of interest in environmental assessments of light-water-reactor fuel cycles. These dose conversion factors represent the 50-year dose commitment per microcurie intake of a given radionuclide for 22target organs including contributions from specified source organs and surplus activity in the rest of the body. These tabulations are particularly significant in their consistent use of contemporary models and data and in the detail of documentation

  12. TRAC-PF1/MOD1 computer code

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1984-01-01

    The TRAC-P1 program was designed primarily for the analysis of large-break loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs). Because of its versatility, however, it can be applied directly to many analyses ranging from blowdowns in simple pipes to integral LOCA tests in multiloop facilities. A refined version, called TRAC-P1A, was released to the National Energy Software Center (NESC) in March 1979. Although it still treats the same class of problems, TRAC-P1A is more efficient than TRAC-P1 and incorporates improved hydrodynamic and heat-transfer models. It also is easier to implement on various computers. TRAC-PD2 contains improved reflood and heat-transfer models and improvements in the numerical solution methods. Although a large LOCA code, it has been applied successfully to small-break problems and to the Three Mile Island incident. Distinguishing characteristics of the TRAC-PF1/MOD1 are summarized

  13. MINIMARS interim report appendix halo model and computer code

    International Nuclear Information System (INIS)

    Santarius, J.F.; Barr, W.L.; Deng, B.Q.; Emmert, G.A.

    1985-01-01

    A tenuous, cool plasma called the halo shields the core plasma in a tandem mirror from neutral gas and impurities. The neutral particles are ionized and then pumped by the halo to the end tanks of the device, since flow of plasma along field lines is much faster than radial flow. Plasma reaching the end tank walls recombines, and the resulting neutral gas is vacuum pumped. The basic geometry of the MINIMARS halo is shown. For halo modeling purposes, the core plasma and cold gas regions may be treated as single radial zones leading to halo source and sink terms. The halo itself is differential into two major radial zones: halo scraper and halo dump. The halo scraper zone is defined by the radial distance required for the ion end plugging potential to drop to the central cell value, and thus have no effect on axial confinement; this distance is typically a sloshing plug ion Larmor diameter. The outer edge of the halo dump zone is defined by the last central cell flux tube to pass through the choke coil. This appendix will summarize the halo model that has been developed for MINIMARS and the methodology used in implementing that model as a computer code

  14. The TESS [Tandem Experiment Simulation Studies] computer code user's manual

    International Nuclear Information System (INIS)

    Procassini, R.J.

    1990-01-01

    TESS (Tandem Experiment Simulation Studies) is a one-dimensional, bounded particle-in-cell (PIC) simulation code designed to investigate the confinement and transport of plasma in a magnetic mirror device, including tandem mirror configurations. Mirror plasmas may be modeled in a system which includes an applied magnetic field and/or a self-consistent or applied electrostatic potential. The PIC code TESS is similar to the PIC code DIPSI (Direct Implicit Plasma Surface Interactions) which is designed to study plasma transport to and interaction with a solid surface. The codes TESS and DIPSI are direct descendants of the PIC code ES1 that was created by A. B. Langdon. This document provides the user with a brief description of the methods used in the code and a tutorial on the use of the code. 10 refs., 2 tabs

  15. 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)

  16. Towards advanced code simulators

    International Nuclear Information System (INIS)

    Scriven, A.H.

    1990-01-01

    The Central Electricity Generating Board (CEGB) uses advanced thermohydraulic codes extensively to support PWR safety analyses. A system has been developed to allow fully interactive execution of any code with graphical simulation of the operator desk and mimic display. The system operates in a virtual machine environment, with the thermohydraulic code executing in one virtual machine, communicating via interrupts with any number of other virtual machines each running other programs and graphics drivers. The driver code itself does not have to be modified from its normal batch form. Shortly following the release of RELAP5 MOD1 in IBM compatible form in 1983, this code was used as the driver for this system. When RELAP5 MOD2 became available, it was adopted with no changes needed in the basic system. Overall the system has been used for some 5 years for the analysis of LOBI tests, full scale plant studies and for simple what-if studies. For gaining rapid understanding of system dependencies it has proved invaluable. The graphical mimic system, being independent of the driver code, has also been used with other codes to study core rewetting, to replay results obtained from batch jobs on a CRAY2 computer system and to display suitably processed experimental results from the LOBI facility to aid interpretation. For the above work real-time execution was not necessary. Current work now centers on implementing the RELAP 5 code on a true parallel architecture machine. Marconi Simulation have been contracted to investigate the feasibility of using upwards of 100 processors, each capable of a peak of 30 MIPS to run a highly detailed RELAP5 model in real time, complete with specially written 3D core neutronics and balance of plant models. This paper describes the experience of using RELAP5 as an analyzer/simulator, and outlines the proposed methods and problems associated with parallel execution of RELAP5

  17. Assessment of the computer code COBRA/CFTL

    International Nuclear Information System (INIS)

    Baxi, C.B.; Burhop, C.J.

    1981-07-01

    The COBRA/CFTL code has been developed by Oak Ridge National Laboratory (ORNL) for thermal-hydraulic analysis of simulated gas-cooled fast breeder reactor (GCFR) core assemblies to be tested in the core flow test loop (CFTL). The COBRA/CFTL code was obtained by modifying the General Atomic code COBRA*GCFR. This report discusses these modifications, compares the two code results for three cases which represent conditions from fully rough turbulent flow to laminar flow. Case 1 represented fully rough turbulent flow in the bundle. Cases 2 and 3 represented laminar and transition flow regimes. The required input for the COBRA/CFTL code, a sample problem input/output and the code listing are included in the Appendices

  18. 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

  19. 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

  20. A computer code to design liquid containers for vehicles

    International Nuclear Information System (INIS)

    Parizi, H.B.; Fard, M.P.; Dolatabadi, A.

    2003-01-01

    We are presenting the development of a modular code for the simulation of the fluid sloshing that occurs in the liquid containers in vehicles. Sloshing occurs when a partially filled container of liquid goes through transient or steady external forces. Under such conditions, the free surface of the liquid may move and the liquid may impact on the walls of the container, exchanging forces. These forces may cause numerous harmful and undesirable consequences in the operation of the vehicle, such as vehicle turn over. The fluid mechanic equations that describe the fluid sloshing in the container and the dynamic equations that describe the movement of the container are solved separately in two different codes. The codes are coupled weekly, such that the output of one code will be used as the input to the other code in the same time step. The outputs of the fluid code are the forces and torques that are applied to the body of the container due to sloshing, whereas the output of the dynamic code are the translational and rotational velocities and accelerations of the container. The proposed software can be used to test the performance of the designed container under various operating condition and allow effective improvements to the container design. The proposed code is different than the presently available codes, in that it will provide a true simulation of the coupled fluid and structure interaction. (author)

  1. User manual for PACTOLUS: a code for computing power costs

    International Nuclear Information System (INIS)

    Huber, H.D.; Bloomster, C.H.

    1979-02-01

    PACTOLUS is a computer code for calculating the cost of generating electricity. Through appropriate definition of the input data, PACTOLUS can calculate the cost of generating electricity from a wide variety of power plants, including nuclear, fossil, geothermal, solar, and other types of advanced energy systems. The purpose of PACTOLUS is to develop cash flows and calculate the unit busbar power cost (mills/kWh) over the entire life of a power plant. The cash flow information is calculated by two principal models: the Fuel Model and the Discounted Cash Flow Model. The Fuel Model is an engineering cost model which calculates the cash flow for the fuel cycle costs over the project lifetime based on input data defining the fuel material requirements, the unit costs of fuel materials and processes, the process lead and lag times, and the schedule of the capacity factor for the plant. For nuclear plants, the Fuel Model calculates the cash flow for the entire nuclear fuel cycle. For fossil plants, the Fuel Model calculates the cash flow for the fossil fuel purchases. The Discounted Cash Flow Model combines the fuel costs generated by the Fuel Model with input data on the capital costs, capital structure, licensing time, construction time, rates of return on capital, tax rates, operating costs, and depreciation method of the plant to calculate the cash flow for the entire lifetime of the project. The financial and tax structure for both investor-owned utilities and municipal utilities can be simulated through varying the rates of return on equity and debt, the debt-equity ratios, and tax rates. The Discounted Cash Flow Model uses the principal that the present worth of the revenues will be equal to the present worth of the expenses including the return on investment over the economic life of the project. This manual explains how to prepare the input data, execute cases, and interpret the output results with the updated version of PACTOLUS. 11 figures, 2 tables

  2. User manual for PACTOLUS: a code for computing power costs.

    Energy Technology Data Exchange (ETDEWEB)

    Huber, H.D.; Bloomster, C.H.

    1979-02-01

    PACTOLUS is a computer code for calculating the cost of generating electricity. Through appropriate definition of the input data, PACTOLUS can calculate the cost of generating electricity from a wide variety of power plants, including nuclear, fossil, geothermal, solar, and other types of advanced energy systems. The purpose of PACTOLUS is to develop cash flows and calculate the unit busbar power cost (mills/kWh) over the entire life of a power plant. The cash flow information is calculated by two principal models: the Fuel Model and the Discounted Cash Flow Model. The Fuel Model is an engineering cost model which calculates the cash flow for the fuel cycle costs over the project lifetime based on input data defining the fuel material requirements, the unit costs of fuel materials and processes, the process lead and lag times, and the schedule of the capacity factor for the plant. For nuclear plants, the Fuel Model calculates the cash flow for the entire nuclear fuel cycle. For fossil plants, the Fuel Model calculates the cash flow for the fossil fuel purchases. The Discounted Cash Flow Model combines the fuel costs generated by the Fuel Model with input data on the capital costs, capital structure, licensing time, construction time, rates of return on capital, tax rates, operating costs, and depreciation method of the plant to calculate the cash flow for the entire lifetime of the project. The financial and tax structure for both investor-owned utilities and municipal utilities can be simulated through varying the rates of return on equity and debt, the debt-equity ratios, and tax rates. The Discounted Cash Flow Model uses the principal that the present worth of the revenues will be equal to the present worth of the expenses including the return on investment over the economic life of the project. This manual explains how to prepare the input data, execute cases, and interpret the output results. (RWR)

  3. Waste Evaporator Accident Simulation Using RELAP5 Computer Code

    International Nuclear Information System (INIS)

    POLIZZI, L.M.

    2004-01-01

    An evaporator is used on liquid waste from processing facilities to reduce the volume of the waste through heating the waste and allowing some of the water to be separated from the waste through boiling. This separation process allows for more efficient processing and storage of liquid waste. Commonly, the liquid waste consists of an aqueous solution of chemicals that over time could induce corrosion, and in turn weaken the tubes in the steam tube bundle of the waste evaporator that are used to heat the waste. This chemically induced corrosion could escalate into a possible tube leakage and/or the severance of a tube(s) in the tube bundle. In this paper, analyses of a waste evaporator system for the processing of liquid waste containing corrosive chemicals are presented to assess the system response to this accident scenario. This accident scenario is evaluated since its consequences can propagate to a release of hazardous material to the outside environment. It is therefore important to ensure that the evaporator system component structural integrity is not compromised, i.e. the design pressure and temperature of the system is not exceeded during the accident transient. The computer code used for the accident simulation is RELAP5-MOD31. The accident scenario analyzed includes a double-ended guillotine break of a tube in the tube bundle of the evaporator. A mitigated scenario is presented to evaluate the excursion of the peak pressure and temperature in the various components of the evaporator system to assess whether the protective actions and controls available are adequate to ensure that the structural integrity of the evaporator system is maintained and that no atmospheric release occurs

  4. Sensitivity and uncertainty studies of the CRAC2 computer code

    International Nuclear Information System (INIS)

    Kocher, D.C.; Ward, R.C.; Killough, G.G.; Dunning, D.E. Jr.; Hicks, B.B.; Hosker, R.P. Jr.; Ku, J.Y.; Rao, K.S.

    1987-01-01

    The authors have studied the sensitivity of health impacts from nuclear reactor accidents, as predicted by the CRAC2 computer code, to the following sources of uncertainty: (1) the model for plume rise, (2) the model for wet deposition, (3) the meteorological bin-sampling procedure for selecting weather sequences with rain, (4) the dose conversion factors for inhalation as affected by uncertainties in the particle size of the carrier aerosol and the clearance rates of radionuclides from the respiratory tract, (5) the weathering half-time for external ground-surface exposure, and (6) the transfer coefficients for terrestrial foodchain pathways. Predicted health impacts usually showed little sensitivity to use of an alternative plume-rise model or a modified rain-bin structure in bin-sampling. Health impacts often were quite sensitive to use of an alternative wet-deposition model in single-trial runs with rain during plume passage, but were less sensitive to the model in bin-sampling runs. Uncertainties in the inhalation dose conversion factors had important effects on early injuries in single-trial runs. Latent cancer fatalities were moderately sensitive to uncertainties in the weathering half-time for ground-surface exposures, but showed little sensitivity to the transfer coefficients for terrestrial foodchain pathways. Sensitivities of CRAC2 predictions to uncertainties in the models and parameters also depended on the magnitude of the source term, and some of the effects on early health effects were comparable to those that were due only to selection of different sets of weather sequences in bin-sampling

  5. Implementation of computer codes for performance assessment of the Republic repository of LLW/ILW Mochovce

    International Nuclear Information System (INIS)

    Hanusik, V.; Kopcani, I.; Gedeon, M.

    2000-01-01

    This paper describes selection and adaptation of computer codes required to assess the effects of radionuclide release from Mochovce Radioactive Waste Disposal Facility. The paper also demonstrates how these codes can be integrated into performance assessment methodology. The considered codes include DUST-MS for source term release, MODFLOW for ground-water flow and BS for transport through biosphere and dose assessment. (author)

  6. Three computer codes for safety and stability of large superconducting magnets

    International Nuclear Information System (INIS)

    Turner, L.R.

    1985-01-01

    For analyzing the safety and stability of large superconducting magnets, three computer codes TASS, SHORTURN, and SSICC have been developed, applicable to bath-cooled magnets, bath-cooled magnets with shorted turns, and magnets with internally cooled conductors respectively. The TASS code is described, and the use of the three codes is reviewed

  7. MLSOIL and DFSOIL - computer codes to estimate effective ground surface concentrations for dose computations

    International Nuclear Information System (INIS)

    Sjoreen, A.L.; Kocher, D.C.; Killough, G.G.; Miller, C.W.

    1984-11-01

    This report is a user's manual for MLSOIL (Multiple Layer SOIL model) and DFSOIL (Dose Factors for MLSOIL) and a documentation of the computational methods used in those two computer codes. MLSOIL calculates an effective ground surface concentration to be used in computations of external doses. This effective ground surface concentration is equal to (the computed dose in air from the concentration in the soil layers)/(the dose factor for computing dose in air from a plane). MLSOIL implements a five compartment linear-transfer model to calculate the concentrations of radionuclides in the soil following deposition on the ground surface from the atmosphere. The model considers leaching through the soil as well as radioactive decay and buildup. The element-specific transfer coefficients used in this model are a function of the k/sub d/ and environmental parameters. DFSOIL calculates the dose in air per unit concentration at 1 m above the ground from each of the five soil layers used in MLSOIL and the dose per unit concentration from an infinite plane source. MLSOIL and DFSOIL have been written to be part of the Computerized Radiological Risk Investigation System (CRRIS) which is designed for assessments of the health effects of airborne releases of radionuclides. 31 references, 3 figures, 4 tables

  8. PAPIRUS - a computer code for FBR fuel performance analysis

    International Nuclear Information System (INIS)

    Kobayashi, Y.; Tsuboi, Y.; Sogame, M.

    1991-01-01

    The FBR fuel performance analysis code PAPIRUS has been developed to design fuels for demonstration and future commercial reactors. A pellet structural model was developed to describe the generation, depletion and transport of vacancies and atomic elements in unified fashion. PAPIRUS results in comparison with the power - to - melt test data from HEDL showed validity of the code at the initial reactor startup. (author)

  9. Utilization of KENO-IV computer code with HANSEN-ROACH library

    International Nuclear Information System (INIS)

    Lima Barros, M. de; Vellozo, S.O.

    1982-01-01

    Several analysis with KENO-IV computer code, which is based in the Monte Carlo method, and the cross section library HANSEN-ROACH, were done, aiming to present the more convenient form to execute criticality calculations with this computer code and this cross sections. (E.G.) [pt

  10. CAT: a computer code for the automated construction of fault trees

    International Nuclear Information System (INIS)

    Apostolakis, G.E.; Salem, S.L.; Wu, J.S.

    1978-03-01

    A computer code, CAT (Computer Automated Tree, is presented which applies decision table methods to model the behavior of components for systematic construction of fault trees. The decision tables for some commonly encountered mechanical and electrical components are developed; two nuclear subsystems, a Containment Spray Recirculation System and a Consequence Limiting Control System, are analyzed to demonstrate the applications of CAT code

  11. TRACMAB. A computer code to form part of the link between the codes TRAC and MABEL

    International Nuclear Information System (INIS)

    Newbon, S.

    1982-05-01

    This report describes the function of the link program TRACMAB and provides a guide for users. The program is required to convert the thermal disequilibrium data output by the transient code TRAC into equilibrium data in a format compatible with the input data required by the code CAIN which in turn produces input data for MABEL. (author)

  12. Efficient Proximity Computation Techniques Using ZIP Code Data for Smart Cities †

    Directory of Open Access Journals (Sweden)

    Muhammad Harist Murdani

    2018-03-01

    Full Text Available In this paper, we are interested in computing ZIP code proximity from two perspectives, proximity between two ZIP codes (Ad-Hoc and neighborhood proximity (Top-K. Such a computation can be used for ZIP code-based target marketing as one of the smart city applications. A naïve approach to this computation is the usage of the distance between ZIP codes. We redefine a distance metric combining the centroid distance with the intersecting road network between ZIP codes by using a weighted sum method. Furthermore, we prove that the results of our combined approach conform to the characteristics of distance measurement. We have proposed a general and heuristic approach for computing Ad-Hoc proximity, while for computing Top-K proximity, we have proposed a general approach only. Our experimental results indicate that our approaches are verifiable and effective in reducing the execution time and search space.

  13. Efficient Proximity Computation Techniques Using ZIP Code Data for Smart Cities †.

    Science.gov (United States)

    Murdani, Muhammad Harist; Kwon, Joonho; Choi, Yoon-Ho; Hong, Bonghee

    2018-03-24

    In this paper, we are interested in computing ZIP code proximity from two perspectives, proximity between two ZIP codes ( Ad-Hoc ) and neighborhood proximity ( Top-K ). Such a computation can be used for ZIP code-based target marketing as one of the smart city applications. A naïve approach to this computation is the usage of the distance between ZIP codes. We redefine a distance metric combining the centroid distance with the intersecting road network between ZIP codes by using a weighted sum method. Furthermore, we prove that the results of our combined approach conform to the characteristics of distance measurement. We have proposed a general and heuristic approach for computing Ad-Hoc proximity, while for computing Top-K proximity, we have proposed a general approach only. Our experimental results indicate that our approaches are verifiable and effective in reducing the execution time and search space.

  14. SWAAM-LT: The long-term, sodium/water reaction analysis method computer code

    International Nuclear Information System (INIS)

    Shin, Y.W.; Chung, H.H.; Wiedermann, A.H.; Tanabe, H.

    1993-01-01

    The SWAAM-LT Code, developed for analysis of long-term effects of sodium/water reactions, is discussed. The theoretical formulation of the code is described, including the introduction of system matrices for ease of computer programming as a general system code. Also, some typical results of the code predictions for available large scale tests are presented. Test data for the steam generator design with the cover-gas feature and without the cover-gas feature are available and analyzed. The capabilities and limitations of the code are then discussed in light of the comparison between the code prediction and the test data

  15. Compendium of computer codes for the safety analysis of fast breeder reactors

    International Nuclear Information System (INIS)

    1977-10-01

    The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available

  16. Cooperation of experts' opinion, experiment and computer code development

    International Nuclear Information System (INIS)

    Wolfert, K.; Hicken, E.

    The connection between code development, code assessment and confidence in the analysis of transients will be discussed. In this manner, the major sources of errors in the codes and errors in applications of the codes will be shown. Standard problem results emphasize that, in order to have confidence in licensing statements, the codes must be physically realistic and the code user must be qualified and experienced. We will discuss why there is disagreement between the licensing authority and vendor concerning assessment of the fullfillment of safety goal requirements. The answer to the question lies in the different confidence levels of the assessment of transient analysis. It is expected that a decrease in the disagreement will result from an increased confidence level. Strong efforts will be made to increase this confidence level through improvements in the codes, experiments and related organizational strcutures. Because of the low probability for loss-of-coolant-accidents in the nuclear industry, assessment must rely on analytical techniques and experimental investigations. (orig./HP) [de

  17. Regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes

    International Nuclear Information System (INIS)

    Vitkova, M.; Kalchev, B.; Stefanova, S.

    2006-01-01

    The paper presents an overview of the regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes, which are used for safety assessment of the fuel design and the fuel utilization. Some requirements to the model development, verification and validation of the codes and analysis of code uncertainties are also define. Questions concerning Quality Assurance during development and implementation of the codes as well as preparation of a detailed verification and validation plan are briefly discussed

  18. MMA, A Computer Code for Multi-Model Analysis

    Science.gov (United States)

    Poeter, Eileen P.; Hill, Mary C.

    2007-01-01

    This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations. Many applications of MMA will

  19. Flux wire measurements in Cavalier for verifying computer code applications

    International Nuclear Information System (INIS)

    Fehr, M.; Stubbs, J.; Hosticka, B.

    1988-01-01

    The Cavalier and UVAR research reactors are to be converted from high-enrichment uranium (HEU) to low-enrichment uranium (LEU) fuel. As a first step, an extensive set of gold wire activation measurements has been taken on the Cavalier reactor. Axial traverses show internal consistency to the order of ±5%, while horizontal traverses show somewhat larger deviations. The activation measurements will be converted to flux measurements via the Thermos code and will then be used to verify the Leopard-2DB codes. The codes will ultimately be used to design an upgraded LEU core for the UVAR

  20. Quick look report for semiscale MOD-2C Test S-FS-11

    International Nuclear Information System (INIS)

    Plessinger, M.P.

    1985-11-01

    Results of a preliminary analysis of the fifth test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-11 simulated a pressurized water reactor transient initiated by a 50% break in a steam generator bottom feedwater line downstream of the check valve. With the exception of primary pressure, the initial conditions represented the initial conditions used for the C-E System 80 Final Safety Analysis Report (FSAR) Appendix 15B calculations. The transient included an initial 600 s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant followed by break isolation and affected loop steam generator refill with auxiliary feedwater. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overpressurization and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overpressurization and primary-to-secondary heat transfer. 64 figs

  1. Quick Look Report for Semiscale MOD-2C Test S-FS-2

    International Nuclear Information System (INIS)

    Boucher, T.J.; Chen, T.H.

    1985-01-01

    Results of a preliminary analysis of the first test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-2 simulated a pressurized water reactor transient initiated by a double-ended offset shear of a steam generator main steam line upstream of the flow restrictor. Initial conditions represented normal ''hot-standby'' operation. The transient included an initial 600-s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant at conditions required to allow a natural circulation cooldown. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overcooling and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overcooling and primary-to-secondary heat transfer. 57 figs., 3 tabs

  2. The assessment of RELAP5/MOD2 based on pressurizer transient experiments

    International Nuclear Information System (INIS)

    Xue Hanjun; Tanrikut, A.; Menzel, R.

    1992-03-01

    Two typical experiments have been performed in Chinese test facility under full pressure load corresponding to typical PWRs, 1) dynamic behavior of pressurizer due to relief valve operations (Case-I) is extremely important in transients and accident conditions regarding depressurization of PWR primary system; 2) Outsurge/Insurge operation is one of the transient which is often encountered and experienced in pressurizer systems due to pressure transients in primary system of PWRs. The simulation capability of RELAP5/MOD2 is good in comparison to experimental results. The physical models (such as interface model, stratification model), playing a major role in such simulation, seems to be realistic. The effect of realistic valve modeling in depressurization simulation is extremely important. Sufficient data for relief valve (the dynamic characteristics of valve) play a major role. The time dependent junction model and the trip valve model with a reduced discharge coefficient of 0.2 give better predictions in agreement with the experiment data while the trip valve models with discharge coefficient 1.0 yield overdepressurization. The simulation of outsurge/insurge transient yields satisfactory results. The thermal non-equilibrium model is important with respect to simulation of complicated physical phenomena in outsurge/insurge transient but has a negligible effect upon the depressurization simulation. (orig./HP)

  3. Fuel behavior modeling using the MARS computer code

    International Nuclear Information System (INIS)

    Faya, S.C.S.; Faya, A.J.G.

    1983-01-01

    The fuel behaviour modeling code MARS against experimental data, was evaluated. Two cases were selected: an early comercial PWR rod (Maine Yankee rod) and an experimental rod from the Canadian BWR program (Canadian rod). The MARS predictions are compared with experimental data and predictions made by other fuel modeling codes. Improvements are suggested for some fuel behaviour models. Mars results are satisfactory based on the data available. (Author) [pt

  4. Sensitivity analysis of FRAPCON-1 computer code to some parameters

    International Nuclear Information System (INIS)

    Chia, C.T.; Silva, C.F. da.

    1987-05-01

    A sensibility study of the code FRAPCON-1 was done for the following inout data: number of axial nodes, number of time steps and the axial power shape. Their influence in the code response concerning to the fuel center line temperature, stored energy, internal gas pressure, clad hoop strain and gap width were analyzed. The number of axial nodes has little influence, but care must be taken in the choice of the power axial profile and the time step length. (Author) [pt

  5. Computer code ANISN multiplying media and shielding calculation 2. Code description (input/output)

    International Nuclear Information System (INIS)

    Maiorino, J.R.

    1991-01-01

    The new code CCC-0514-ANISN/PC is described, as well as a ''GENERAL DESCRIPTION OF ANISN/PC code''. In addition to the ANISN/PC code, the transmittal package includes an interactive input generation programme called APE (ANISN Processor and Evaluator), which facilitates the work of the user in giving input. Also, a 21 group photon cross section master library FLUNGP.LIB in ISOTX format, which can be edited by an executable file LMOD.EXE, is included in the package. The input and output subroutines are reviewed. 6 refs, 1 fig., 1 tab

  6. Multi keno-VAX a modified version of the reactor computer code Multi keno-2

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    The reactor computer code Multi keno-2 is developed in Japan from the original Monte Carlo Keno-IV. By applications of this code on some real problems, fatal errors were detected. These errors are related to the restart option in the code. The restart option is essential for solving time-consuming problems on mini-computer like VAX-6320. These errors were corrected and other modifications were carried out in the code. Because of these modifications new input data description was written for the code. Thus a new VAX/VMS version for the program was developed which is also adaptable for mini-mainframes. This new developed program, called Multi keno-VAX is accepted in the Nea-IAEA data bank and is added to its international computer codes library. 1 fig.

  7. Multi keno-VAX a modified version of the reactor computer code Multi keno-2

    International Nuclear Information System (INIS)

    Imam, M.

    1995-01-01

    The reactor computer code Multi keno-2 is developed in Japan from the original Monte Carlo Keno-IV. By applications of this code on some real problems, fatal errors were detected. These errors are related to the restart option in the code. The restart option is essential for solving time-consuming problems on mini-computer like VAX-6320. These errors were corrected and other modifications were carried out in the code. Because of these modifications new input data description was written for the code. Thus a new VAX/VMS version for the program was developed which is also adaptable for mini-mainframes. This new developed program, called Multi keno-VAX is accepted in the Nea-IAEA data bank and is added to its international computer codes library. 1 fig

  8. FIRAC: a computer code to predict fire-accident effects in nuclear facilities

    International Nuclear Information System (INIS)

    Bolstad, J.W.; Krause, F.R.; Tang, P.K.; Andrae, R.W.; Martin, R.A.; Gregory, W.S.

    1983-01-01

    FIRAC is a medium-sized computer code designed to predict fire-induced flows, temperatures, and material transport within the ventilating systems and other airflow pathways in nuclear-related facilities. The code is designed to analyze the behavior of interconnected networks of rooms and typical ventilation system components. This code is one in a family of computer codes that is designed to provide improved methods of safety analysis for the nuclear industry. The structure of this code closely follows that of the previously developed TVENT and EVENT codes. Because a lumped-parameter formulation is used, this code is particularly suitable for calculating the effects of fires in the far field (that is, in regions removed from the fire compartment), where the fire may be represented parametrically. However, a fire compartment model to simulate conditions in the enclosure is included. This model provides transport source terms to the ventilation system that can affect its operation and in turn affect the fire

  9. Computer codes for simulating atomic-displacement cascades in solids subject to irradiation

    International Nuclear Information System (INIS)

    Asaoka, Takumi; Taji, Yukichi; Tsutsui, Tsuneo; Nakagawa, Masayuki; Nishida, Takahiko

    1979-03-01

    In order to study atomic displacement cascades originating from primary knock-on atoms in solids subject to incident radiation, the simulation code CASCADE/CLUSTER is adapted for use on FACOM/230-75 computer system. In addition, the code is modified so as to plot the defect patterns in crystalline solids. As other simulation code of the cascade process, MARLOWE is also available for use on the FACOM system. To deal with the thermal annealing of point defects produced in the cascade process, the code DAIQUIRI developed originally for body-centered cubic crystals is modified to be applicable also for face-centered cubic lattices. By combining CASCADE/CLUSTER and DAIQUIRI, we then prepared a computer code system CASCSRB to deal with heavy irradiation or saturation damage state of solids at normal temperature. Furthermore, a code system for the simulation of heavy irradiations CASCMARL is available, in which MARLOWE code is substituted for CASCADE in the CASCSRB system. (author)

  10. Citham a computer code for calculating fuel depletion-description, tests, modifications and evaluation

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1984-12-01

    The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author) [pt

  11. Recommendations for computer code selection of a flow and transport code to be used in undisturbed vadose zone calculations for TWRS immobilized wastes environmental analyses

    International Nuclear Information System (INIS)

    VOOGD, J.A.

    1999-01-01

    An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis

  12. Qualification of the new version of HAMMER computer code

    International Nuclear Information System (INIS)

    Chia, C.T.

    1984-06-01

    (HTEC) code were tested with a great number of diferent type of experiments. This experiments covers the most important parameters in neutronic calculations, such as the cell geometry and composition. The HTEC code results have been analysed and compared with experimental data and results given by the literature and simulated by HAMMER and LEOPARD codes. The quantities used for analysis were Keff and the following integral parameters: R28 - ratio of epicadmium-to-subcadmium 238 U captures; D25 - ratio of epicadmium-to-subcadmium 235 U fission; D28 - ratio of 238 U fissions to 235 U fissions; C - ratio of 238 U captures to 235 U fissions; RC02 - ratio of epicadmium-to-subcadmium 232 Th capture. The analysis shows that the results given by the code are in good agreement with the experimental data and the results given by the other codes. The calculation that have been done with the detailed ressonance profile tabulations of plutonium isotopes shows worst results than that obtained with the ressonance parameters. Almost all the simulated cases, shows that the HTEC results are closest to the experimental data than the HAMMER results, when one do not use the detailed ressonance profile tabulations of the plutonium isotopes. (Author) [pt

  13. Italian electricity supply contracts optimization: ECO computer code

    International Nuclear Information System (INIS)

    Napoli, G.; Savelli, D.

    1993-01-01

    The ECO (Electrical Contract Optimization) code written in the Microsoft WINDOWS 3.1 language can be handled with a 286 PC and a minimum of RAM. It consists of four modules, one for the calculation of ENEL (Italian National Electricity Board) tariffs, one for contractual time-of-use tariffs optimization, a table of tariff coefficients, and a module for monthly power consumption calculations based on annual load diagrams. The optimization code was developed by ENEA (Italian Agency for New Technology, Energy and the Environment) to help Italian industrial firms comply with new and complex national electricity supply contractual regulations and tariffs. In addition to helping industrial firms determine optimum contractual arrangements, the code also assists them in optimizing their choice of equipment and production cycles

  14. Fast Computation of Pulse Height Spectra Using SGRD Code

    Directory of Open Access Journals (Sweden)

    Humbert Philippe

    2017-01-01

    Full Text Available SGRD (Spectroscopy, Gamma rays, Rapid, Deterministic code is used for fast calculation of the gamma ray spectrum produced by a spherical shielded source and measured by a detector. The photon source lines originate from the radioactive decay of the unstable isotopes. The emission rate and spectrum of these primary sources are calculated using the DARWIN code. The leakage spectrum is separated in two parts, the uncollided component is transported by ray-tracing and the scattered component is calculated using a multigroup discrete ordinates method. The pulsed height spectrum is then simulated by folding the leakage spectrum with the detector response functions which are pre-calculated using MCNP5 code for each considered detector type. An application to the simulation of the gamma spectrum produced by a natural uranium ball coated with plexiglass and measured using a NaI detector is presented.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  16. 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)

  17. Development of a tracer transport option for the NAPSAC fracture network computer code

    International Nuclear Information System (INIS)

    Herbert, A.W.

    1990-06-01

    The Napsac computer code predicts groundwater flow through fractured rock using a direct fracture network approach. This paper describes the development of a tracer transport algorithm for the NAPSAC code. A very efficient particle-following approach is used enabling tracer transport to be predicted through large fracture networks. The new algorithm is tested against three test examples. These demonstrations confirm the accuracy of the code for simple networks, where there is an analytical solution to the transport problem, and illustrates the use of the computer code on a more realistic problem. (author)

  18. Interface code between WIMS-AECL and RFSP-IST for coupling computing

    International Nuclear Information System (INIS)

    Xu Liangwang; Liu Yu; Jia Baoshan

    2007-01-01

    A code based on the protocols of Telnet and FTP is developed with C++ for coupling computing between WIMS-AECL and RFSP-IST. the input document of WIMS-AECL and RFSP-ISP cna be generated automatically and be submitted to server, the output document will be downloaded by the end of computing. the function of analyzing standard output document is also included in this code. After simple updating, this code can meet the requirement of other code using input document, e.g. CATHENA. A pilot study of the relation between void fraction and reactivity in TACR, some valuable conclusions has been achieved. (authors)

  19. 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

  20. Quantum computation with topological codes from qubit to topological fault-tolerance

    CERN Document Server

    Fujii, Keisuke

    2015-01-01

    This book presents a self-consistent review of quantum computation with topological quantum codes. The book covers everything required to understand topological fault-tolerant quantum computation, ranging from the definition of the surface code to topological quantum error correction and topological fault-tolerant operations. The underlying basic concepts and powerful tools, such as universal quantum computation, quantum algorithms, stabilizer formalism, and measurement-based quantum computation, are also introduced in a self-consistent way. The interdisciplinary fields between quantum information and other fields of physics such as condensed matter physics and statistical physics are also explored in terms of the topological quantum codes. This book thus provides the first comprehensive description of the whole picture of topological quantum codes and quantum computation with them.

  1. Use of NESTLE computer code for NPP transition process analysis

    International Nuclear Information System (INIS)

    Gal'chenko, V.V.

    2001-01-01

    A newly created WWER-440 reactor model with use NESTLE code is discussed. Results of 'fast' and 'slow' transition processes based on it are presented. This model was developed for Rovno NPP reactor and it can be used also for WWER-1000 reactor in Zaporozhe NPP

  2. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  3. Validation of thermohydraulic codes by comparison of experimental results with computer simulations

    International Nuclear Information System (INIS)

    Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

    1989-01-01

    The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

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

    International Nuclear Information System (INIS)

    Tsujimoto, Iwao; Naito, Yoshitaka.

    1992-11-01

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

  5. Development and application of computer codes for multidimensional thermalhydraulic analyses of nuclear reactor components

    International Nuclear Information System (INIS)

    Carver, M.B.

    1983-01-01

    Components of reactor systems and related equipment are identified in which multidimensional computational thermal hydraulics can be used to advantage to assess and improve design. Models of single- and two-phase flow are reviewed, and the governing equations for multidimensional analysis are discussed. Suitable computational algorithms are introduced, and sample results from the application of particular multidimensional computer codes are given

  6. Wake structure measurements at the MOD-2 cluster test facility at Goodnoe Hills, Washington

    Energy Technology Data Exchange (ETDEWEB)

    Zambrano, T.G.; Gyatt, G.W.

    1983-12-01

    A field measurement programme was carried out at the cluster of three MOD-2 wind turbines located at Goodnoe Hills, Washington, to determine the rate of decay of wake velocity deficit with down-wind distance in various meteorological conditions. Measurements were taken at hub height (60 m). Wake wind speeds were measured using a radiosonde suspended from a tethered balloon, its position being determined from a grid of ground stakes. Instantaneous readings were recorded by each system every two seconds and averaged over ten-minute periods. As a control experiment, the sonde was also operated next to the meteorological tower to calibrate the instrumentation. Measurements were also made down wind with the turbine off to determine the magnitude of terrain-induced variations in wind speed. Downstream distances of 274.3, 457.2, 640.1 and 823.0 m from the turbine, corresponding to 3, 5, 7 and 9 rotor diameters D, were investigated. There was considerable scatter in the observed 10 min average downstream/free-stream velocity ratios. Turbine-on velocity ratios showed even greater scatter, suggesting that only some measurements were, in fact, representative of wake centre-line velocities, and that others were made off centre line due to wake meander or wind shift. Isolation of the high wind speed (13.4 to 20.1 m/s) velocity ratios, however, revealed velocity deficits of up to about 50% at 3D and 5% at 5D downstream. Measurements at greater downstream distances showed no wake deficit within the limits of resolution of the experiment, indicating that the wake had recovered to free-stream conditions.

  7. Computer code for the costing and sizing of TNS tokamaks

    International Nuclear Information System (INIS)

    Sink, D.A.; Iwinski, E.M.

    1977-01-01

    A FORTRAN code for the COsting And Sizing of Tokamaks (COAST) is described. The code was written to conduct detailed analyses on the engineering features of the next tokamak fusion device following TFTR. The ORNL/Westinghouse study of TNS (The Next Step) has involved the investigation of a number of device options, each over a wide range of plasma sizes. A generalized description of TNS is incorporated in the code and includes refined modeling of over forty systems and subsystems. Considerable detailed design and analyses have provided the basis for the thermal, electrical, mechanical, nuclear, chemical, vacuum, and facility engineering of the various subsystems. Currently, the code provides a tool for the systematic comparison of four toroidal field (TF) coil technologies allowing both D-shaped and circular coils. The coil technologies are: (1) copper (both room temperature and liquid-nitrogen cooled), (2) superconducting NbTi, (3) superconducting Nb 3 Sn, and (4) a Cu/NbTi/ hybrid. For the poloidal field (PF) coil systems copper conductors are assumed. The ohmic heating (OH) coils are located within the machine bore and have an air core, while the shaping field (SF) coils are located either within or outside the TF coils. The PF coil self and mutual inductances are calculated from the geometry, and the PF coil power supplies are modeled to account for time-dependent profiles for voltages and currents as governed by input data. Plasma heating is assumed to be by neutral beams, and impurity control is either passive or by a poloidal divertor system. The size modeling allows considerable freedom in specifying physics assumptions, operating scenarios, TF operating margin, and component geometric and performance parameters. Cost relationships have been developed for both plant and capital equipment and for annual utility and fuel expenses. The code has been used successfully to reproduce the sizing and costing of TFTR in order to calibrate the various models

  8. Development of a Computer Code for the Estimation of Fuel Rod Failure

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I.H.; Ahn, H.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    Much research has already been performed to obtain the information on the degree of failed fuel rods from the primary coolant activities of operating PWRs in the last few decades. The computer codes that are currently in use for domestic nuclear power plants, such as CADE code and ABB-CE codes developed by Westinghouse and ABB-CE, respectively, still give significant overall errors in estimating the failed fuel rods. In addition, with the CADE code, it is difficult to predict the degree of fuel rod failures during the transient period of nuclear reactor operation, where as the ABB-CE codes are relatively more difficult to use for end-users. In particular, the rapid progresses made recently in the area of the computer hardware and software systems that their computer programs be more versatile and user-friendly. While the MS windows system that is centered on the graphic user interface and multitasking is now in widespread use, the computer codes currently employed at the nuclear power plants, such as CADE and ABB-CE codes, can only be run on the DOS system. Moreover, it is desirable to have a computer code for the fuel rod failure estimation that can directly use the radioactivity data obtained from the on-line monitoring system of the primary coolant activity. The main purpose of this study is, therefore, to develop a Windows computer code that can predict the location, the number of failed fuel rods,and the degree of failures using the radioactivity data obtained from the primary coolant activity for PWRs. Another objective is to combine this computer code with the on-line monitoring system of the primary coolant radioactivity at Kori 3 and 4 operating nuclear power plants and enable their combined use for on-line evaluation of the number and degree of fuel rod failures. (author). 49 refs., 85 figs., 30 tabs.

  9. Assessment of RELAP5/MOD2 against critical flow data from Marviken tests JIT 11 and CFT 21

    International Nuclear Information System (INIS)

    Rosdahl, O.; Caraher, D.

    1986-09-01

    RELAP5/MOD2 simulations of the critical flow of saturated steam are reported together with simulations of the critical flow of subcooled liquid and a low quality two-phase mixture. The experiments which were simulated used nozzle diameters of 0.3 m and 0.5 m. RELAP5 overpredicted the experimental flow rates by 10 to 25% unless discharge coefficients were applied

  10. RELAP5/MOD2 post-test calculation of the OECD LOFT experiment LP-SB-2

    International Nuclear Information System (INIS)

    Perez, J.; Mendizabal, R.

    1992-04-01

    This document presents the analysis of the OECD LOFT LP-SB-2 Experiment performed by the Consejo de Seguridad Nuclear of Spain working group making use of RELAP5/MOD2 in the frame of the Spanish LOFT Project. LB-SB-2 experiment studies the effect of a delayed pump trip in a small break LOCA scenario with a 3-inch equivalent diameter break in the hot leg of a commercial PWR

  11. Selection of a computer code for Hanford low-level waste engineered-system performance assessment

    International Nuclear Information System (INIS)

    McGrail, B.P.; Mahoney, L.A.

    1995-10-01

    Planned performance assessments for the proposed disposal of low-level waste (LLW) glass produced from remediation of wastes stored in underground tanks at Hanford, Washington will require calculations of radionuclide release rates from the subsurface disposal facility. These calculations will be done with the aid of computer codes. Currently available computer codes were ranked in terms of the feature sets implemented in the code that match a set of physical, chemical, numerical, and functional capabilities needed to assess release rates from the engineered system. The needed capabilities were identified from an analysis of the important physical and chemical process expected to affect LLW glass corrosion and the mobility of radionuclides. The highest ranked computer code was found to be the ARES-CT code developed at PNL for the US Department of Energy for evaluation of and land disposal sites

  12. Report on nuclear industry quality assurance procedures for safety analysis computer code development and use

    International Nuclear Information System (INIS)

    Sheron, B.W.; Rosztoczy, Z.R.

    1980-08-01

    As a result of a request from Commissioner V. Gilinsky to investigate in detail the causes of an error discovered in a vendor Emergency Core Cooling System (ECCS) computer code in March, 1978, the staff undertook an extensive investigation of the vendor quality assurance practices applied to safety analysis computer code development and use. This investigation included inspections of code development and use practices of the four major Light Water Reactor Nuclear Steam Supply System vendors and a major reload fuel supplier. The conclusion reached by the staff as a result of the investigation is that vendor practices for code development and use are basically sound. A number of areas were identified, however, where improvements to existing vendor procedures should be made. In addition, the investigation also addressed the quality assurance (QA) review and inspection process for computer codes and identified areas for improvement

  13. LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

    International Nuclear Information System (INIS)

    Bordner, G.L.

    1979-10-01

    The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

  14. Development and validation of GWHEAD, a three-dimensional groundwater head computer code

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Root, R.W.; Routt, K.R.

    1980-03-01

    A computer code has been developed to solve the groundwater flow equation in three dimensions. The code has finite-difference approximations solved by the strongly implicit solution procedure. Input parameters to the code include hydraulic conductivity, specific storage, porosity, accretion (recharge), and initial hydralic head. These parameters may be input as varying spatially. The hydraulic conductivity may be input as isotropic or anisotropic. The boundaries either may permit flow across them or may be impermeable. The code has been used to model leaky confined groundwater conditions and spherical flow to a continuous point sink, both of which have exact analytical solutions. The results generated by the computer code compare well with those of the analytical solutions. The code was designed to be used to model groundwater flow beneath fuel reprocessing and waste storage areas at the Savannah River Plant

  15. A new 3-D integral code for computation of accelerator magnets

    International Nuclear Information System (INIS)

    Turner, L.R.; Kettunen, L.

    1991-01-01

    For computing accelerator magnets, integral codes have several advantages over finite element codes; far-field boundaries are treated automatically, and computed field in the bore region satisfy Maxwell's equations exactly. A new integral code employing edge elements rather than nodal elements has overcome the difficulties associated with earlier integral codes. By the use of field integrals (potential differences) as solution variables, the number of unknowns is reduced to one less than the number of nodes. Two examples, a hollow iron sphere and the dipole magnet of Advanced Photon Source injector synchrotron, show the capability of the code. The CPU time requirements are comparable to those of three-dimensional (3-D) finite-element codes. Experiments show that in practice it can realize much of the potential CPU time saving that parallel processing makes possible. 8 refs., 4 figs., 1 tab

  16. Use of ETOG and ETOT computer codes for preparating the Library of LEOPARD with data from ENDFIB-IV

    International Nuclear Information System (INIS)

    Cunha Menezes Filho, A. da.

    1983-01-01

    The modifications carried out in the ETOT-3 and ETOG-3 computer codes used for preparating the thermal (172 energy groups) and epithermal (54 energy groups) libraries, respectivelly, of LEOPARD computer code, are presented. (M.C.K.) [pt

  17. Intercomparison on the usage of computational codes in radiation dosimetry

    International Nuclear Information System (INIS)

    Ilic, R.; Pesic, M.; Pavlovic, R.

    2003-01-01

    SRNA-2KG software package was modified for this work to include necessary input and output data and for predicted voxelized geometry and dosimetry. SRNA is a Monte Carlo code developed for applications in proton transport, radiotherapy and dosimetry. Protons within energy range from 100 keV to 250 MeV with predefined spectra are transported in 3D geometry through material zones confined by planes and second order surfaces or in 3D voxelized geometry. The code can treat proton transport in a few hundred different materials including elements from Z=1 to Z=98. Simulation of proton transport is based on the multiple scattering theory of charged particles and on the model for compound nucleus decay

  18. New computational methods used in the lattice code DRAGON

    International Nuclear Information System (INIS)

    Marleau, G.; Hebert, A.; Roy, R.

    1992-01-01

    The lattice code DRAGON is used to perform transport calculations inside cells and assemblies for multidimensional geometry using the collision probability method, including the interface current and J ± techniques. Typical geometries that can be treated using this code include CANDU 2-dimensional clusters, CANDU 3-dimensional assemblies, pressurized water reactor (PWR) rectangular and hexagonal assemblies. It contains a self-shielding module for the treatment of microscopic cross section libraries and a depletion module for burnup calculations. DRAGON was written in a modular form in such a way as to accept easily new collision probability options and make them readily available to all the modules that require collision probability matrices like the self-shielding module, the flux solution module and the homogenization module. In this paper the authors present an overview of DRAGON and discuss some of the methods that were implemented in DRAGON in order to improve on its performance

  19. Computer codes used during upgrading activities at MINT TRIGA reactor

    International Nuclear Information System (INIS)

    Mohammad Suhaimi Kassim; Adnan Bokhari; Mohd Idris Taib

    1999-01-01

    MINT TRIGA Reactor is a 1-MW swimming pool nuclear research reactor commissioned in 1982. In 1993, a project was initiated to upgrade the thermal power to 2 MW. The IAEA assistance was sought to assist the various activities relevant to an upgrading exercise. For neutronics calculations, the IAEA has provided expert assistance to introduce the WIMS code, TRIGAP, and EXTERMINATOR2. For thermal-hydraulics calculations, PARET and RELAP5 were introduced. Shielding codes include ANISN and MERCURE. However, in the middle of 1997, MINT has decided to change the scope of the project to safety upgrading of the MINT Reactor. This paper describes some of the activities carried out during the upgrading process. (author)

  20. Computations for Truck Sliding with TRUCK 3.1 Code

    Science.gov (United States)

    1989-08-01

    16 REFERENCES 1. L u. \\Villiam N.. Hobbs. Norman P. and Atkinson, Michael. TRUCK 3.1-An Improrcd Digital (’oiputtr Program for Calculating the Response...for Operations and Plans ATIN: Technical Libary Director of Chemical & Nuear Operations Dpartnt of the AIW Waskbington, DC 20310 1 Cocaeder US Ay...Lawrenoe Livermore Lab. ATIN: Code 2124, Tedhnical ATTN: Tech Info Dept L-3 Reports Libary P.O. Be 808 Monterey, CA 93940 Livermore, CA 94550 AFSC

  1. GATE: computation code for medical imagery, radiotherapy and dosimetry

    International Nuclear Information System (INIS)

    Jan, S.

    2010-01-01

    The author presents the GATE code, a simulation software based on the Geant4 development environment developed by the CERN (the European organization for nuclear research) which enables Monte-Carlo type simulation to be developed for tomography imagery using ionizing radiation, and radiotherapy examinations (conventional and hadron therapy) to be simulated. The authors concentrate on the use of medical imagery in carcinology. They comment some results obtained in nuclear imagery and in radiotherapy

  2. A user's guide to GENEX, SDR, and related computer codes

    International Nuclear Information System (INIS)

    Brissenden, R.J.; Durston, C.

    1968-08-01

    This series of codes will be of use in a variety of fields connected with reactor physics, examples of which are: (a) In evaluation of nuclear data in which the RESP-GENEX part of the system would be used to examine and produce a cross-section set based on the theories and experiments of the nuclear physicists. The approximations in GENEX must however be kept in mind, the chief one being the diagonal expansion approximation of the inverse level matrix originally due to Bethe which precludes a correct representation of strong interference effects (the Lynn effect). (b) In the calculation of Doppler effects or other resonance effects such as establishing equivalence relationships, approximate resonance treatments, etc. A given set of tapes generated by GENEX (or by some other means into the GENEX format) would be used to run the SDH code. The SDR code produces cross-sections and reaction rates over any group structure within its working range. In situations with complex geometries the spatial representation of SDR is liable to be inadequate and in these circumstances it is recommended that the reaction rates are not used directly but instead the cross-sections are used in a more accurate spatial calculation to produce revised reaction rates. (c) Finally the system may be used for a variety of special investigations such as an analysis of the variance of the Doppler coefficient in fast reactors or the accurate assessment of ideal integral measurements, (for instance the Aldermaston sphere experiment

  3. Code and papers: computing publication patterns in the LHC era

    CERN Multimedia

    CERN. Geneva

    2012-01-01

    Publications in scholarly journals establish the body of knowledge deriving from scientific research; they also play a fundamental role in the career path of scientists and in the evaluation criteria of funding agencies. This presentation reviews the evolution of computing-oriented publications in HEP following the start of operation of LHC. Quantitative analyses are illustrated, which document the production of scholarly papers on computing-related topics by HEP experiments and core tools projects (including distributed computing R&D), and the citations they receive. Several scientometric indicators are analyzed to characterize the role of computing in HEP literature. Distinctive features of scholarly publication production in the software-oriented and hardware-oriented experimental HEP communities are highlighted. Current patterns and trends are compared to the situation in previous generations' HEP experiments at LEP, Tevatron and B-factories. The results of this scientometric analysis document objec...

  4. ENDF/B Pre-Processing Codes: Implementing and testing on a Personal Computer

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1987-05-01

    This document describes the contents of the diskettes containing the ENDF/B Pre-Processing codes by D.E. Cullen, and example data for use in implementing and testing these codes on a Personal Computer of the type IBM-PC/AT. Upon request the codes are available from the IAEA Nuclear Data Section, free of charge, on a series of 7 diskettes. (author)

  5. User's guide for vectorized code EQUIL for calculating equilibrium chemistry on Control Data STAR-100 computer

    Science.gov (United States)

    Kumar, A.; Graves, R. A., Jr.; Weilmuenster, K. J.

    1980-01-01

    A vectorized code, EQUIL, was developed for calculating the equilibrium chemistry of a reacting gas mixture on the Control Data STAR-100 computer. The code provides species mole fractions, mass fractions, and thermodynamic and transport properties of the mixture for given temperature, pressure, and elemental mass fractions. The code is set up for the electrons H, He, C, O, N system of elements. In all, 24 chemical species are included.

  6. 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

  7. Development of a computer code for Dalat research reactor transient analysis

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong

    2003-01-01

    DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)

  8. Structural dynamics in LMFBR containment analysis: a brief survey of computational methods and codes

    International Nuclear Information System (INIS)

    Chang, Y.W.; Gvildys, J.

    1977-01-01

    In recent years, the use of computer codes to study the response of primary containment of large, liquid-metal fast breeder reactors (LMFBR) under postulated accident conditions has been adopted by most fast reactor projects. Since the first introduction of REXCO-H containment code in 1969, a number of containment codes have evolved and been reported in the literature. The paper briefly summarizes the various numerical methods commonly used in containment analysis in computer programs. They are compared on the basis of truncation errors resulting in the numerical approximation, the method of integration, the resolution of the computed results, and the ease of programming in computer codes. The aim of the paper is to provide enough information to an analyst so that he can suitably define his choice of method, and hence his choice of programs

  9. Comparison of Computational Electromagnetic Codes for Prediction of Low-Frequency Radar Cross Section

    National Research Council Canada - National Science Library

    Lash, Paul C

    2006-01-01

    .... The goal of this research is to compare the capabilities of three computational electromagnetic codes for use in production of RCS signature assessments at low frequencies in terms of performance...

  10. EXTRAN: A computer code for estimating concentrations of toxic substances at control room air intakes

    International Nuclear Information System (INIS)

    Ramsdell, J.V.

    1991-03-01

    This report presents the NRC staff with a tool for assessing the potential effects of accidental releases of radioactive materials and toxic substances on habitability of nuclear facility control rooms. The tool is a computer code that estimates concentrations at nuclear facility control room air intakes given information about the release and the environmental conditions. The name of the computer code is EXTRAN. EXTRAN combines procedures for estimating the amount of airborne material, a Gaussian puff dispersion model, and the most recent algorithms for estimating diffusion coefficients in building wakes. It is a modular computer code, written in FORTRAN-77, that runs on personal computers. It uses a math coprocessor, if present, but does not require one. Code output may be directed to a printer or disk files. 25 refs., 8 figs., 4 tabs

  11. Double folding model of nucleus-nucleus potential: formulae, iteration method and computer code

    International Nuclear Information System (INIS)

    Luk'yanov, K.V.

    2008-01-01

    Method of construction of the nucleus-nucleus double folding potential is described. Iteration procedure for the corresponding integral equation is presented. Computer code and numerical results are presented

  12. Computer code for thermal-hydraulic simulation of heat pressurizer tanks operation (Simterm-H)

    International Nuclear Information System (INIS)

    Sellos, R.F.

    1987-01-01

    It is presented the Simtherm-H computer code, developed for calculating the thermodynamic properties of the high pressure heating system and the feedwater tank in transient state for PWR nuclear power plants (1300 MWe). (E.G.) [pt

  13. The Unified English Braille Code: Examination by Science, Mathematics, and Computer Science Technical Expert Braille Readers

    Science.gov (United States)

    Holbrook, M. Cay; MacCuspie, P. Ann

    2010-01-01

    Braille-reading mathematicians, scientists, and computer scientists were asked to examine the usability of the Unified English Braille Code (UEB) for technical materials. They had little knowledge of the code prior to the study. The research included two reading tasks, a short tutorial about UEB, and a focus group. The results indicated that the…

  14. PAD: a one-dimensional, coupled neutronic-thermodynamic-hydrodynamic computer code

    International Nuclear Information System (INIS)

    Peterson, D.M.; Stratton, W.R.; McLaughlin, T.P.

    1976-12-01

    Theoretical and numerical foundations, utilization guide, sample problems, and program listing and glossary are given for the PAD computer code which describes dynamic systems with interactive neutronics, thermodynamics, and hydrodynamics in one-dimensional spherical, cylindrical, and planar geometries. The code has been applied to prompt critical excursions in various fissioning systems (solution, metal, LMFBR, etc.) as well as to nonfissioning systems

  15. A Coding System for Qualitative Studies of the Information-Seeking Process in Computer Science Research

    Science.gov (United States)

    Moral, Cristian; de Antonio, Angelica; Ferre, Xavier; Lara, Graciela

    2015-01-01

    Introduction: In this article we propose a qualitative analysis tool--a coding system--that can support the formalisation of the information-seeking process in a specific field: research in computer science. Method: In order to elaborate the coding system, we have conducted a set of qualitative studies, more specifically a focus group and some…

  16. Computer codes for tasks in the fields of isotope and radiation research

    International Nuclear Information System (INIS)

    Friedrich, K.; Gebhardt, O.

    1978-11-01

    Concise descriptions of computer codes developed for solving problems in the fields of isotope and radiation research at the Zentralinstitut fuer Isotopen- und Strahlenforschung (ZfI) are compiled. In part two the structure of the ZfI program library MABIF is outlined and a complete list of all codes available is given

  17. HADES. A computer code for fast neutron cross section from the Optical Model

    International Nuclear Information System (INIS)

    Guasp, J.; Navarro, C.

    1973-01-01

    A FORTRAN V computer code for UNIVAC 1108/6 using a local Optical Model with spin-orbit interaction is described. The code calculates fast neutron cross sections, angular distribution, and Legendre moments for heavy and intermediate spherical nuclei. It allows for the possibility of automatic variation of potential parameters for experimental data fitting. (Author) 55 refs

  18. FRAP-T1: a computer code for the transient analysis of oxide fuel rods

    International Nuclear Information System (INIS)

    Dearien, J.A.; Miller, R.L.; Hobbins, R.R.; Siefken, L.J.; Baston, V.F.; Coleman, D.R.

    1977-02-01

    FRAP-T is a FORTRAN IV computer code which can be used to solve for the transient response of a light water reactor (LWR) fuel rod during accident transients such as loss-of-coolant accident (LOCA) or a power-cooling-mismatch (PCM). The coupled effects of mechanical, thermal, internal gas, and material property response on the behavior of the fuel rod are considered. FRAP-T is a modular code with each major computational model isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAP-T is coupled to a materials properties subcode (MATPRO) which is used to provide gas, fuel, and cladding properties to the FRAP-T computational subcodes. No material properties need be supplied by the code user. The needed water properties are stored in tables built into the code. Critical heat flux (CHF) and heat transfer correlations for a wide range of coolant conditions are contained in modular subroutines. FRAP-T has been evaluated by making extensive comparisons between predictions of the code and experimental data. Comparison of predicted and experimental results are presented for a range of FRAP-T calculated parameters. The code is presently programmed and running on an IBM-360/75 and a CDC 7600 computer

  19. Developing a coding scheme for detecting usability and fun problems in computer games for young children

    NARCIS (Netherlands)

    Barendregt, W.; Bekker, M.M.

    2006-01-01

    This article describes the development and assessment of a coding scheme for finding both usability and fun problems through observations of young children playing computer games during user tests. The proposed coding scheme is based on an existing list of breakdown indication types of the detailed

  20. LWR-WIMS, a computer code for light water reactor lattice calculations

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1982-06-01

    LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)

  1. Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'

    International Nuclear Information System (INIS)

    Ito, Masahiro; Uwaba, Tomoyuki

    2005-04-01

    JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)

  2. Computing the Feng-Rao distances for codes from order domains

    DEFF Research Database (Denmark)

    Ruano Benito, Diego

    2007-01-01

    We compute the Feng–Rao distance of a code coming from an order domain with a simplicial value semigroup. The main tool is the Apéry set of a semigroup that can be computed using a Gröbner basis.......We compute the Feng–Rao distance of a code coming from an order domain with a simplicial value semigroup. The main tool is the Apéry set of a semigroup that can be computed using a Gröbner basis....

  3. Distribution of absorbed dose in human eye simulated by SRNA-2KG computer code

    International Nuclear Information System (INIS)

    Ilic, R.; Pesic, M.; Pavlovic, R.; Mostacci, D.

    2003-01-01

    Rapidly increasing performances of personal computers and development of codes for proton transport based on Monte Carlo methods will allow, very soon, the introduction of the computer planning proton therapy as a normal activity in regular hospital procedures. A description of SRNA code used for such applications and results of calculated distributions of proton-absorbed dose in human eye are given in this paper. (author)

  4. PEBBLES: A COMPUTER CODE FOR MODELING PACKING, FLOW AND RECIRCULATIONOF PEBBLES IN A PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-10-01

    A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.

  5. Modification in the CITATION computer code: change of microscopic cross sections by zone

    International Nuclear Information System (INIS)

    Yamaguchi, M.; Kosaka, N.

    1983-01-01

    Some modifications done in the CITATION computer code are presented, aiming to calculate the accumulated burnup for each reactor zone in each step of burnup and allow changing the microscopic cross sections for each zone in accordance to the burnup accumulated after each step of burnup. Some input data were put in the computer code. The alterations were tested and the results were compared with and without modifications. (E.G.) [pt

  6. Model of nuclear reactor type VVER-1000/V-320 built by computer code ATHLET-CD

    International Nuclear Information System (INIS)

    Georgiev, Yoto; Filipov, Kalin; Velev, Vladimir

    2014-01-01

    A model of nuclear reactor type VVER-1000 V-320 developed for computer code ATHLET-CD2.1A is presented. Validation of the has been made, in the analysis of the station blackout scenario with LOCA on fourth cold leg is shown. As the calculation has been completed, the results are checked through comparison with the results from the computer codes ATHLET-2.1A, ASTEC-2.1 and RELAP5mod3.2

  7. A restructuring of the CF/EDF packages for the MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S.H.; Kim, K.R.; Kim, D.H.

    2004-01-01

    The CF and EDF packages, which allow the user to define the functions of variables in a database and the usage of an external data file, have been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To restructure the code, the data transferring methods of the current MELCOR code are modified and then partially adopted into the CF/EDF packages. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as pointers are used to define their addresses. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type without pointers leading to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF/EDF packages addressed in this paper includes a module development and subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code and the trends are almost the same to each other. Therefore the similar approach could be extended to the entire code package for code restructuring. It is expected that the code restructuring will accelerate the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

  8. Challenges in the twentieth century and beyond: Computer codes and data

    International Nuclear Information System (INIS)

    Kirk, B.L.

    1995-01-01

    The second half of the twentieth century has seen major changes in computer architecture. From the early fifties to the early seventies, the word open-quotes computerclose quotes demanded reverence, respect, and even fear. Computers, then, were almost open-quotes untouchable.close quotes Computers have become the mainstream of communication on rapidly expanding communication highways. They have become necessities of life. This report describes computer codes and packaging, as well as compilers and operating systems

  9. Plutonium explosive dispersal modeling using the MACCS2 computer code

    International Nuclear Information System (INIS)

    Steele, C.M.; Wald, T.L.; Chanin, D.I.

    1998-01-01

    The purpose of this paper is to derive the necessary parameters to be used to establish a defensible methodology to perform explosive dispersal modeling of respirable plutonium using Gaussian methods. A particular code, MACCS2, has been chosen for this modeling effort due to its application of sophisticated meteorological statistical sampling in accordance with the philosophy of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.145, ''Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants''. A second advantage supporting the selection of the MACCS2 code for modeling purposes is that meteorological data sets are readily available at most Department of Energy (DOE) and NRC sites. This particular MACCS2 modeling effort focuses on the calculation of respirable doses and not ground deposition. Once the necessary parameters for the MACCS2 modeling are developed and presented, the model is benchmarked against empirical test data from the Double Tracks shot of project Roller Coaster (Shreve 1965) and applied to a hypothetical plutonium explosive dispersal scenario. Further modeling with the MACCS2 code is performed to determine a defensible method of treating the effects of building structure interaction on the respirable fraction distribution as a function of height. These results are related to the Clean Slate 2 and Clean Slate 3 bunkered shots of Project Roller Coaster. Lastly a method is presented to determine the peak 99.5% sector doses on an irregular site boundary in the manner specified in NRC Regulatory Guide 1.145 (1983). Parametric analyses are performed on the major analytic assumptions in the MACCS2 model to define the potential errors that are possible in using this methodology

  10. The computer code SEURBNUK/EURDYN. Pt. 2

    International Nuclear Information System (INIS)

    Yerkess, A.; Broadhouse, B.J.; Smith, B.L.

    1987-01-01

    SEURBNUK-2 is a two-dimensional, axisymmetric, Eulerian, finite difference containment code. The numerical procedure adopted in SEURBNUK to solve the hydrodynamic equations is based on the semi-implicit ICE method which itself is an extension of the MAC algorithm. SEURBNUK has a finite difference thin shell treatment for vessels and internal structures of arbitrary shape and includes the effects of the compressibility of the fluid. Fluid flow through porous media and porous structures can also be accommodated. SEURBNUK/EURDYN is an extension of SEURBNUK-2 in which the finite difference thin shell treatment is replaced by a finite element calculation for both thin or thick structures. This has been achieved by coupling the shell elements and axisymmetric triangular elements. Within the code, the equations of motion for the structures are solved quite separately from those for the fluid, and the timestep for the fluid can be an integer multiple of that for the structures. The interaction of the structures with the fluid is then considered as a modification to the coefficients in the pressure equations, the modifications naturally depending on the behaviour of the structures within the fluid cell. The code is limited to dealing with a single fluid, the coolant, and the bubble and the cover gas are treated as cavities of uniform pressure calculated via appropriate pressure-volume-energy relationships. This manual describes the input data specifications needed for the execution of SEURBNUK/EURDYN calculations. After explaining the output facilities information is included to aid users to avoid some common pit-falls

  11. Plutonium explosive dispersal modeling using the MACCS2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Steele, C.M.; Wald, T.L.; Chanin, D.I.

    1998-11-01

    The purpose of this paper is to derive the necessary parameters to be used to establish a defensible methodology to perform explosive dispersal modeling of respirable plutonium using Gaussian methods. A particular code, MACCS2, has been chosen for this modeling effort due to its application of sophisticated meteorological statistical sampling in accordance with the philosophy of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.145, ``Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants``. A second advantage supporting the selection of the MACCS2 code for modeling purposes is that meteorological data sets are readily available at most Department of Energy (DOE) and NRC sites. This particular MACCS2 modeling effort focuses on the calculation of respirable doses and not ground deposition. Once the necessary parameters for the MACCS2 modeling are developed and presented, the model is benchmarked against empirical test data from the Double Tracks shot of project Roller Coaster (Shreve 1965) and applied to a hypothetical plutonium explosive dispersal scenario. Further modeling with the MACCS2 code is performed to determine a defensible method of treating the effects of building structure interaction on the respirable fraction distribution as a function of height. These results are related to the Clean Slate 2 and Clean Slate 3 bunkered shots of Project Roller Coaster. Lastly a method is presented to determine the peak 99.5% sector doses on an irregular site boundary in the manner specified in NRC Regulatory Guide 1.145 (1983). Parametric analyses are performed on the major analytic assumptions in the MACCS2 model to define the potential errors that are possible in using this methodology.

  12. Implementation of burnup in FERM nodal computer code

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Nakata, H.

    1986-01-01

    In this work a spatial burnup scheme and feedback effects has been implemented into the FERM [1] ('Finite Element Response Matrix') program. The spatially dependent neutronic parameters have been considered in three levels: zonewise calculation, assemblywise calculation and pointwise calculation. The results have been compared with the results obtained by CITATION [2] program and showed that the processing time in the FERM code has been hundred of times shorter and no significant difference has been observed in the assembly average power distribution. (Author) [pt

  13. Status Report on Hydrogen Management and Related Computer Codes

    International Nuclear Information System (INIS)

    Liang, Z.; Chan, C.K.; Sonnenkalb, M.; Bentaib, A.; Malet, J.; Sangiorgi, M.; Gryffroy, D.; Gyepi-Garbrah, S.; Duspiva, J.; Sevon, T.; Kelm, S.; Reinecke, E.A.; Xu, Z.J.; Cervone, A.; Utsuno, H.; Hotta, A.; Hong, S.W.; Kim, J.T.; Visser, D.C.; Stempniewicz, M.M.; Kuriene, L.; Prusinski, P.; Martin-Valdepenas, J.M.; Frid, W.; Isaksson, P.; Dreier, J.; Paladino, D.; Algama, D.; Notafrancesco, A.; Amri, A.; Kissane, M.; )

    2014-01-01

    In follow-up to the Fukushima Daiichi NPP accident, the Committee on the Safety of Nuclear Installations (CSNI) decided to launch several high priority activities. At the 14. plenary meeting of the Working Group on Analysis and Management of Accidents (WGAMA), a proposal for a status paper on hydrogen generation, transport and mitigation under severe accident conditions was approved. The proposed activity is in line with the WGAMA mandate and it was considered to be needed to revisit the hydrogen issue. The report is broken down into five Chapters and two appendixes. Chapter 1 provides background information for this activity and expected topics defined by the WGAMA members. A general understanding of hydrogen behavior and control in severe accidents is discussed. A brief literature review is included in this chapter to summarize the progress obtained from the early US NRC sponsored research on hydrogen and recent international OECD or EC sponsored projects on hydrogen related topics (generation, distribution, combustion and mitigation). Chapter 2 provides a general overview of the various reactor designs of Western PWRs, BWRs, Eastern European VVERs and PHWRs (CANDUs). The purpose is to understand the containment design features in relation to hydrogen management measures. Chapter 3 provides a detailed description of national requirements on hydrogen management and hydrogen mitigation measures inside the containment and other places (e.g., annulus space, secondary buildings, spent fuel pool, etc.). Discussions are followed on hydrogen analysis approaches, application of safety systems (e.g., spray, containment ventilation, local air cooler, suppression pool, and latch systems), hydrogen measurement strategies as well as lessons learnt from the Fukushima Daiichi nuclear power accident. Chapter 4 provides an overview of various codes that are being used for hydrogen risk assessment, and the codes capabilities and validation status in terms of hydrogen related

  14. Computational fluid mechanics qualification calculations for the code TEACH

    International Nuclear Information System (INIS)

    DeGrazia, M.C.; Fitzsimmons, L.B.; Reynolds, J.T.

    1979-11-01

    KAPL is developing a predictive method for three-dimensional (3-D) turbulent fluid flow configurations typically encountered in the thermal-hydraulic design of a nuclear reactor. A series of experiments has been selected for analysis to investigate the adequacy of the two-equation turbulence model developed at Imperial College, London, England for predicting the flow patterns in simple geometries. The analysis of these experiments is described with the two-dimensional (2-D) turbulent fluid flow code TEACH. This work qualifies TEACH for a variety of geometries and flow conditions

  15. Verification of the network flow and transport/distributed velocity (NWFT/DVM) computer code

    International Nuclear Information System (INIS)

    Duda, L.E.

    1984-05-01

    The Network Flow and Transport/Distributed Velocity Method (NWFT/DVM) computer code was developed primarily to fulfill a need for a computationally efficient ground-water flow and contaminant transport capability for use in risk analyses where, quite frequently, large numbers of calculations are required. It is a semi-analytic, quasi-two-dimensional network code that simulates ground-water flow and the transport of dissolved species (radionuclides) in a saturated porous medium. The development of this code was carried out under a program funded by the US Nuclear Regulatory Commission (NRC) to develop a methodology for assessing the risk from disposal of radioactive wastes in deep geologic formations (FIN: A-1192 and A-1266). In support to the methodology development program, the NRC has funded a separate Maintenance of Computer Programs Project (FIN: A-1166) to ensure that the codes developed under A-1192 or A-1266 remain consistent with current operating systems, are as error-free as possible, and have up-to-date documentations for reference by the NRC staff. Part of this effort would include verification and validation tests to assure that a code correctly performs the operations specified and/or is representing the processes or system for which it is intended. This document contains four verification problems for the NWFT/DVM computer code. Two of these problems are analytical verifications of NWFT/DVM where results are compared to analytical solutions. The other two are code-to-code verifications where results from NWFT/DVM are compared to those of another computer code. In all cases NWFT/DVM showed good agreement with both the analytical solutions and the results from the other code

  16. Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2001-01-01

    This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

  17. The failure mechanisms of HTR coated particle fuel and computer code

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Shao Youlin; Liang Tongxiang; Tang Chunhe

    2010-01-01

    The basic constituent unit of fuel element in HTR is ceramic coated particle fuel. And the performance of coated particle fuel determines the safety of HTR. In addition to the traditional detection of radiation experiments, establishing computer code is of great significance to the research. This paper mainly introduces the structure and the failure mechanism of TRISO-coated particle fuel, as well as a few basic assumptions,principles and characteristics of some existed main overseas codes. Meanwhile, this paper has proposed direction of future research by comparing the advantages and disadvantages of several computer codes. (authors)

  18. Status of computer codes available in AEOI for reactor physics analysis

    International Nuclear Information System (INIS)

    Karbassiafshar, M.

    1986-01-01

    Many of the nuclear computer codes available in Atomic Energy Organization of Iran AEOI can be used for physics analysis of an operating reactor or design purposes. Grasp of the various methods involved and practical experience with these codes would be the starting point for interesting design studies or analysis of operating conditions of presently existing and future reactors. A review of the objectives and flowchart of commonly practiced procedures in reactor physics analysis of LWRs and related computer codes was made, extrapolating to the nationally and internationally available resources. Finally, effective utilization of the existing facilities is discussed and called upon

  19. Development of a system of computer codes for severe accident analyses and its applications

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.

  20. TPASS: a gamma-ray spectrum analysis and isotope identification computer code

    International Nuclear Information System (INIS)

    Dickens, J.K.

    1981-03-01

    The gamma-ray spectral data-reduction and analysis computer code TPASS is described. This computer code is used to analyze complex Ge(Li) gamma-ray spectra to obtain peak areas corrected for detector efficiencies, from which are determined gamma-ray yields. These yields are compared with an isotope gamma-ray data file to determine the contributions to the observed spectrum from decay of specific radionuclides. A complete FORTRAN listing of the code and a complex test case are given

  1. Development of a system of computer codes for severe accident analyses and its applications

    International Nuclear Information System (INIS)

    Chang, Soon Hong; Cheon, Moon Heon; Cho, Nam jin; No, Hui Cheon; Chang, Hyeon Seop; Moon, Sang Kee; Park, Seok Jeong; Chung, Jee Hwan

    1991-12-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy

  2. CITADEL: a computer code for the analysis of iodine behavior in steam generator tube rupture accidents

    International Nuclear Information System (INIS)

    1982-04-01

    The computer code CITADEL was written to analyze iodine behavior during steam generator tube rupture accidents. The code models the transport and deposition of iodine from its point of escape at the steam generator primary break until its release to the environment. This report provides a brief description of the code including its input requirements and the nature and form of its output. A user's guide describing the manner in which the input data are required to be set up to run the code is also provided

  3. Validation of computer codes used in the safety analysis of Canadian research reactors

    International Nuclear Information System (INIS)

    Bishop, W.E.; Lee, A.G.

    1998-01-01

    AECL has embarked on a validation program for the suite of computer codes that it uses in performing the safety analyses for its research reactors. Current focus is on codes used for the analysis of the two MAPLE reactors under construction at Chalk River but the program will be extended to include additional codes that will be used for the Irradiation Research Facility. The program structure is similar to that used for the validation of codes used in the safety analyses for CANDU power reactors. (author)

  4. Development of DUST: A computer code that calculates release rates from a LLW disposal unit

    International Nuclear Information System (INIS)

    Sullivan, T.M.

    1992-01-01

    Performance assessment of a Low-Level Waste (LLW) disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the disposal unit source term). The major physical processes that influence the source term are water flow, container degradation, waste form leaching, and radionuclide transport. A computer code, DUST (Disposal Unit Source Term) has been developed which incorporates these processes in a unified manner. The DUST code improves upon existing codes as it has the capability to model multiple container failure times, multiple waste form release properties, and radionuclide specific transport properties. Verification studies performed on the code are discussed

  5. Proceedings of the conference on computer codes and the linear accelerator community

    International Nuclear Information System (INIS)

    Cooper, R.K.

    1990-07-01

    The conference whose proceedings you are reading was envisioned as the second in a series, the first having been held in San Diego in January 1988. The intended participants were those people who are actively involved in writing and applying computer codes for the solution of problems related to the design and construction of linear accelerators. The first conference reviewed many of the codes both extant and under development. This second conference provided an opportunity to update the status of those codes, and to provide a forum in which emerging new 3D codes could be described and discussed. The afternoon poster session on the second day of the conference provided an opportunity for extended discussion. All in all, this conference was felt to be quite a useful interchange of ideas and developments in the field of 3D calculations, parallel computation, higher-order optics calculations, and code documentation and maintenance for the linear accelerator community. A third conference is planned

  6. Proceedings of the conference on computer codes and the linear accelerator community

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, R.K. (comp.)

    1990-07-01

    The conference whose proceedings you are reading was envisioned as the second in a series, the first having been held in San Diego in January 1988. The intended participants were those people who are actively involved in writing and applying computer codes for the solution of problems related to the design and construction of linear accelerators. The first conference reviewed many of the codes both extant and under development. This second conference provided an opportunity to update the status of those codes, and to provide a forum in which emerging new 3D codes could be described and discussed. The afternoon poster session on the second day of the conference provided an opportunity for extended discussion. All in all, this conference was felt to be quite a useful interchange of ideas and developments in the field of 3D calculations, parallel computation, higher-order optics calculations, and code documentation and maintenance for the linear accelerator community. A third conference is planned.

  7. Visualization of elastic wavefields computed with a finite difference code

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, S. [Lawrence Livermore National Lab., CA (United States); Harris, D.

    1994-11-15

    The authors have developed a finite difference elastic propagation model to simulate seismic wave propagation through geophysically complex regions. To facilitate debugging and to assist seismologists in interpreting the seismograms generated by the code, they have developed an X Windows interface that permits viewing of successive temporal snapshots of the (2D) wavefield as they are calculated. The authors present a brief video displaying the generation of seismic waves by an explosive source on a continent, which propagate to the edge of the continent then convert to two types of acoustic waves. This sample calculation was part of an effort to study the potential of offshore hydroacoustic systems to monitor seismic events occurring onshore.

  8. Computer code for general analysis of radon risks (GARR)

    International Nuclear Information System (INIS)

    Ginevan, M.

    1984-09-01

    This document presents a computer model for general analysis of radon risks that allow the user to specify a large number of possible models with a small number of simple commands. The model is written in a version of BASIC which conforms closely to the American National Standards Institute (ANSI) definition for minimal BASIC and thus is readily modified for use on a wide variety of computers and, in particular, microcomputers. Model capabilities include generation of single-year life tables from 5-year abridged data, calculation of multiple-decrement life tables for lung cancer for the general population, smokers, and nonsmokers, and a cohort lung cancer risk calculation that allows specification of level and duration of radon exposure, the form of the risk model, and the specific population assumed at risk. 36 references, 8 figures, 7 tables

  9. Spent fuel management fee methodology and computer code user's manual

    International Nuclear Information System (INIS)

    Engel, R.L.; White, M.K.

    1982-01-01

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively

  10. High-Performance Java Codes for Computational Fluid Dynamics

    Science.gov (United States)

    Riley, Christopher; Chatterjee, Siddhartha; Biswas, Rupak; Biegel, Bryan (Technical Monitor)

    2001-01-01

    The computational science community is reluctant to write large-scale computationally -intensive applications in Java due to concerns over Java's poor performance, despite the claimed software engineering advantages of its object-oriented features. Naive Java implementations of numerical algorithms can perform poorly compared to corresponding Fortran or C implementations. To achieve high performance, Java applications must be designed with good performance as a primary goal. This paper presents the object-oriented design and implementation of two real-world applications from the field of Computational Fluid Dynamics (CFD): a finite-volume fluid flow solver (LAURA, from NASA Langley Research Center), and an unstructured mesh adaptation algorithm (2D_TAG, from NASA Ames Research Center). This work builds on our previous experience with the design of high-performance numerical libraries in Java. We examine the performance of the applications using the currently available Java infrastructure and show that the Java version of the flow solver LAURA performs almost within a factor of 2 of the original procedural version. Our Java version of the mesh adaptation algorithm 2D_TAG performs within a factor of 1.5 of its original procedural version on certain platforms. Our results demonstrate that object-oriented software design principles are not necessarily inimical to high performance.

  11. Characterization of the MCNPX computer code in micro processed architectures

    International Nuclear Information System (INIS)

    Almeida, Helder C.; Dominguez, Dany S.; Orellana, Esbel T.V.; Milian, Felix M.

    2009-01-01

    The MCNPX (Monte Carlo N-Particle extended) can be used to simulate the transport of several types of nuclear particles, using probabilistic methods. The technique used for MCNPX is to follow the history of each particle from its origin to its extinction that can be given by absorption, escape or other reasons. To obtain accurate results in simulations performed with the MCNPX is necessary to process a large number of histories, which demand high computational cost. Currently the MCNPX can be installed in virtually all computing platforms available, however there is virtually no information on the performance of the application in each. This paper studies the performance of MCNPX, to work with electrons and photons in phantom Faux on two platforms used by most researchers, Windows and Li nux. Both platforms were tested on the same computer to ensure the reliability of the hardware in the measures of performance. The performance of MCNPX was measured by time spent to run a simulation, making the variable time the main measure of comparison. During the tests the difference in performance between the two platforms MCNPX was evident. In some cases we were able to gain speed more than 10% only with the exchange platforms, without any specific optimization. This shows the relevance of the study to optimize this tool on the platform most appropriate for its use. (author)

  12. Monte Carlo simulation of Ising models by multispin coding on a vector computer

    Science.gov (United States)

    Wansleben, Stephan; Zabolitzky, John G.; Kalle, Claus

    1984-11-01

    Rebbi's efficient multispin coding algorithm for Ising models is combined with the use of the vector computer CDC Cyber 205. A speed of 21.2 million updates per second is reached. This is comparable to that obtained by special- purpose computers.

  13. User's manual for the G.T.M.-1 computer code

    International Nuclear Information System (INIS)

    Prado-Herrero, P.

    1992-01-01

    This document describes the GTM-1 ( Geosphere Transport Model, release-1) computer code and is intended to provide the reader with enough detailed information in order to use the code. GTM-1 was developed for the assessment of radionuclide migration by the ground water through geologic deposits whose properties can change along the pathway.GTM-1 solves the transport equation by the finite differences method ( Crank-Nicolson scheme ). It was developped for specific use within Probabilistic System Assessment (PSA) Monte Carlo Method codes; in this context the first application of GTM-1 was within the LISA (Long Term Isolation System Assessment) code. GTM-1 is also available as an independent model, which includes various submodels simulating a multi-barrier disposal system. The code has been tested with the PSACOIN ( Probabilistic System Assessment Codes intercomparison) benchmarks exercises from PSAC User Group (OECD/NEA). 10 refs., 6 Annex., 2 tabs

  14. Benchmark testing and independent verification of the VS2DT computer code

    International Nuclear Information System (INIS)

    McCord, J.T.

    1994-11-01

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

  15. ASAS: Computational code for Analysis and Simulation of Atomic Spectra

    Directory of Open Access Journals (Sweden)

    Jhonatha R. dos Santos

    2017-01-01

    Full Text Available The laser isotopic separation process is based on the selective photoionization principle and, because of this, it is necessary to know the absorption spectrum of the desired atom. Computational resource has become indispensable for the planning of experiments and analysis of the acquired data. The ASAS (Analysis and Simulation of Atomic Spectra software presented here is a helpful tool to be used in studies involving atomic spectroscopy. The input for the simulations is friendly and essentially needs a database containing the energy levels and spectral lines of the atoms subjected to be studied.

  16. GIANT: a computer code for General Interactive ANalysis of Trajectories

    International Nuclear Information System (INIS)

    Jaeger, J.; Lee, M.; Servranckx, R.; Shoaee, H.

    1985-04-01

    Many model-driven diagnostic and correction procedures have been developed at SLAC for the on-line computer controlled operation of SPEAR, PEP, the LINAC, and the Electron Damping Ring. In order to facilitate future applications and enhancements, these procedures are being collected into a single program, GIANT. The program allows interactive diagnosis as well as performance optimization of any beam transport line or circular machine. The test systems for GIANT are those of the SLC project. The organization of this program and some of the recent applications of the procedures will be described in this paper

  17. The Uncertainty Test for the MAAP Computer Code

    International Nuclear Information System (INIS)

    Park, S. H.; Song, Y. M.; Park, S. Y.; Ahn, K. I.; Kim, K. R.; Lee, Y. J.

    2008-01-01

    After the Three Mile Island Unit 2 (TMI-2) and Chernobyl accidents, safety issues for a severe accident are treated in various aspects. Major issues in our research part include a level 2 PSA. The difficulty in expanding the level 2 PSA as a risk information activity is the uncertainty. In former days, it attached a weight to improve the quality in a internal accident PSA, but the effort is insufficient for decrease the phenomenon uncertainty in the level 2 PSA. In our country, the uncertainty degree is high in the case of a level 2 PSA model, and it is necessary to secure a model to decrease the uncertainty. We have not yet experienced the uncertainty assessment technology, the assessment system itself depends on advanced nations. In advanced nations, the severe accident simulator is implemented in the hardware level. But in our case, basic function in a software level can be implemented. In these circumstance at home and abroad, similar instances are surveyed such as UQM and MELCOR. Referred to these instances, SAUNA (Severe Accident UNcertainty Analysis) system is being developed in our project to assess and decrease the uncertainty in a level 2 PSA. It selects the MAAP code to analyze the uncertainty in a severe accident

  18. The modification and application of RAMS computer code. Final report

    International Nuclear Information System (INIS)

    McKee, T.B.

    1995-01-01

    The Regional Atmospheric Modeling System (RAMS) has been utilized in its most updated form, version 3a, to simulate a case night from the Atmospheric Studies in COmplex Terrain (ASCOT) experimental program. ASCOT held a wintertime observational campaign during February, 1991 to observe the often strong drainage flows which form on the Great Plains and in the canyons embedded within the slope from the Continental Divide to the Great Plains. A high resolution (500 m grid spacing) simulation of the 4-5 February 1991 case night using the more advanced turbulence closure now available in RAMS 3a allowed greater analysis of the physical processes governing the drainage flows. It is found that shear interaction above and within the drainage flow are important, and are overpredicted with the new scheme at small grid spacing (< ∼1000 m). The implication is that contaminants trapped in nighttime stable flows such as these, will be mixed too strongly in the vertical reducing predicted ground concentrations. The HYPACT code has been added to the capability at LANL, although due to the reduced scope of work, no simulations with HYPACT were performed

  19. Automatic Parallelization Tool: Classification of Program Code for Parallel Computing

    Directory of Open Access Journals (Sweden)

    Mustafa Basthikodi

    2016-04-01

    Full Text Available Performance growth of single-core processors has come to a halt in the past decade, but was re-enabled by the introduction of parallelism in processors. Multicore frameworks along with Graphical Processing Units empowered to enhance parallelism broadly. Couples of compilers are updated to developing challenges forsynchronization and threading issues. Appropriate program and algorithm classifications will have advantage to a great extent to the group of software engineers to get opportunities for effective parallelization. In present work we investigated current species for classification of algorithms, in that related work on classification is discussed along with the comparison of issues that challenges the classification. The set of algorithms are chosen which matches the structure with different issues and perform given task. We have tested these algorithms utilizing existing automatic species extraction toolsalong with Bones compiler. We have added functionalities to existing tool, providing a more detailed characterization. The contributions of our work include support for pointer arithmetic, conditional and incremental statements, user defined types, constants and mathematical functions. With this, we can retain significant data which is not captured by original speciesof algorithms. We executed new theories into the device, empowering automatic characterization of program code.

  20. The extensive international use of commercial computational fluid dynamics (CFD) codes

    International Nuclear Information System (INIS)

    Hartmut Wider

    2005-01-01

    What are the main reasons for the extensive international success of commercial CFD codes? This is due to their ability to calculate the fine structures of the investigated processes due to their versatility, their numerical stability and that they can guarantee the proper solution in most cases. This was made possible by the constantly increasing computer power at an ever more affordable prize. Furthermore it is much more efficient to have researchers use a CFD code rather than to develop a similar code system due to the time consuming nature of this activity and the high probability of hidden coding errors. The centralized development and upgrading makes these reliable CFD codes possible and affordable. However, the CFD companies' developments are naturally concentrated on the most profitable areas, and thus, if one works in a 'non-priority' field one cannot use them. Moreover, the prize of renting CFD codes, applications to complex systems such as whole nuclear reactors and the need to teach students gives the development of self-made codes still plenty of room. But CFD codes can model detailed aspects of large systems and subroutines generated by users can be added. Since there are only a few heavily used CFD codes such as FLUENT, STAR-CD, ANSYS CFX, these are used in many countries. Also international training courses are given and the news bulletins of these codes help to spread the news on further developments. A larger number of international codes would increase the competition but would at the same time make it harder to select the most appropriate CFD code for a given problem. Examples will be presented of uses of CFD codes as more detailed system codes for the decay heat removal from reactors, the application to aerosol physics and the application to heavy metal fluids using different turbulence models. (author)

  1. Verification of thermal-hydraulic computer codes against standard problems for WWER reflooding

    International Nuclear Information System (INIS)

    Alexander D Efanov; Vladimir N Vinogradov; Victor V Sergeev; Oleg A Sudnitsyn

    2005-01-01

    Full text of publication follows: The computational assessment of reactor core components behavior under accident conditions is impossible without knowledge of the thermal-hydraulic processes occurring in this case. The adequacy of the results obtained using the computer codes to the real processes is verified by carrying out a number of standard problems. In 2000-2003, the fulfillment of three Russian standard problems on WWER core reflooding was arranged using the experiments on full-height electrically heated WWER 37-rod bundle model cooldown in regimes of bottom (SP-1), top (SP-2) and combined (SP-3) reflooding. The representatives from the eight MINATOM's organizations took part in this work, in the course of which the 'blind' and posttest calculations were performed using various versions of the RELAP5, ATHLET, CATHARE, COBRA-TF, TRAP, KORSAR computer codes. The paper presents a brief description of the test facility, test section, test scenarios and conditions as well as the basic results of computational analysis of the experiments. The analysis of the test data revealed a significantly non-one-dimensional nature of cooldown and rewetting of heater rods heated up to a high temperature in a model bundle. This was most pronounced at top and combined reflooding. The verification of the model reflooding computer codes showed that most of computer codes fairly predict the peak rod temperature and the time of bundle cooldown. The exception is provided by the results of calculations with the ATHLET and CATHARE codes. The nature and rate of rewetting front advance in the lower half of the bundle are fairly predicted practically by all computer codes. The disagreement between the calculations and experimental results for the upper half of the bundle is caused by the difficulties of computational simulation of multidimensional effects by 1-D computer codes. In this regard, a quasi-two-dimensional computer code COBRA-TF offers certain advantages. Overall, the closest

  2. Issues in computational fluid dynamics code verification and validation

    Energy Technology Data Exchange (ETDEWEB)

    Oberkampf, W.L.; Blottner, F.G.

    1997-09-01

    A broad range of mathematical modeling errors of fluid flow physics and numerical approximation errors are addressed in computational fluid dynamics (CFD). It is strongly believed that if CFD is to have a major impact on the design of engineering hardware and flight systems, the level of confidence in complex simulations must substantially improve. To better understand the present limitations of CFD simulations, a wide variety of physical modeling, discretization, and solution errors are identified and discussed. Here, discretization and solution errors refer to all errors caused by conversion of the original partial differential, or integral, conservation equations representing the physical process, to algebraic equations and their solution on a computer. The impact of boundary conditions on the solution of the partial differential equations and their discrete representation will also be discussed. Throughout the article, clear distinctions are made between the analytical mathematical models of fluid dynamics and the numerical models. Lax`s Equivalence Theorem and its frailties in practical CFD solutions are pointed out. Distinctions are also made between the existence and uniqueness of solutions to the partial differential equations as opposed to the discrete equations. Two techniques are briefly discussed for the detection and quantification of certain types of discretization and grid resolution errors.

  3. Toward Reproducible Computational Research: An Empirical Analysis of Data and Code Policy Adoption by Journals.

    Directory of Open Access Journals (Sweden)

    Victoria Stodden

    Full Text Available Journal policy on research data and code availability is an important part of the ongoing shift toward publishing reproducible computational science. This article extends the literature by studying journal data sharing policies by year (for both 2011 and 2012 for a referent set of 170 journals. We make a further contribution by evaluating code sharing policies, supplemental materials policies, and open access status for these 170 journals for each of 2011 and 2012. We build a predictive model of open data and code policy adoption as a function of impact factor and publisher and find higher impact journals more likely to have open data and code policies and scientific societies more likely to have open data and code policies than commercial publishers. We also find open data policies tend to lead open code policies, and we find no relationship between open data and code policies and either supplemental material policies or open access journal status. Of the journals in this study, 38% had a data policy, 22% had a code policy, and 66% had a supplemental materials policy as of June 2012. This reflects a striking one year increase of 16% in the number of data policies, a 30% increase in code policies, and a 7% increase in the number of supplemental materials policies. We introduce a new dataset to the community that categorizes data and code sharing, supplemental materials, and open access policies in 2011 and 2012 for these 170 journals.

  4. Toward Reproducible Computational Research: An Empirical Analysis of Data and Code Policy Adoption by Journals.

    Science.gov (United States)

    Stodden, Victoria; Guo, Peixuan; Ma, Zhaokun

    2013-01-01

    Journal policy on research data and code availability is an important part of the ongoing shift toward publishing reproducible computational science. This article extends the literature by studying journal data sharing policies by year (for both 2011 and 2012) for a referent set of 170 journals. We make a further contribution by evaluating code sharing policies, supplemental materials policies, and open access status for these 170 journals for each of 2011 and 2012. We build a predictive model of open data and code policy adoption as a function of impact factor and publisher and find higher impact journals more likely to have open data and code policies and scientific societies more likely to have open data and code policies than commercial publishers. We also find open data policies tend to lead open code policies, and we find no relationship between open data and code policies and either supplemental material policies or open access journal status. Of the journals in this study, 38% had a data policy, 22% had a code policy, and 66% had a supplemental materials policy as of June 2012. This reflects a striking one year increase of 16% in the number of data policies, a 30% increase in code policies, and a 7% increase in the number of supplemental materials policies. We introduce a new dataset to the community that categorizes data and code sharing, supplemental materials, and open access policies in 2011 and 2012 for these 170 journals.

  5. A restructuring of the MELCOR fission product packages for the MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S.H.; Kim, K.R.; Kim, D.H.

    2004-01-01

    The RN1/RN2 packages, which are the fission product-related packages in MELCOR, have been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the RN1/RN2 package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1/RN2 package addressed in this paper includes a module development, subroutine modification, and the treatment of MELGEN, which generates the data file, as well as MELCOR, which is processing the calculation. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerate the code domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

  6. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    International Nuclear Information System (INIS)

    Wren, D.J.; Popov, N.; Snell, V.G.

    2004-01-01

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  7. Superimposed Code Theoretic Analysis of Deoxyribonucleic Acid (DNA) Codes and DNA Computing

    Science.gov (United States)

    2010-01-01

    DNA strand and its Watson - Crick complement can be used to perform mathematical computation. This research addresses how the...Acid dsDNA double stranded DNA MOSAIC Mobile Stream Processing Cluster PCR Polymerase Chain Reaction RAM Random Access Memory ssDNA single stranded DNA WC Watson – Crick A Adenine C Cytosine G Guanine T Thymine ...are 5′→3′ and strands with strikethrough are 3′→5′. A dsDNA duplex formed between a strand and its reverse complement is called a

  8. Algorithms and computer codes for atomic and molecular quantum scattering theory

    International Nuclear Information System (INIS)

    Thomas, L.

    1979-01-01

    This workshop has succeeded in bringing up 11 different coupled equation codes on the NRCC computer, testing them against a set of 24 different test problems and making them available to the user community. These codes span a wide variety of methodologies, and factors of up to 300 were observed in the spread of computer times on specific problems. A very effective method was devised for examining the performance of the individual codes in the different regions of the integration range. Many of the strengths and weaknesses of the codes have been identified. Based on these observations, a hybrid code has been developed which is significantly superior to any single code tested. Thus, not only have the original goals been fully met, the workshop has resulted directly in an advancement of the field. All of the computer programs except VIVS are available upon request from the NRCC. Since an improved version of VIVS is contained in the hybrid program, VIVAS, it was not made available for distribution. The individual program LOGD is, however, available. In addition, programs which compute the potential energy matrices of the test problems are also available. The software library names for Tests 1, 2 and 4 are HEH2, LICO, and EN2, respectively

  9. Development of computing code system for level 3 PSA

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan.

    1997-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic

  10. Development of computing code system for level 3 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan

    1997-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic

  11. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  12. UCODE, a computer code for universal inverse modeling

    Science.gov (United States)

    Poeter, E.P.; Hill, M.C.

    1999-01-01

    This article presents the US Geological Survey computer program UCODE, which was developed in collaboration with the US Army Corps of Engineers Waterways Experiment Station and the International Ground Water Modeling Center of the Colorado School of Mines. UCODE performs inverse modeling, posed as a parameter-estimation problem, using nonlinear regression. Any application model or set of models can be used; the only requirement is that they have numerical (ASCII or text only) input and output files and that the numbers in these files have sufficient significant digits. Application models can include preprocessors and postprocessors as well as models related to the processes of interest (physical, chemical and so on), making UCODE extremely powerful for model calibration. Estimated parameters can be defined flexibly with user-specified functions. Observations to be matched in the regression can be any quantity for which a simulated equivalent value can be produced, thus simulated equivalent values are calculated using values that appear in the application model output files and can be manipulated with additive and multiplicative functions, if necessary. Prior, or direct, information on estimated parameters also can be included in the regression. The nonlinear regression problem is solved by minimizing a weighted least-squares objective function with respect to the parameter values using a modified Gauss-Newton method. Sensitivities needed for the method are calculated approximately by forward or central differences and problems and solutions related to this approximation are discussed. Statistics are calculated and printed for use in (1) diagnosing inadequate data or identifying parameters that probably cannot be estimated with the available data, (2) evaluating estimated parameter values, (3) evaluating the model representation of the actual processes and (4) quantifying the uncertainty of model simulated values. UCODE is intended for use on any computer operating

  13. FIRAC - a computer code to predict fire accident effects in nuclear facilities

    International Nuclear Information System (INIS)

    Bolstad, J.W.; Foster, R.D.; Gregory, W.S.

    1983-01-01

    FIRAC is a medium-sized computer code designed to predict fire-induced flows, temperatures, and material transport within the ventilating systems and other airflow pathways in nuclear-related facilities. The code is designed to analyze the behavior of interconnected networks of rooms and typical ventilation system components. This code is one in a family of computer codes that is designed to provide improved methods of safety analysis for the nuclear industry. The structure of this code closely follows that of the previously developed TVENT and EVENT codes. Because a lumped-parameter formulation is used, this code is particularly suitable for calculating the effects of fires in the far field (that is, in regions removed from the fire compartment), where the fire may be represented parametrically. However, a fire compartment model to simulate conditions in the enclosure is included. This model provides transport source terms to the ventilation system that can affect its operation and in turn affect the fire. A basic material transport capability that features the effects of convection, deposition, entrainment, and filtration of material is included. The interrelated effects of filter plugging, heat transfer, gas dynamics, and material transport are taken into account. In this paper the authors summarize the physical models used to describe the gas dynamics, material transport, and heat transfer processes. They also illustrate how a typical facility is modeled using the code

  14. Computer-aided software understanding systems to enhance confidence of scientific codes

    International Nuclear Information System (INIS)

    Sheng, G.; Oeren, T.I.

    1991-01-01

    A unique characteristic of nuclear waste disposal is the very long time span over which the combined engineered and natural containment system must remain effective: hundreds of thousands of years. Since there is no precedent in human history for such an endeavour, simulation with the use of computers is the only means we have of forecasting possible future outcomes quantitatively. The need for reliable models and software to make such forecasts so far into the future is obvious. One of the critical elements necessary to ensure reliability is the degree of reviewability of the computer program. Among others, there are two very important reasons for this. Firstly, if there is to be any chance at all of validating the conceptual models as implemented by the computer code, peer reviewers must be able to see and understand what the program is doing. It is all but impossible to achieve this understanding by just looking at the code due to possible unfamiliarity with the language and often due as well to the length and complexity of the code. Secondly, a thorough understanding of the code is also necessary to carry out code maintenance activities which include among others, error detection, error correction and code modification for purposes of enhancing its performance, functionality or to adapt it to a changed environment. The emerging concepts of computer-aided software understanding and reverse engineering can answer precisely these needs. This paper will discuss the role they can play in enhancing the confidence one has on computer codes and several examples will be provided. Finally a brief discussion of combining state-of-art forward engineering systems with reverse engineering systems will show how powerfully they can contribute to the overall quality assurance of a computer program. (13 refs., 7 figs.)

  15. Assessment of a pressurizer spray valve faulty opening transient at Asco Nuclear Power Plant with RELAP5/MOD2

    International Nuclear Information System (INIS)

    Reventos, F.; Baptista, J.S.; Navas, A.P.; Moreno, P.

    1993-12-01

    The Asociacion Nuclear Asco has prepared a model of Asco NPP using RELAP5/MOD2. This model, which include thermalhydraulics, kinetics and protection and controls, has been qualified in previous calculations of several actual plant transients. One of the transients of the qualification process is a ''Pressurizer spray valve faulty opening'' presented in this report. It consists in a primary coolant depressurization that causes the reactor trip by overtemperature and later on the actuation of the safety injection. The results are in close agreement with plant data

  16. Computer codes in nuclear safety, radiation transport and dosimetry; Les codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    Bordy, J M; Kodeli, I; Menard, St; Bouchet, J L; Renard, F; Martin, E; Blazy, L; Voros, S; Bochud, F; Laedermann, J P; Beaugelin, K; Makovicka, L; Quiot, A; Vermeersch, F; Roche, H; Perrin, M C; Laye, F; Bardies, M; Struelens, L; Vanhavere, F; Gschwind, R; Fernandez, F; Quesne, B; Fritsch, P; Lamart, St; Crovisier, Ph; Leservot, A; Antoni, R; Huet, Ch; Thiam, Ch; Donadille, L; Monfort, M; Diop, Ch; Ricard, M

    2006-07-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.

  17. Development of the computer code to monitor gamma radiation in the nuclear facility environment

    International Nuclear Information System (INIS)

    Akhmad, Y. R.; Pudjiyanto, M.S.

    1998-01-01

    Computer codes for gamma radiation monitoring in the vicinity of nuclear facility which have been developed could be introduced to the commercial potable gamma analyzer. The crucial stage of the first year activity was succeeded ; that is the codes have been tested to transfer data file (pulse high distribution) from Micro NOMAD gamma spectrometer (ORTEC product) and the convert them into dosimetry and physics quantities. Those computer codes are called as GABATAN (Gamma Analyzer of Batan) and NAGABAT (Natural Gamma Analyzer of Batan). GABATAN code can isable to used at various nuclear facilities for analyzing gamma field up to 9 MeV, while NAGABAT could be used for analyzing the contribution of natural gamma rays to the exposure rate in the certain location

  18. A computer code for Cohort Analysis of Increased Risks of Death (CAIRD). Technical report

    International Nuclear Information System (INIS)

    Cook, J.R.; Bunger, B.M.; Barrick, M.K.

    1978-06-01

    The most serious health risk confronting individuals exposed to radiation is death from an induced cancer. Since cancers usually do no develop until many years after exposure, other causes of death may intervene and take the lives of those destined to die from cancer. This computer code has been developed to aid risk analysis by calculating the number of premature deaths and loss of years of life produced by a hypothetical population after exposure to a given risk situation. The code generates modified life tables and estimates the impact of increased risk through several numerical comparisons with the appropriate reference life tables. One of the code's frequent applications is in estimating the number of radiation induced deaths that would result from exposing an initial population of 100,000 individuals to an annual radiation dose. For each risk situation analyzed, the computer code generates a summary table which documents the input, data and contains the results of the comparisons with reference life tables

  19. Theory of the space-dependent fuel management computer code ''UAFCC''

    International Nuclear Information System (INIS)

    El-Meshad, Y.; Morsy, S.; El-Osery, I.A.

    1981-01-01

    This report displays the theory of the spatial burnup computer code ''UAFCC'' which has been constructed as a part of an integrated reactor calculation scheme proposed at the Reactors Department of the ARE Atomic Energy Authority. The ''UAFCC'' is a single energy-one-dimensional diffusion burnup FORTRAN computer code for well moderated, multiregion, cylindrical thermal reactors. The effect of reactivity variation with burnup is introduced in the steady state diffusion equation by a fictitious neutron source. The infinite multiplication factor, the total migration area, and the power density per unit thermal flux are calculated from the point model burnup code ''UABUC'' fitted to polynomials of suitable degree in the flux-time, and then used as an input data to the ''UAFCC'' code. The proposed burnup spatial model has been used to study the different stratogemes of the incore fuel management schemes. The conclusions of this study will be presented in a future publication. (author)

  20. Compilation of documented computer codes applicable to environmental assessment of radioactivity releases

    International Nuclear Information System (INIS)

    Hoffman, F.O.; Miller, C.W.; Shaeffer, D.L.; Garten, C.T. Jr.; Shor, R.W.; Ensminger, J.T.

    1977-04-01

    The objective of this paper is to present a compilation of computer codes for the assessment of accidental or routine releases of radioactivity to the environment from nuclear power facilities. The capabilities of 83 computer codes in the areas of environmental transport and radiation dosimetry are summarized in tabular form. This preliminary analysis clearly indicates that the initial efforts in assessment methodology development have concentrated on atmospheric dispersion, external dosimetry, and internal dosimetry via inhalation. The incorporation of terrestrial and aquatic food chain pathways has been a more recent development and reflects the current requirements of environmental legislation and the needs of regulatory agencies. The characteristics of the conceptual models employed by these codes are reviewed. The appendixes include abstracts of the codes and indexes by author, key words, publication description, and title