WorldWideScience

Sample records for stress analysis code

  1. Electromagnetic field and mechanical stress analysis code

    International Nuclear Information System (INIS)

    1978-01-01

    Analysis TEXMAGST is a two stage linear finite element code for the analysis of static magnetic fields in three dimensional structures and associated mechanical stresses produced by the anti J x anti B forces within these structures. The electromagnetic problem is solved in terms of magnetic vector potential A for a given current density anti J as curl 1/μ curl anti A = anti J considering the magnetic permeability as constant. The Coulombian gauge (div anti A = o) was chosen and was implemented through the use of Lagrange multipliers. The second stage of the problem - the calculation of mechanical stresses in the same three dimensional structure is solved by using the same code with few modifications - through a restart card. Body forces anti J x anti B within each element are calculated from the solution of the first stage run and represent the input to the second stage run which will give the solution for the stress problem

  2. Analysis of pipe stress using CAESAR II code

    International Nuclear Information System (INIS)

    Sitandung, Y.B.; Bandriyana, B.

    2002-01-01

    Analysis of this piping stress with the purpose of knowing stress distribution piping system in order to determine pipe supports configuration. As an example of analysis, Gas Exchanger to Warm Separator Line was chosen with, input data was firstly prepared in a document, i.e. piping analysis specification that its content named as pipe characteristics, material properties, operation conditions, guide equipment's and so on. Analysis result such as stress, load, displacement and the use support type were verified based on requirements in the code, standard, and regularities were suitable with piping system condition analyzed. As the proof that piping system is in safety condition, it can be indicated from analysis results (actual loads) which still under allowable load. From the analysis steps that have been done CAESAR II code fulfill requirements to be used as a tool of piping stress analysis as well as nuclear and non nuclear installation piping system

  3. PIPE STRESS and VERPIP codes for stress analysis and verifications of PEC reactor piping

    International Nuclear Information System (INIS)

    Cesari, F.; Ferranti, P.; Gasparrini, M.; Labanti, L.

    1975-01-01

    To design LMFBR piping systems following ASME Sct. III requirements unusual flexibility computer codes are to be adopted to consider piping and its guard-tube. For this purpose PIPE STRESS code previously prepared by Southern-Service, has been modified. Some subroutine for detailed stress analysis and principal stress calculations on all the sections of piping have been written and fitted in the code. Plotter can also be used. VERPIP code for automatic verifications of piping as class 1 Sct. III prescriptions has been also prepared. The results of PIPE STRESS and VERPIP codes application to PEC piping are in section III of this report

  4. BILAM: a composite laminate failure-analysis code using bilinear stress-strain approximations

    Energy Technology Data Exchange (ETDEWEB)

    McLaughlin, P.V. Jr.; Dasgupta, A.; Chun, Y.W.

    1980-10-01

    The BILAM code which uses constant strain laminate analysis to generate in-plane load/deformation or stress/strain history of composite laminates to the point of laminate failure is described. The program uses bilinear stress-strain curves to model layer stress-strain behavior. Composite laminates are used for flywheels. The use of this computer code will help to develop data on the behavior of fiber composite materials which can be used by flywheel designers. In this program the stress-strain curves are modelled by assuming linear response in axial tension while using bilinear approximations (2 linear segments) for stress-strain response to axial compressive, transverse tensile, transverse compressive and axial shear loadings. It should be noted that the program attempts to empirically simulate the effects of the phenomena which cause nonlinear stress-strain behavior, instead of mathematically modelling the micromechanics involved. This code, therefore, performs a bilinear laminate analysis, and, in conjunction with several user-defined failure interaction criteria, is designed to provide sequential information on all layer failures up to and including the first fiber failure. The modus operandi is described. Code BILAM can be used to: predict the load-deformation/stress-strain behavior of a composite laminate subjected to a given combination of in-plane loads, and make analytical predictions of laminate strength.

  5. Development of an inelastic stress analysis code 'KINE-T' and its evaluations

    International Nuclear Information System (INIS)

    Kobatake, K.; Takahashi, S.; Suzuki, M.

    1977-01-01

    Referring to the ASME B and PVC Code Case 1592-7, the inelastic stress analysis is required for the designs of the class 1 components in elevated temperature if the results of the elastic stress analysis and/or simplified inelastic analysis do not satisfy the requirements. Authors programmed a two-dimensional axisymmetric inelastic analysis code 'KINE-T', and carried out its evaluations and an application. This FEM code is based on the incremental method and the following: elastic-plastic constitutive equation (yield condition of von Mises; flow rule of Prandtl-Reuss; Prager's hardening rule); creep constitutive equation (equation of state approach; flow rule of von Mises; strain hardening rule); the temperature dependency of the yield function is considered; solution procedure of the assembled stiffness matrix is the 'initial stress method'. After the completion of the programming, authors compared the output with not only theoretical results but also with those of the MARC code and the ANSYS code. In order to apply the code to the practical designing, authors settled a quasi-component two-dimensional axisymmetric model and a loading cycle (500 cycles). Then, an inelastic analysis and its integrity evaluation are carried out

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

    International Nuclear Information System (INIS)

    1989-01-01

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

  7. Elastic-plastic stress analysis and ASME code evaluation of a bottomhead penetration in a reactor pressure vessel

    International Nuclear Information System (INIS)

    Ranganath, S.

    1979-01-01

    Nuclear pressure vessel components are designed to meet the requirements of Section III of the ASME Boiler and Pressure Vessel Code. Specifically, the design must satisfy the limits on stress range and fatigue usage prescribed in NB-3200, Section III ASME Code for the various design and operating conditions for the component. The Code requirements assure that the component does not experience gross yielding and that in general, elastic shakedown occurs following cyclic loading. When elastic stress analysis is performed this can be shown by meeting the limits in the Code on Primary and Primary plus Secondary (P+Q) stress intensities. However, when the P+Q limits cannot be met and elastic Shakedown cannot be demonstrated, plastic analysis may be performed to meet the requirements of the Code. This paper describes the elastic-plastic stress analysis of a Boiling Water Reactor Vessel bottom head in-core penetration and illustrates how plastic analysis can be used in ASME Code evaluations to show Code compliance. Details of the thermal analysis, elastic-plastic stress analysis and fatigue evaluation are presented and it is shown that the in-core penetration satisfies the code requirements. 6 refs

  8. FERRET data analysis code

    International Nuclear Information System (INIS)

    Schmittroth, F.

    1979-09-01

    A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples

  9. Fuel performance analysis code 'FAIR'

    International Nuclear Information System (INIS)

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

    1994-01-01

    For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

  10. The evaluation of failure stress and released amount of fission product gas of power ramped rod by fuel behaviour analysis code 'FEMAXI-III'

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujita, Misao

    1984-01-01

    Pellet-Cladding Interaction(PCI) related in-pile failure of Zircaloy sheathed fuel rod is in general considered to be caused by combination of pellet-cladding mechanical interaction(PCMI) with fuel-cladding chemical interaction(FCCI). An understanding of a basic mechanism of PCI-related fuel failure is therefore necessary to get actual cladding hoop stress from mechanical interaction and released amounts of fission product(FP) gas of aggressive environmental agency from chemical interaction. This paper describes results of code analysis performed on fuel failure to cladding hoop stress and amounts of FP gas released under the condition associated with power ramping. Data from Halden(HBWR) and from Studsvik(R2) are used for code analysis. The fuel behaviour analysis code ''FEMAXI-III'' is used as an analytical tool. The followings are revealed from the study: (1) PCI-related fuel failure is dependent upon cladding hoop stress and released amounts of FP gas at power ramping. (2) Preliminary calculated threshold values of hoop stress and of released amounts of FP gas to PCI failure are respectively 330MPa, 10% under the Halden condition, 190MPa, 5% under the Inter ramp(BWR) condition, and 270MPa, 14% under the Over ramp(PWR) condition. The values of hoop stress calculated are almost in the similar range of those obtained from ex-reactor PCI simulated tests searched from references published. (3) The FEMAXI-III code verification is made in mechanical manner by using in-pile deformation data(diametral strain) obtained from power ramping test undertaken by JAERI. While, the code verification is made in thermal manner by using punctured FP gas data obtained from post irradiation examination performed on non-defected power ramped fuel rods. The calculations are resulted in good agreements to both, mechanical and thermal experimental data suggesting the validity of the code evaluation. (J.P.N.)

  11. Nuclear class 1 piping stress analysis

    International Nuclear Information System (INIS)

    Lucas, J.C.R.; Maneschy, J.E.; Mariano, L.A.; Tamura, M.

    1981-01-01

    A nuclear class 1 piping stress analysis, according to the ASME code, is presented. The TRHEAT computer code has been used to determine the piping wall thermal gradient. The Nupipe computer code was employed for the piping stress analysis. Computer results were compared with the allowable criteria from the ASME code. (Author) [pt

  12. Identification and characterization of wheat long non-protein coding RNAs responsive to powdery mildew infection and heat stress by using microarray analysis and SBS sequencing

    Directory of Open Access Journals (Sweden)

    Peng Huiru

    2011-04-01

    Full Text Available Abstract Background Biotic and abiotic stresses, such as powdery mildew infection and high temperature, are important limiting factors for yield and grain quality in wheat production. Emerging evidences suggest that long non-protein coding RNAs (npcRNAs are developmentally regulated and play roles in development and stress responses of plants. However, identification of long npcRNAs is limited to a few plant species, such as Arabidopsis, rice and maize, no systematic identification of long npcRNAs and their responses to abiotic and biotic stresses is reported in wheat. Results In this study, by using computational analysis and experimental approach we identified 125 putative wheat stress responsive long npcRNAs, which are not conserved among plant species. Among them, some were precursors of small RNAs such as microRNAs and siRNAs, two long npcRNAs were identified as signal recognition particle (SRP 7S RNA variants, and three were characterized as U3 snoRNAs. We found that wheat long npcRNAs showed tissue dependent expression patterns and were responsive to powdery mildew infection and heat stress. Conclusion Our results indicated that diverse sets of wheat long npcRNAs were responsive to powdery mildew infection and heat stress, and could function in wheat responses to both biotic and abiotic stresses, which provided a starting point to understand their functions and regulatory mechanisms in the future.

  13. A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

    International Nuclear Information System (INIS)

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya

    2001-05-01

    In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10 25 m -2 , respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

  14. RNA-Seq analysis of D. radiodurans find non coding RNAs expressed in response to radiation stress

    International Nuclear Information System (INIS)

    Gadewal, Nikhil; Mukhopadhyaya, Rita

    2015-01-01

    In bacteria discovery of functional RNA molecules that are not translated into protein, noncoding RNAs, became possible with advent of Next Generation Sequencing technology. Bacterial non coding RNAs are typically 50-300 nucleotides long and work as internal signals controlling various levels of gene expression. Deep sequencing of total cellular RNA captures all coding and noncoding transcripts with their differential levels of expression in the transcriptome. It provides a powerful approach to study bacterial gene expression and mechanisms of gene regulation. We subjected the 3 h transcriptome of Deinococcus radiodurans R1 cells post exposure to 6 KGy gamma radiation to 100 x 2 cycles of deep sequencing on the Illumina HiSeq 2000 to look for ncRNA transcripts. Bioinformatics pipeline for analysis and interpretation of RNA Seq data was done in house using Softwares available in public domains. Our sequence data aligned with 21 putative ncRNAs expressed in the intergenic regions of annotated genome of D radiodurans. Verification of 2 ncRNA candidates and 3 transcription factor genes by Real Time PCR confirmed presence of these transcripts in the 3 h transcriptome sequenced by us. Any relationship between ncRNAs and control of radiation induced gene expression in D radiodurans can be proved only after specific gene knock outs in future. (author)

  15. 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

  16. Stress analysis of pressure vessels

    International Nuclear Information System (INIS)

    Kim, B.K.; Song, D.H.; Son, K.H.; Kim, K.S.; Park, K.B.; Song, H.K.; So, J.Y.

    1979-01-01

    This interim report contains the results of the effort to establish the stress report preparation capability under the research project ''Stress analysis of pressure vessels.'' 1978 was the first year in this effort to lay the foundation through the acquisition of SAP V structural analysis code and a graphic terminal system for improved efficiency of using such code. Software programming work was developed in pre- and post processing, such as graphic presentation of input FEM mesh geometry and output deformation or mode shope patterns, which was proven to be useful when using the FEM computer code. Also, a scheme to apply fracture mechanics concept was developed in fatigue analysis of pressure vessels. (author)

  17. Development and verification of coupled fluid-structural dynamic codes for stress analysis of reactor vessel internals under blowdown loading

    International Nuclear Information System (INIS)

    Krieg, R.; Schlechtendahl, E.G.

    1977-01-01

    YAQUIR has been applied to large PWR blowdown problems and compared with LECK results. The structural model of CYLDY2 and the fluid model of YAQUIR have been coupled in the code STRUYA. First tests with the fluid dynamic systems code FLUST have been successful. The incompressible fluid version of the 3D coupled code FLUX for HDR-geometry was checked against some analytical test cases and was used for evaluation of the eigenfrequencies of the coupled system. Several test cases were run with the two phase flow code SOLA-DF with satisfactory results. Remarkable agreement was found between YAQUIR results and experimental data obtained from shallow water analogy experiments. A test for investigation of nonequilibrium twophase flow dynamics has been specified in some detail. The test is to be performed early 1978 in the water loop of the IRB. Good agreement was found between the natural frequency predictions for the core barrel obtained from CYLDY2 and STRUDL/DYNAL. Work started on improvement of the beam mode treatment in CYLDY2. The name of this modified version will be CYLDY3. The fluiddynamic code SING1, based on an advanced singularity method and applicable to a broad class of highly transient, incompressible 3D-problems with negligible viscosity has been developed and tested. It will be used in connection with the planned laboratory experiments in order to investigate the effect of the core structure on the blowdown process. Coupling of SING1 with structural dynamics is on the way. (orig./RW) [de

  18. Stress Analysis

    DEFF Research Database (Denmark)

    Burcharth, Hans F.

    The following types of forces contribute to the stresses in a Dolos in a pack exposed to waves: 1)Gravity forces Compaction forces (mainly due to settlements, gravity and flow forces) 2) Flow forces 3) Impact forces (impacts between moving concrete blocks)......The following types of forces contribute to the stresses in a Dolos in a pack exposed to waves: 1)Gravity forces Compaction forces (mainly due to settlements, gravity and flow forces) 2) Flow forces 3) Impact forces (impacts between moving concrete blocks)...

  19. CINETHICA - Core accident analysis code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

  20. ZERBERUS - the code for reliability analysis of crack containing structures

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.

    1992-04-01

    Brief description of the First- and Second Order Reliability Methods, being the theoretical background of the code, is given. The code structure is described in detail, with special emphasis to the new application fields. The numerical example investigates failure probability of steam generator tubing affected by stress corrosion cracking. The changes necessary to accommodate this analysis within the ZERBERUS code are explained. Analysis results are compared to different Monte Carlo techniques. (orig./HP) [de

  1. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  2. Stress: a concept analysis.

    Science.gov (United States)

    Goodnite, Patricia M

    2014-01-01

    To analyze the concept of stress and provide an operational definition of stress. Literature review revealed that stress is a commonly used, but often ambiguous, term. Findings supported a definition of stress entailing an individual's perception of a stimulus as overwhelming, which in turn elicits a measurable response resulting in a transformed state. This analysis adopts a dynamic definition of stress that may serve to encourage communication, promote reflection, and enhance concept understanding. This definition may provide direction for future work, as well as enhance efforts to serve patients affected by stress. © 2013 Wiley Periodicals, Inc.

  3. Extending CANTUP code analysis to probabilistic evaluations

    International Nuclear Information System (INIS)

    Florea, S.

    2001-01-01

    The structural analysis with numerical methods based on final element method plays at present a central role in evaluations and predictions of structural systems which require safety and reliable operation in aggressive environmental conditions. This is the case too for the CANDU - 600 fuel channel, where besides the corrosive and thermal aggression upon the Zr97.5Nb2.5 pressure tubes, a lasting irradiation adds which has marked consequences upon the materials properties evolution. This results in an unavoidable spreading in the materials properties in time, affected by high uncertainties. Consequently, the deterministic evaluation with computation codes based on finite element method are supplemented by statistic and probabilistic methods of evaluation of the response of structural components. This paper reports the works on extending the thermo-mechanical evaluation of the fuel channel components in the frame of probabilistic structure mechanics based on statistical methods and developed upon deterministic CANTUP code analyses. CANTUP code was adapted from LAHEY 77 platform onto Microsoft Developer Studio - Fortran Power Station 4.0 platform. To test the statistical evaluation of the creeping behaviour of pressure tube, the value of longitudinal elasticity modulus (Young) was used, as random variable, with a normal distribution around value, as used in deterministic analyses. The influence of the random quantity upon the hog and effective stress developed in the pressure tube for to time values, specific to primary and secondary creep was studied. The results obtained after a five year creep, corresponding to the secondary creep are presented

  4. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2009-01-01

    In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

  5. Stability analysis by ERATO code

    International Nuclear Information System (INIS)

    Tsunematsu, Toshihide; Takeda, Tatsuoki; Matsuura, Toshihiko; Azumi, Masafumi; Kurita, Gen-ichi

    1979-12-01

    Problems in MHD stability calculations by ERATO code are described; which concern convergence property of results, equilibrium codes, and machine optimization of ERATO code. It is concluded that irregularity on a convergence curve is not due to a fault of the ERATO code itself but due to inappropriate choice of the equilibrium calculation meshes. Also described are a code to calculate an equilibrium as a quasi-inverse problem and a code to calculate an equilibrium as a result of a transport process. Optimization of the code with respect to I/O operations reduced both CPU time and I/O time considerably. With the FACOM230-75 APU/CPU multiprocessor system, the performance is about 6 times as high as with the FACOM230-75 CPU, showing the effectiveness of a vector processing computer for the kind of MHD computations. This report is a summary of the material presented at the ERATO workshop 1979(ORNL), supplemented with some details. (author)

  6. Uncertainty analysis of the FRAP code

    International Nuclear Information System (INIS)

    Peck, S.O.

    1978-01-01

    A user oriented, automated uncertainty analysis capability has been built into the FRAP code (Fuel Rod Analysis Program) and applied to a PWR fuel rod undergoing a LOCA. The method of uncertainty analysis is the Response Surface Method (RSM). (author)

  7. Comparative analysis of design codes for timber bridges in Canada, the United States, and Europe

    Science.gov (United States)

    James Wacker; James (Scott) Groenier

    2010-01-01

    The United States recently completed its transition from the allowable stress design code to the load and resistance factor design (LRFD) reliability-based code for the design of most highway bridges. For an international perspective on the LRFD-based bridge codes, a comparative analysis is presented: a study addressed national codes of the United States, Canada, and...

  8. Basic stress analysis

    CERN Document Server

    Iremonger, M J

    1982-01-01

    BASIC Stress Analysis aims to help students to become proficient at BASIC programming by actually using it in an important engineering subject. It also enables the student to use computing as a means of learning stress analysis because writing a program is analogous to teaching-it is necessary to understand the subject matter. The book begins by introducing the BASIC approach and the concept of stress analysis at first- and second-year undergraduate level. Subsequent chapters contain a summary of relevant theory, worked examples containing computer programs, and a set of problems. Topics c

  9. SASSYS LMFBR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.

    1982-01-01

    The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

  10. Parallel processing of structural integrity analysis codes

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.

    1996-01-01

    Structural integrity analysis forms an important role in assessing and demonstrating the safety of nuclear reactor components. This analysis is performed using analytical tools such as Finite Element Method (FEM) with the help of digital computers. The complexity of the problems involved in nuclear engineering demands high speed computation facilities to obtain solutions in reasonable amount of time. Parallel processing systems such as ANUPAM provide an efficient platform for realising the high speed computation. The development and implementation of software on parallel processing systems is an interesting and challenging task. The data and algorithm structure of the codes plays an important role in exploiting the parallel processing system capabilities. Structural analysis codes based on FEM can be divided into two categories with respect to their implementation on parallel processing systems. The first category codes such as those used for harmonic analysis, mechanistic fuel performance codes need not require the parallelisation of individual modules of the codes. The second category of codes such as conventional FEM codes require parallelisation of individual modules. In this category, parallelisation of equation solution module poses major difficulties. Different solution schemes such as domain decomposition method (DDM), parallel active column solver and substructuring method are currently used on parallel processing systems. Two codes, FAIR and TABS belonging to each of these categories have been implemented on ANUPAM. The implementation details of these codes and the performance of different equation solvers are highlighted. (author). 5 refs., 12 figs., 1 tab

  11. Analysis of stress distribution of timing belts by FEM; Yugen yosoho ni yoru timing belt oryoku kaiseki (belt code oryoku bunpu kaiseki hokoku)

    Energy Technology Data Exchange (ETDEWEB)

    Furukawa, Y; Tomono, K; Takahashi, H; Uchida, T [Honda R and D Co. Ltd., Tokyo (Japan)

    1997-10-01

    A model of the belt analyzed by-ABAQUS (: a general nonlinear finite element program) successfully confirmed the mechanism that generates the belt cord stress. A quite good agreement between experimental and computed results for the stress distribution of the belt cord. It is found that maximum stress of the cords occurs near the root of the tooth by calculation, where the belt cords break off. 3 refs., 9 figs.

  12. ENSDF analysis codes. IBM version, August 1982

    International Nuclear Information System (INIS)

    Lorenz, A.

    1982-01-01

    The nuclear structure analysis programme tape consists of physics computer processing codes used in the evaluation of mass-chain structure data. This tape was generated by the National Nuclear Data Centre, Brookhaven National Laboratory in the USA. (author)

  13. ENSDF analysis codes: IBM version. August 1982

    International Nuclear Information System (INIS)

    Lorenz, A.

    1982-09-01

    The nuclear structure analysis programme tape consists of physics computer processing codes used in the evaluation of mass-chain structure data. This tape was generated by the National Nuclear Data Centre, Brookhaven National Laboratory in the USA. (author)

  14. R-matrix analysis code (RAC)

    International Nuclear Information System (INIS)

    Chen Zhenpeng; Qi Huiquan

    1990-01-01

    A comprehensive R-matrix analysis code has been developed. It is based on the multichannel and multilevel R-matrix theory and runs in VAX computer with FORTRAN-77. With this code many kinds of experimental data for one nuclear system can be fitted simultaneously. The comparisions between code RAC and code EDA of LANL are made. The data show both codes produced the same calculation results when one set of R-matrix parameters was used. The differential cross section of 10 B (n, α) 7 Li for E n = 0.4 MeV and the polarization of 16 O (n,n) 16 O for E n = 2.56 MeV are presented

  15. Stress Analysis of Composites.

    Science.gov (United States)

    1981-01-01

    8217, Finite Elements in Nonlinear Mechanics, 1., 109-130, Tapir Publishers, Norway (1978). 9. A.J. Barnard and P.W. Sharman, ’Elastic-Plastic Analysis Using...Hybrid Stress Finite Elements,’ Finite Elements in Nonlinear Mechanics, 1, 131-148, Tapir Publishers Norway, (1978). ’.........Pian, ’Variational

  16. Appropriate nominal stresses for use with ASME Code pressure-loading stress indices for nozzles

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.

    1976-06-01

    This program is part of a cooperative effort with industry to develop and verify analytical methods for assessing the safety of nuclear pressure-vessel and piping-system design. The study of nominal stresses and stress indices described is part of a continuing study of design rules for nozzles in pressure vessels being coordinated by the PVRC Subcommittee on Reinforced Openings and External Loadings. Results from these studies are used by appropriate ASME Code groups in drafting new and improved design rules

  17. Two-dimensional disruption thermal analysis code DREAM

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Kobayashi, Takeshi; Seki, Masahiro.

    1988-08-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)

  18. Centrifugal Compressor Aeroelastic Analysis Code

    Science.gov (United States)

    Keith, Theo G., Jr.; Srivastava, Rakesh

    2002-01-01

    Centrifugal compressors are very widely used in the turbomachine industry where low mass flow rates are required. Gas turbine engines for tanks, rotorcraft and small jets rely extensively on centrifugal compressors for rugged and compact design. These compressors experience problems related with unsteadiness of flowfields, such as stall flutter, separation at the trailing edge over diffuser guide vanes, tip vortex unsteadiness, etc., leading to rotating stall and surge. Considerable interest exists in small gas turbine engine manufacturers to understand and eventually eliminate the problems related to centrifugal compressors. The geometric complexity of centrifugal compressor blades and the twisting of the blade passages makes the linear methods inapplicable. Advanced computational fluid dynamics (CFD) methods are needed for accurate unsteady aerodynamic and aeroelastic analysis of centrifugal compressors. Most of the current day industrial turbomachines and small aircraft engines are designed with a centrifugal compressor. With such a large customer base and NASA Glenn Research Center being, the lead center for turbomachines, it is important that adequate emphasis be placed on this area as well. Currently, this activity is not supported under any project at NASA Glenn.

  19. NORTICA - a new code for cyclotron analysis

    International Nuclear Information System (INIS)

    Gorelov, D.; Johnson, D.; Marti, F.

    2001-01-01

    The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state

  20. Implications of stress range for inelastic analysis

    International Nuclear Information System (INIS)

    Karabin, M.E.; Dhalla, A.K.

    1981-01-01

    The elastic stress range over a complete load cycle is routinely used to formulate simplified rules regarding the inelastic behavior of structures operating at elevated temperature. For example, a 300 series stainless steel structure operating at elevated temperature, in all probability, would satisfy the ASME Boiler and Pressure Vessel Code criteria if the linearized elastic stress range is less than three times the material yield strength. However, at higher elastic stress ranges it is difficult to judge, a priori, that a structural component would comply with inelastic Code criteria after a detailed inelastic analysis. The purpose of this paper is to illustrate that it is not the elastic stress range but the stress intensities at specific times during a thermal transient which provide a better insight into the inelastic response of the structure. The specific example of the CRBRP flued head design demonstrates that the temperature differential between various parts of the structure can be changed by modifying the insulation pattern and heat flow path in the structure, without significantly altering the elastic stress range over a complete load cycle. However, the modified design did reduce the stress intensity during steady state elevated temperature operation. This modified design satisfied the inelastic Code criteria whereas the initial design failed to comply with the strain accumulation criterion

  1. CATHENA 4. A thermalhydraulics network analysis code

    International Nuclear Information System (INIS)

    Aydemir, N.U.; Hanna, B.N.

    2009-01-01

    Canadian Algorithm for THErmalhydraulic Network Analysis (CATHENA) is a one-dimensional, non-equilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. The objective of the present paper is to describe the design, application and future development plans for the CATHENA 4 thermalhydraulics network analysis code, which is a modernized version of the present frozen CATHENA 3 code. The new code is designed in modular form, using the Fortran 95 (F95) programming language. The semi-implicit numerical integration scheme of CATHENA 3 is re-written to implement a fully-implicit methodology using Newton's iterative solution scheme suitable for nonlinear equations. The closure relations, as a first step, have been converted from the existing CATHENA 3 implementation to F95 but modularized to achieve ease of maintenance. The paper presents the field equations, followed by a description of the Newton's scheme used. The finite-difference form of the field equations is given, followed by a discussion of convergence criteria. Two applications of CATHENA 4 are presented to demonstrate the temporal and spatial convergence of the new code for problems with known solutions or available experimental data. (author)

  2. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  3. Static Code Analysis with Gitlab-CI

    CERN Document Server

    Datko, Szymon Tomasz

    2016-01-01

    Static Code Analysis is a simple but efficient way to ensure that application’s source code is free from known flaws and security vulnerabilities. Although such analysis tools are often coming with more advanced code editors, there are a lot of people who prefer less complicated environments. The easiest solution would involve education – where to get and how to use the aforementioned tools. However, counting on the manual usage of such tools still does not guarantee their actual usage. On the other hand, reducing the required effort, according to the idea “setup once, use anytime without sweat” seems like a more promising approach. In this paper, the approach to automate code scanning, within the existing CERN’s Gitlab installation, is described. For realization of that project, the Gitlab-CI service (the “CI” stands for "Continuous Integration"), with Docker assistance, was employed to provide a variety of static code analysers for different programming languages. This document covers the gene...

  4. 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)

  5. Uncertainty analysis of the FRAP code

    International Nuclear Information System (INIS)

    Peck, S.O.

    1978-01-01

    A user oriented, automated uncertainty analysis capability has been built into the Fuel Rod Analysis Program (FRAP) code and has been applied to a pressurized water reactor (PWR) fuel rod undergoing a loss-of-coolant accident (LOCA). The method of uncertainty analysis is the response surface method. The automated version significantly reduced the time required to complete the analysis and, at the same time, greatly increased the problem scope. Results of the analysis showed a significant difference in the total and relative contributions to the uncertainty of the response parameters between steady state and transient conditions

  6. Benchmark calculation of subchannel analysis codes

    International Nuclear Information System (INIS)

    1996-02-01

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

  7. Sandia National Laboratories analysis code data base

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, C.W.

    1994-11-01

    Sandia National Laboratories, mission is to solve important problems in the areas of national defense, energy security, environmental integrity, and industrial technology. The Laboratories` strategy for accomplishing this mission is to conduct research to provide an understanding of the important physical phenomena underlying any problem, and then to construct validated computational models of the phenomena which can be used as tools to solve the problem. In the course of implementing this strategy, Sandia`s technical staff has produced a wide variety of numerical problem-solving tools which they use regularly in the design, analysis, performance prediction, and optimization of Sandia components, systems and manufacturing processes. This report provides the relevant technical and accessibility data on the numerical codes used at Sandia, including information on the technical competency or capability area that each code addresses, code ``ownership`` and release status, and references describing the physical models and numerical implementation.

  8. Sandia National Laboratories analysis code data base

    Science.gov (United States)

    Peterson, C. W.

    1994-11-01

    Sandia National Laboratories' mission is to solve important problems in the areas of national defense, energy security, environmental integrity, and industrial technology. The laboratories' strategy for accomplishing this mission is to conduct research to provide an understanding of the important physical phenomena underlying any problem, and then to construct validated computational models of the phenomena which can be used as tools to solve the problem. In the course of implementing this strategy, Sandia's technical staff has produced a wide variety of numerical problem-solving tools which they use regularly in the design, analysis, performance prediction, and optimization of Sandia components, systems, and manufacturing processes. This report provides the relevant technical and accessibility data on the numerical codes used at Sandia, including information on the technical competency or capability area that each code addresses, code 'ownership' and release status, and references describing the physical models and numerical implementation.

  9. Development of chemical equilibrium analysis code 'CHEEQ'

    International Nuclear Information System (INIS)

    Nagai, Shuichiro

    2006-08-01

    'CHEEQ' code which calculates the partial pressure and the mass of the system consisting of ideal gas and pure condensed phase compounds, was developed. Characteristics of 'CHEEQ' code are as follows. All the chemical equilibrium equations were described by the formation reactions from the mono-atomic gases in order to simplify the code structure and input preparation. Chemical equilibrium conditions, Σν i μ i =0 for the gaseous compounds and precipitated condensed phase compounds and Σν i μ i > 0 for the non-precipitated condensed phase compounds, were applied. Where, ν i and μ i are stoichiometric coefficient and chemical potential of component i. Virtual solid model was introduced to perform the calculation of constant partial pressure condition. 'CHEEQ' was consisted of following 3 parts, (1) analysis code, zc132. f. (2) thermodynamic data base, zmdb01 and (3) input data file, zindb. 'CHEEQ' code can calculate the system which consisted of elements (max.20), condensed phase compounds (max.100) and gaseous compounds. (max.200). Thermodynamic data base, zmdb01 contains about 1000 elements and compounds, and 200 of them were Actinide elements and their compounds. This report describes the basic equations, the outline of the solution procedure and instructions to prepare the input data and to evaluate the calculation results. (author)

  10. Code comparison for accelerator design and analysis

    International Nuclear Information System (INIS)

    Parsa, Z.

    1988-01-01

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

  11. Fire-accident analysis code (FIRAC) verification

    International Nuclear Information System (INIS)

    Nichols, B.D.; Gregory, W.S.; Fenton, D.L.; Smith, P.R.

    1986-01-01

    The FIRAC computer code predicts fire-induced transients in nuclear fuel cycle facility ventilation systems. FIRAC calculates simultaneously the gas-dynamic, material transport, and heat transport transients that occur in any arbitrarily connected network system subjected to a fire. The network system may include ventilation components such as filters, dampers, ducts, and blowers. These components are connected to rooms and corridors to complete the network for moving air through the facility. An experimental ventilation system has been constructed to verify FIRAC and other accident analysis codes. The design emphasizes network system characteristics and includes multiple chambers, ducts, blowers, dampers, and filters. A larger industrial heater and a commercial dust feeder are used to inject thermal energy and aerosol mass. The facility is instrumented to measure volumetric flow rate, temperature, pressure, and aerosol concentration throughout the system. Aerosol release rates and mass accumulation on filters also are measured. We have performed a series of experiments in which a known rate of thermal energy is injected into the system. We then simulated this experiment with the FIRAC code. This paper compares and discusses the gas-dynamic and heat transport data obtained from the ventilation system experiments with those predicted by the FIRAC code. The numerically predicted data generally are within 10% of the experimental data

  12. Web interface for plasma analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Emoto, M. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)], E-mail: emo@nifs.ac.jp; Murakami, S. [Kyoto University, Yoshida-Honmachi, Sakyo-ku, Kyoto 606-8501 (Japan); Yoshida, M.; Funaba, H.; Nagayama, Y. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)

    2008-04-15

    There are many analysis codes that analyze various aspects of plasma physics. However, most of them are FORTRAN programs that are written to be run in supercomputers. On the other hand, many scientists use GUI (graphical user interface)-based operating systems. For those who are not familiar with supercomputers, it is a difficult task to run analysis codes in supercomputers, and they often hesitate to use these programs to substantiate their ideas. Furthermore, these analysis codes are written for personal use, and the programmers do not expect these programs to be run by other users. In order to make these programs to be widely used by many users, the authors developed user-friendly interfaces using a Web interface. Since the Web browser is one of the most common applications, it is useful for both the users and developers. In order to realize interactive Web interface, AJAX technique is widely used, and the authors also adopted AJAX. To build such an AJAX based Web system, Ruby on Rails plays an important role in this system. Since this application framework, which is written in Ruby, abstracts the Web interfaces necessary to implement AJAX and database functions, it enables the programmers to efficiently develop the Web-based application. In this paper, the authors will introduce the system and demonstrate the usefulness of this approach.

  13. Web interface for plasma analysis codes

    International Nuclear Information System (INIS)

    Emoto, M.; Murakami, S.; Yoshida, M.; Funaba, H.; Nagayama, Y.

    2008-01-01

    There are many analysis codes that analyze various aspects of plasma physics. However, most of them are FORTRAN programs that are written to be run in supercomputers. On the other hand, many scientists use GUI (graphical user interface)-based operating systems. For those who are not familiar with supercomputers, it is a difficult task to run analysis codes in supercomputers, and they often hesitate to use these programs to substantiate their ideas. Furthermore, these analysis codes are written for personal use, and the programmers do not expect these programs to be run by other users. In order to make these programs to be widely used by many users, the authors developed user-friendly interfaces using a Web interface. Since the Web browser is one of the most common applications, it is useful for both the users and developers. In order to realize interactive Web interface, AJAX technique is widely used, and the authors also adopted AJAX. To build such an AJAX based Web system, Ruby on Rails plays an important role in this system. Since this application framework, which is written in Ruby, abstracts the Web interfaces necessary to implement AJAX and database functions, it enables the programmers to efficiently develop the Web-based application. In this paper, the authors will introduce the system and demonstrate the usefulness of this approach

  14. Introduction of thermal-hydraulic analysis code and system analysis code for HTGR

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1984-01-01

    Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)

  15. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

    2007-07-01

    For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

  16. Development of disruption thermal analysis code DREAM

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, Seiichiro; Kobayahsi, Takeshi [Kawasaki Heavy Industries Ltd., Kobe (Japan); Seki, Masahiro

    1989-07-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author).

  17. Development of disruption thermal analysis code DREAM

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Kobayahsi, Takeshi; Seki, Masahiro.

    1989-01-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author)

  18. CONTEMPT-DG containment analysis code

    International Nuclear Information System (INIS)

    Deem, R.E.; Rousseau, K.

    1982-01-01

    The assessment of hydrogen burning in a containment building during a degraded core event requires a knowledge of various system responses. These system responses (i.e. heat sinks, fan cooler units, sprays, etc.) can have a marked effect on the overall containment integrity results during a hydrogen burn. In an attempt to properly handle the various system responses and still retain the capability to perform sensitivity analysis on various parameters, the CONTEMPT-DG computer code was developed. This paper will address the historical development of the code, its various features, and the rationale for its development. Comparisons between results from the CONTEMPT-DG analyses and results from similar MARCH analyses will also be given

  19. 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.)

  20. SLSF loop handling system. Volume III. AISC code evaluations and analysis of critical attachments

    International Nuclear Information System (INIS)

    Ahmed, H.; Cowie, A.; Malek, R.A.; Rafer, A.; Ma, D.; Tebo, F.

    1978-10-01

    SLSF loop handling system was analyzed for deadweight and postulated dynamic loading conditions using a linear elastic static equivalent method of stress analysis. Stress computations of Cradle and critical attachments per AISC Code guidelines are presented. HFEF is credited with in-depth review of initial phase of work

  1. Numerical analysis and nuclear standard code application to thermal fatigue

    International Nuclear Information System (INIS)

    Merola, M.

    1992-01-01

    The present work describes the Joint Research Centre Ispra contribution to the IAEA benchmark exercise 'Lifetime Behaviour of the First Wall of Fusion Machines'. The results of the numerical analysis of the reference thermal fatigue experiment are presented. Then a discussion on the numerical analysis of thermal stress is tackled, pointing out its particular aspects in view of their influence on the stress field evaluation. As far as the design-allowable number of cycles are concerned the American nuclear code ASME and the French code RCC-MR are applied and the reasons for the different results obtained are investigated. As regards a realistic fatigue lifetime evaluation, the main problems to be solved are brought out. This work, is intended as a preliminary basis for a discussion focusing on the main characteristics of the thermal fatigue problem from both a numerical and a lifetime assessment point of view. In fact the present margin of discretion left to the analyst may cause undue discrepancies in the results obtained. A sensitivity analysis of the main parameters involved is desirable and more precise design procedures should be stated

  2. 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

  3. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

    To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

  4. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2005-02-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

  5. Code-B-1 for stress/strain calculation for TRISO fuel particle (Contract research)

    International Nuclear Information System (INIS)

    Aihara, Jun; Ueta, Shohei; Shibata, Taiju; Sawa, Kazuhiro

    2011-12-01

    We have developed Code-B-1 for the prediction of the failure probabilities of the coated fuel particles for the high temperature gas-cooled reactors (HTGRs) under operation by modification of an existing code. A finite element method (FEM) is employed for the stress calculation part and Code-B-1 can treat the plastic deformation of the coating layer of the coated fuel particles which the existing code cannot treat. (author)

  6. Current lead thermal analysis code 'CURRENT'

    International Nuclear Information System (INIS)

    Yamaguchi, Masahito; Tada, Eisuke; Shimamoto, Susumu; Hata, Kenichiro.

    1985-08-01

    Large gas-cooled current lead with the capacity more than 30 kA and 22 kV is required for superconducting toroidal and poloidal coils for fusion application. The current lead is used to carry electrical current from the power supply system at room temperature to the superconducting coil at 4 K. Accordingly, the thermal performance of the current lead is significantly important to determine the heat load requirements of the coil system at 4 K. Japan Atomic Energy Research Institute (JAERI) has being developed the large gas-cooled current leads with the optimum condition in which the heat load is around 1 W per 1 kA at 4 K. In order to design the current lead with the optimum thermal performances, JAERI developed thermal analysis code named as ''CURRENT'' which can theoretically calculate the optimum geometric shape and cooling conditions of the current lead. The basic equations and the instruction manual of the analysis code are described in this report. (author)

  7. Spartan Release Engagement Mechanism (REM) stress and fracture analysis

    Science.gov (United States)

    Marlowe, D. S.; West, E. J.

    1984-01-01

    The revised stress and fracture analysis of the Spartan REM hardware for current load conditions and mass properties is presented. The stress analysis was performed using a NASTRAN math model of the Spartan REM adapter, base, and payload. Appendix A contains the material properties, loads, and stress analysis of the hardware. The computer output and model description are in Appendix B. Factors of safety used in the stress analysis were 1.4 on tested items and 2.0 on all other items. Fracture analysis of the items considered fracture critical was accomplished using the MSFC Crack Growth Analysis code. Loads and stresses were obtaind from the stress analysis. The fracture analysis notes are located in Appendix A and the computer output in Appendix B. All items analyzed met design and fracture criteria.

  8. A Nucleus-localized Long Non-Coding RNA Enhances Drought and Salt Stress Tolerance

    KAUST Repository

    Qin, Tao

    2017-09-09

    Long non-coding RNAs (lncRNAs) affect gene expression through a wide range of mechanisms and are considered as important regulators in many essential biological processes. A large number of lncRNA transcripts have been predicted or identified in plants in recent years. However, the biological functions for most of them are still unknown. In this study, we identified an Arabidopsis thaliana lncRNA, Drought induced RNA (DRIR), as a novel positive regulator of plant response to drought and salt stress. DRIR was expressed at a low level under non-stress conditions but can be significantly activated by drought and salt stress as well as by abscisic acid (ABA) treatment. We identified a T-DNA insertion mutant, drirD, which had higher expression of the DRIR gene than the wild type plants. The drirD mutant exhibits increased tolerance to drought and salt stress. Overexpressing DRIR in Arabidopsis also increased tolerance to drought and salt stress of the transgenic plants. The drirD mutant and the overexpressing seedlings are more sensitive to ABA than the wild type in stomata closure and seedling growth. Genome-wide transcriptome analysis demonstrated that the expression of a large number of genes was altered in drirD and the overexpressing plants. These include genes involved in ABA signaling, water transport and other stress-relief processes. Our study reveals a mechanism whereby DRIR regulates plant response to abiotic stress by modulating the expression of a series of genes involved in stress response.

  9. Stress analysis of PCV nozzle junction

    International Nuclear Information System (INIS)

    Uchiyama, Shoichi; Oikawa, Tsuneo; Hoshino, Seizo

    1976-01-01

    Most of various pressure vessels comprise each one cylindrical shell and one or more nozzles. In this study, in order to analyze the stress in the structures of this type as minutely and exactly as possible, the program for stress analysis by the finite element method was made, which is required for the strength analysis for three-dimensional structures. Especially, the problem of the stress distribution around nozzle junctions was solved theoretically with the program. The program for the analysis developed in this study is provided with various functions, such as the input generator for cylindrical, conical and spherical shells, and plotter, and is very covenient. The accuracy of analysis is very good. The method of analysis and the calculation of the rigidity matrices for the deformation in plane and bending are explained. The result of the stress analysis around the nozzle junctions of a containment vessel with this program was in good agreement with experimental data and the result with SAP-4 code, therefore the propriety of the calculated result with this program was proved. Also calculations were carried out on three cases, namely a flat plate fixed at one end with distributed load, a cylinder fixed at one end with internal pressure, and an I-beam fixed at one end with concentrated load. The calculated results agreed well with theoretical solutions in all cases. (Kako, I.)

  10. Performance testing of thermal analysis codes for nuclear fuel casks

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

    In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

  11. Error-correction coding and decoding bounds, codes, decoders, analysis and applications

    CERN Document Server

    Tomlinson, Martin; Ambroze, Marcel A; Ahmed, Mohammed; Jibril, Mubarak

    2017-01-01

    This book discusses both the theory and practical applications of self-correcting data, commonly known as error-correcting codes. The applications included demonstrate the importance of these codes in a wide range of everyday technologies, from smartphones to secure communications and transactions. Written in a readily understandable style, the book presents the authors’ twenty-five years of research organized into five parts: Part I is concerned with the theoretical performance attainable by using error correcting codes to achieve communications efficiency in digital communications systems. Part II explores the construction of error-correcting codes and explains the different families of codes and how they are designed. Techniques are described for producing the very best codes. Part III addresses the analysis of low-density parity-check (LDPC) codes, primarily to calculate their stopping sets and low-weight codeword spectrum which determines the performance of these codes. Part IV deals with decoders desi...

  12. 76 FR 64931 - Building Energy Codes Cost Analysis

    Science.gov (United States)

    2011-10-19

    ...-0046] Building Energy Codes Cost Analysis AGENCY: Office of Energy Efficiency and Renewable Energy... reopening of the time period for submitting comments on the request for information on Building Energy Codes... the request for information on Building Energy Code Cost Analysis and provide docket number EERE-2011...

  13. Probabilistic analysis of structures involving random stress-strain behavior

    Science.gov (United States)

    Millwater, H. R.; Thacker, B. H.; Harren, S. V.

    1991-01-01

    The present methodology for analysis of structures with random stress strain behavior characterizes the uniaxial stress-strain curve in terms of (1) elastic modulus, (2) engineering stress at initial yield, (3) initial plastic-hardening slope, (4) engineering stress at point of ultimate load, and (5) engineering strain at point of ultimate load. The methodology is incorporated into the Numerical Evaluation of Stochastic Structures Under Stress code for probabilistic structural analysis. The illustrative problem of a thick cylinder under internal pressure, where both the internal pressure and the stress-strain curve are random, is addressed by means of the code. The response value is the cumulative distribution function of the equivalent plastic strain at the inner radius.

  14. User's manuals of probabilistic fracture mechanics analysis code for aged piping, PASCAL-SP

    International Nuclear Information System (INIS)

    Itoh, Hiroto; Nishikawa, Hiroyuki; Onizawa, Kunio; Kato, Daisuke; Osakabe, Kazuya

    2010-03-01

    As a part of research on the material degradation and structural integrity assessment for aged LWR components, a PFM (Probabilistic Fracture Mechanics) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed. This code evaluates the failure probabilities at welded joints of aged piping by a Monte Carlo method. PASCAL-SP treats stress corrosion cracking (SCC) and fatigue crack growth in piping, according to the approaches of NISA and JSME FFS Code. The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the latest knowledge in the SCC assessment and fracture criteria of piping. In addition, the accuracy of flaw detection and sizing at in-service inspection and residual stress distribution were modeled based on experimental data and introduced into PASCAL-SP. This code has been developed for a cross-check use by the regulatory body in Japan. In addition to this, this code can also be used for a research purpose by researchers in academia and industries. This report provides the user's manual and theoretical background of the code. (author)

  15. Civil engineering: calculations of pre-stressed concrete structures using CodeAster

    International Nuclear Information System (INIS)

    Gerard, B.; Ulm, F.

    1997-11-01

    This document presents an analysis of the different calculation methods for pre-stressed concrete structure which can be performed by using finite element methods. Two methods of calculating the pre-stressing of concrete structures with finite elements have been determined. The equivalent method which consists of replacing the action of pre-stressing the concrete by equivalent forces. These method is well suited to dimensioning and studying the overall stability of a structure. It is not an easy matter to take into account the coupled or time-varying phenomena. This approach ignores the evolution of the interaction between the pre-stressing and the concrete. The explicit method which consists of including the mechanical resolution of the pre-stressed cables in that of a concrete structure. Not only does this allow a local study of the pre-stressed to be made, it also allows the coupling which developed over time to be determined, e.g. slip, deferred deformation and coupling between the steel and concrete behaviours. This method enables non-linear phenomena with varying degrees of complexity, such as fracture or yielding of the steels, drying out of the concrete, creep, etc to be described. The two methods are complementary. This document presents the mathematical and computer developments relating to each of this method. In the case of the explicit method, certain of the Code-Aster functions already make it possible to meet several EDF application requirements. Several couplings can be taken into account, such as thermomechanical, shrinkage in drying, creep, relaxation and injection of the cables. Three immediate developments of Code-Aster are proposed for the following applications: - a procedure for calculating the pre-stress losses along the pre-stressing cables; - a command to allocate these forces in the form of an initial force field in the bar elements associated with the cables; - a procedure for linking elements whose nodes do not coincide with each other

  16. Voice stress analysis and evaluation

    Science.gov (United States)

    Haddad, Darren M.; Ratley, Roy J.

    2001-02-01

    Voice Stress Analysis (VSA) systems are marketed as computer-based systems capable of measuring stress in a person's voice as an indicator of deception. They are advertised as being less expensive, easier to use, less invasive in use, and less constrained in their operation then polygraph technology. The National Institute of Justice have asked the Air Force Research Laboratory for assistance in evaluating voice stress analysis technology. Law enforcement officials have also been asking questions about this technology. If VSA technology proves to be effective, its value for military and law enforcement application is tremendous.

  17. Analysis of quantum error-correcting codes: Symplectic lattice codes and toric codes

    Science.gov (United States)

    Harrington, James William

    Quantum information theory is concerned with identifying how quantum mechanical resources (such as entangled quantum states) can be utilized for a number of information processing tasks, including data storage, computation, communication, and cryptography. Efficient quantum algorithms and protocols have been developed for performing some tasks (e.g. , factoring large numbers, securely communicating over a public channel, and simulating quantum mechanical systems) that appear to be very difficult with just classical resources. In addition to identifying the separation between classical and quantum computational power, much of the theoretical focus in this field over the last decade has been concerned with finding novel ways of encoding quantum information that are robust against errors, which is an important step toward building practical quantum information processing devices. In this thesis I present some results on the quantum error-correcting properties of oscillator codes (also described as symplectic lattice codes) and toric codes. Any harmonic oscillator system (such as a mode of light) can be encoded with quantum information via symplectic lattice codes that are robust against shifts in the system's continuous quantum variables. I show the existence of lattice codes whose achievable rates match the one-shot coherent information over the Gaussian quantum channel. Also, I construct a family of symplectic self-dual lattices and search for optimal encodings of quantum information distributed between several oscillators. Toric codes provide encodings of quantum information into two-dimensional spin lattices that are robust against local clusters of errors and which require only local quantum operations for error correction. Numerical simulations of this system under various error models provide a calculation of the accuracy threshold for quantum memory using toric codes, which can be related to phase transitions in certain condensed matter models. I also present

  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. Genetic Code Analysis Toolkit: A novel tool to explore the coding properties of the genetic code and DNA sequences

    Science.gov (United States)

    Kraljić, K.; Strüngmann, L.; Fimmel, E.; Gumbel, M.

    2018-01-01

    The genetic code is degenerated and it is assumed that redundancy provides error detection and correction mechanisms in the translation process. However, the biological meaning of the code's structure is still under current research. This paper presents a Genetic Code Analysis Toolkit (GCAT) which provides workflows and algorithms for the analysis of the structure of nucleotide sequences. In particular, sets or sequences of codons can be transformed and tested for circularity, comma-freeness, dichotomic partitions and others. GCAT comes with a fertile editor custom-built to work with the genetic code and a batch mode for multi-sequence processing. With the ability to read FASTA files or load sequences from GenBank, the tool can be used for the mathematical and statistical analysis of existing sequence data. GCAT is Java-based and provides a plug-in concept for extensibility. Availability: Open source Homepage:http://www.gcat.bio/

  20. Application of coupled codes for safety analysis and licensing issues

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    2006-01-01

    An overview is given on the development and the advantages of coupled codes which integrate 3D neutron kinetics into thermal-hydraulic system codes. The work performed within GRS by coupling the thermal-hydraulic system code ATHLET and the 3D neutronics code QUABOX/CUBBOX is described as an example. The application of the coupled codes as best-estimate simulation tools for safety analysis is discussed. Some examples from German licensing practices are given which demonstrate how the improved analytical methods of coupled codes have contributed to solve licensing issues related to optimized and more economical use of fuel. (authors)

  1. Analytical validation of the CACECO containment analysis code

    International Nuclear Information System (INIS)

    Peak, R.D.

    1979-08-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. This report covers the verification of the CACECO code by problems that can be solved by hand calculations or by reference to textbook and literature examples. The verification concentrates on the accuracy of the material and energy balances maintained by the code and on the independence of the four cells analyzed by the code so that the user can be assured that the code analyses are numerically correct and independent of the organization of the input data submitted to the code

  2. Users' guide to CACECO containment analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.

  3. Verification of the CONPAS (CONtainment Performance Analysis System) code package

    International Nuclear Information System (INIS)

    Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho.

    1997-09-01

    CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs

  4. Uranium Isotopic Analysis with the FRAM Isotopic Analysis Code

    International Nuclear Information System (INIS)

    Vo, D.T.; Sampson, T.E.

    1999-01-01

    FRAM is the acronym for Fixed-Energy Response-Function Analysis with Multiple efficiency. This software was developed at Los Alamos National Laboratory originally for plutonium isotopic analysis. Later, it was adapted for uranium isotopic analysis in addition to plutonium. It is a code based on a self-calibration using several gamma-ray peaks for determining the isotopic ratios. The versatile-parameter database structure governs all facets of the data analysis. User editing of the parameter sets allows great flexibility in handling data with different isotopic distributions, interfering isotopes, and different acquisition parameters such as energy calibration and detector type

  5. Codeword Structure Analysis for LDPC Convolutional Codes

    Directory of Open Access Journals (Sweden)

    Hua Zhou

    2015-12-01

    Full Text Available The codewords of a low-density parity-check (LDPC convolutional code (LDPC-CC are characterised into structured and non-structured. The number of the structured codewords is dominated by the size of the polynomial syndrome former matrix H T ( D , while the number of the non-structured ones depends on the particular monomials or polynomials in H T ( D . By evaluating the relationship of the codewords between the mother code and its super codes, the low weight non-structured codewords in the super codes can be eliminated by appropriately choosing the monomials or polynomials in H T ( D , resulting in improved distance spectrum of the mother code.

  6. The analysis of thermal-hydraulic models in MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-15

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

  7. Concept Analysis: Alzheimer's Caregiver Stress.

    Science.gov (United States)

    Llanque, Sarah; Savage, Lynette; Rosenburg, Neal; Caserta, Michael

    2016-01-01

    The aim of this article was to analyze the concept of caregiver stress in the context of caring for a person with Alzheimer's disease and related dementias. Currently, there are more than 15 million unpaid caregivers for persons suffering from Alzheimer's disease and related dementias. This unpaid care can be stressful for caregivers due to the chronic nature of the disease process, as well as other factors. The paper incorporates the modified method of Wilson's concept analysis procedure to analyze the concept of caregiver stress. A review of the literature was undertaken using the Cumulative Index to Nursing and Allied Health Literature, Google Scholar, and PubMed. A theoretical definition of caregiver stress is provided, and the defining attributes, related concepts, antecedents, and consequences of caregiver stress are proposed, and case studies are presented. The analysis demonstrates that caregiver stress is the unequal exchange of assistance among people who stand in close relationship to one another, which results in emotional and physical stress on the caregiver. Implications for future nursing research and practice conclude the paper. © 2014 Wiley Periodicals, Inc.

  8. PDX vacuum vessel stress analysis

    International Nuclear Information System (INIS)

    Nikodem, Z.D.

    1975-01-01

    A stress analysis of PDX vacuum vessel is described and the summary of results is presented. The vacuum vessel is treated as a toroidal shell of revolution subjected to an internal vacuum. The critical buckling pressure is calculated. The effects of the geometrical discontinuity at the juncture of toroidal shell head and cylindrical outside wall, and the concavity of the cylindrical wall are examined. An effect of the poloidal field coil supports and the vessel outside supports on the stress distribution in the vacuum vessel is determined. A method evaluating the influence of circular ports in the vessel wall on the stress level in the vessel is outlined

  9. Parallelization of Subchannel Analysis Code MATRA

    International Nuclear Information System (INIS)

    Kim, Seongjin; Hwang, Daehyun; Kwon, Hyouk

    2014-01-01

    A stand-alone calculation of MATRA code used up pertinent computing time for the thermal margin calculations while a relatively considerable time is needed to solve the whole core pin-by-pin problems. In addition, it is strongly required to improve the computation speed of the MATRA code to satisfy the overall performance of the multi-physics coupling calculations. Therefore, a parallel approach to improve and optimize the computability of the MATRA code is proposed and verified in this study. The parallel algorithm is embodied in the MATRA code using the MPI communication method and the modification of the previous code structure was minimized. An improvement is confirmed by comparing the results between the single and multiple processor algorithms. The speedup and efficiency are also evaluated when increasing the number of processors. The parallel algorithm was implemented to the subchannel code MATRA using the MPI. The performance of the parallel algorithm was verified by comparing the results with those from the MATRA with the single processor. It is also noticed that the performance of the MATRA code was greatly improved by implementing the parallel algorithm for the 1/8 core and whole core problems

  10. Module type plant system dynamics analysis code (MSG-COPD). Code manual

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2002-11-01

    MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)

  11. OPR1000 RCP Flow Coastdown Analysis using SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong-Hyuk; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Korean nuclear industry developed a thermal-hydraulic analysis code for the safety analysis of PWRs, named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). Current loss of flow transient analysis of OPR1000 uses COAST code to calculate transient RCS(Reactor Coolant System) flow. The COAST code calculates RCS loop flow using pump performance curves and RCP(Reactor Coolant Pump) inertia. In this paper, SPACE code is used to reproduce RCS flowrates calculated by COAST code. The loss of flow transient is transient initiated by reduction of forced reactor coolant circulation. Typical loss of flow transients are complete loss of flow(CLOF) and locked rotor(LR). OPR1000 RCP flow coastdown analysis was performed using SPACE using simplified nodalization. Complete loss of flow(4 RCP trip) was analyzed. The results show good agreement with those from COAST code, which is CE code for calculating RCS flow during loss of flow transients. Through this study, we confirmed that SPACE code can be used instead of COAST code for RCP flow coastdown analysis.

  12. Fast neutron analysis code SAD1

    International Nuclear Information System (INIS)

    Jung, M.; Ott, C.

    1985-01-01

    A listing and an example of outputs of the M.C. code SAD1 are given here. This code has been used many times to predict responses of fast neutrons in hydrogenic materials (in our case emulsions or plastics) towards the elastic n, p scattering. It can be easily extended to other kinds of such materials and to any kind of incident fast neutron spectrum

  13. Code Lavender: Cultivating Intentional Acts of Kindness in Response to Stressful Work Situations.

    Science.gov (United States)

    Davidson, Judy E; Graham, Patricia; Montross-Thomas, Lori; Norcross, William; Zerbi, Giovanna

    Providing healthcare can be stressful. Gone unchecked, clinicians may experience decreased compassion, and increased burnout or secondary traumatic stress. Code Lavender is designed to increase acts of kindness after stressful workplace events occur. To test the feasibility of providing Code Lavender. After stressful events in the workplace, staff will provide, receive, and recommend Code Lavender to others. The provision of Code Lavender will improve Professional Quality of Life Scale (ProQoL) scores, general job satisfaction, and feeling cared for in the workplace. Pilot program testing and evaluation. Staff and physicians on four hospital units were informed of the Code Lavender kit availability, which includes words of comfort, chocolate, lavender essential oil, and employee health referral information. Feasibility data and ProQoL scores were collected at baseline and three months. At baseline, 48% (n = 164) reported a stressful event at work in the last three months. Post-intervention, 51% reported experiencing a stressful workplace event, with 32% receiving a Code Lavender kit from their co-workers as a result (n = 83). Of those who received the Code Lavender intervention; 100% found it helpful, and 84% would recommend it to others. No significant changes were demonstrated before and after the intervention in ProQoL scores or job satisfaction, however the emotion of feeling cared-for improved. Results warrant continuation and further dissemination of Code Lavender. Investigators have received requests to expand the program implying positive reception of the intervention. Additional interventions are needed to overcome workplace stressors. A more intense peer support program is being tested. Copyright © 2017. Published by Elsevier Inc.

  14. Stress analysis and evaluation of a rectangular pressure vessel

    International Nuclear Information System (INIS)

    Rezvani, M.A.; Ziada, H.H.; Shurrab, M.S.

    1992-10-01

    This study addresses structural analysis and evaluation of an abnormal rectangular pressure vessel, designed to house equipment for drilling and collecting samples from Hanford radioactive waste storage tanks. It had to be qualified according to ASME boiler and pressure vessel code, Section VIII; however, it had the cover plate bolted along the long face, a configuration not addressed by the code. Finite element method was used to calculate stresses resulting from internal pressure; these stresses were then used to evaluate and qualify the vessel. Fatigue is not a concern; thus, it can be built according to Section VIII, Division I instead of Division 2. Stress analysis was checked against the code. A stayed plate was added to stiffen the long side of the vessel

  15. Configuration analysis of pipe support for primary cooling using Ps + Caepipe code

    International Nuclear Information System (INIS)

    Sitandung, Y. B.; Pustandyo, W.; Sujalmo, S.

    1998-01-01

    Pipe stress evaluation and support loads has been analyzed on piping segment of RSG-GAS primary cooling system. This paper describes an analysis method of piping system with the use of computer Code PS + CAEPIPE Version 3.4.05.W. From the selected pipe segment, the data of pipe characteristic, material properties, operation condition, equipment and supports were used input. The final evaluation result of primary cooling pipe segment show that actual stress dead weight and seismic load are less than allowable limits (stress ratio 0.101 for deadweight 0.35 for seismic load). From the above ratio, it can be concluded that ratio of pipe support configuration to stress distribution is acceptable, and based on analysis result, the Code used by INTERATOM was sufficiently accurate

  16. Challenges on innovations of newly-developed safety analysis codes

    International Nuclear Information System (INIS)

    Yang, Yanhua; Zhang, Hao

    2016-01-01

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  17. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  18. 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)

  19. Long Non-coding RNAs in Response to Genotoxic Stress

    Institute of Scientific and Technical Information of China (English)

    Xiaoman Li; Dong Pan; Baoquan Zhao; Burong Hu

    2016-01-01

    Long non-coding RNAs(lncRNAs) are increasingly involved in diverse biological processes.Upon DNA damage,the DNA damage response(DDR) elicits a complex signaling cascade,which includes the induction of lncRNAs.LncRNA-mediated DDR is involved in non-canonical and canonical manners.DNA-damage induced lncRNAs contribute to the regulation of cell cycle,apoptosis,and DNA repair,thereby playing a key role in maintaining genome stability.This review summarizes the emerging role of lncRNAs in DNA damage and repair.

  20. Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Han, Kee Soo; Moon, Bok Ja; Jang, Misuk [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    In this presentation, simple thermal hydraulic behavior is analyzed using three codes to show the possibility of using alternative codes. We established three models of simple u-tube manometer using three different codes. CATHENA (Canadian Algorithm for Thermal hydraulic Network Analysis), RELAP (Reactor Excursion and Leak Analysis Program), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are used for this analysis. CATHENA and RELAP are widely used codes for the analysis of system behavior of CANDU and PWR. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. In this paper, the internal behavior of u-tube manometer was analyzed using 3 codes, CATHENA, RELAP and GOTHIC. The general transient behavior is similar among 3 codes. However, the behavior simulated using GOTHIC shows some different trend compared with the results from the other 2 codes at the end of the transient. It would be resulted from the use of different physical model in GOTHIC, which is specialized for the multi-phase thermal hydraulic behavior analysis of containment system unlike the other two codes.

  1. Content Analysis Coding Schemes for Online Asynchronous Discussion

    Science.gov (United States)

    Weltzer-Ward, Lisa

    2011-01-01

    Purpose: Researchers commonly utilize coding-based analysis of classroom asynchronous discussion contributions as part of studies of online learning and instruction. However, this analysis is inconsistent from study to study with over 50 coding schemes and procedures applied in the last eight years. The aim of this article is to provide a basis…

  2. A Semantic Analysis Method for Scientific and Engineering Code

    Science.gov (United States)

    Stewart, Mark E. M.

    1998-01-01

    This paper develops a procedure to statically analyze aspects of the meaning or semantics of scientific and engineering code. The analysis involves adding semantic declarations to a user's code and parsing this semantic knowledge with the original code using multiple expert parsers. These semantic parsers are designed to recognize formulae in different disciplines including physical and mathematical formulae and geometrical position in a numerical scheme. In practice, a user would submit code with semantic declarations of primitive variables to the analysis procedure, and its semantic parsers would automatically recognize and document some static, semantic concepts and locate some program semantic errors. A prototype implementation of this analysis procedure is demonstrated. Further, the relationship between the fundamental algebraic manipulations of equations and the parsing of expressions is explained. This ability to locate some semantic errors and document semantic concepts in scientific and engineering code should reduce the time, risk, and effort of developing and using these codes.

  3. Subchannel analysis code development for CANDU fuel channel

    International Nuclear Information System (INIS)

    Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

    1998-07-01

    Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

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

    International Nuclear Information System (INIS)

    Engelman, G.P.

    1978-10-01

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

  5. Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1983-01-01

    In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)

  6. Evaluation of the DRAGON code for VHTR design analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-12

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.

  7. Evaluation of the DRAGON code for VHTR design analysis

    International Nuclear Information System (INIS)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-01

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR

  8. Verification of thermal-irradiation stress analytical code VIENUS of graphite block

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Shiozawa, Shusaku; Shirai, Hiroshi; Minato, Kazuo.

    1992-02-01

    The core graphite components of the High Temperature Engineering Test Reactor (HTTR) show both the dimensional change (irradiation shrinkage) and creep behavior due to fast neutron irradiation under the temperature and the fast neutron irradiation conditions of the HTTR. Therefore, thermal/irradiation stress analytical code, VIENUS, which treats these graphite irradiation behavior, is to be employed in order to design the core components such as fuel block etc. of the HTTR. The VIENUS is a two dimensional finite element viscoelastic stress analytical code to take account of changes in mechanical properties, thermal strain, irradiation-induced dimensional change and creep in the fast neutron irradiation environment. Verification analyses were carried out in order to prove the validity of this code based on the irradiation tests of the 8th OGL-1 fuel assembly and the fuel element of the Peach Bottom reactor. This report describes the outline of the VIENUS code and its verification analyses. (author)

  9. Stress analysis of the tokamak engineering test breeder blanket

    International Nuclear Information System (INIS)

    Huang Zhongqi

    1992-01-01

    The design features of the hybrid reactor blanket and main parameters are presented. The stress analysis is performed by using computer codes SAP5p and SAP6 with the three kinds of blanket module loadings, i.e, the pressure of coolant, the blanket weight and the thermal loading. Numerical calculation results indicate that the stresses of the blanket are smaller than the allowable ones of the material, the blanket design is therefore reasonable

  10. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  11. 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

  12. 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

  13. Light water reactor fuel analysis code FEMAXI-IV(Ver.2). Detailed structure and user's manual

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Saitou, Hiroaki.

    1997-11-01

    A light water reactor fuel behavior analysis code FEMAXI-IV(Ver.2) was developed as an improved version of FEMAXI-IV. Development of FEMAXI-IV has been already finished in 1992, though a detailed structure and input manual of the code have not been open to users yet. Here, the basic theories and structure, the models and numerical solutions applied to FEMAXI-IV(Ver.2), and the material properties adopted in the code are described in detail. In FEMAXI-IV(Ver.2), programming bugs in previous FEMAXI-IV were eliminated, renewal of the pellet thermal conductivity was performed, and a model of thermal-stress restraint on FP gas release was incorporated. For facilitation of effective and wide-ranging application of the code, methods of input/output of the code are also described in detail, and sample output is included. (author)

  14. Development of the integrated system reliability analysis code MODULE

    International Nuclear Information System (INIS)

    Han, S.H.; Yoo, K.J.; Kim, T.W.

    1987-01-01

    The major components in a system reliability analysis are the determination of cut sets, importance measure, and uncertainty analysis. Various computer codes have been used for these purposes. For example, SETS and FTAP are used to determine cut sets; Importance for importance calculations; and Sample, CONINT, and MOCUP for uncertainty analysis. There have been problems when the codes run each other and the input and output are not linked, which could result in errors when preparing input for each code. The code MODULE was developed to carry out the above calculations simultaneously without linking input and outputs to other codes. MODULE can also prepare input for SETS for the case of a large fault tree that cannot be handled by MODULE. The flow diagram of the MODULE code is shown. To verify the MODULE code, two examples are selected and the results and computation times are compared with those of SETS, FTAP, CONINT, and MOCUP on both Cyber 170-875 and IBM PC/AT. Two examples are fault trees of the auxiliary feedwater system (AFWS) of Korea Nuclear Units (KNU)-1 and -2, which have 54 gates and 115 events, 39 gates and 92 events, respectively. The MODULE code has the advantage that it can calculate the cut sets, importances, and uncertainties in a single run with little increase in computing time over other codes and that it can be used in personal computers

  15. Development of the next generation reactor analysis code system, MARBLE

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

    2011-03-01

    A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

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

    Science.gov (United States)

    Bari, Md. S.; Das, T.

    2013-09-01

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

  17. Analysis and application of ratcheting evaluation procedure of Japanese high temperature design code DDS

    International Nuclear Information System (INIS)

    Lee, H. Y.; Kim, J. B.; Lee, J. H.

    2002-01-01

    In this study, the evaluation procedure of Japanese DDS code which was recently developed to assess the progressive inelastic deformation occurring under repetition of secondary stresses was analyzed and the evaluation results according to DDS was compared those of the thermal ratchet structural test carried out by KAERI to analyze the conservativeness of the code. The existing high temperature codes of US ASME-NH and French RCC-MR suggest the limited ratcheting procedures for only the load cases of cyclic secondary stresses under primary stresses. So they are improper to apply to the actual ratcheting problem which can occur under cyclic secondary membrane stresses due to the movement of hot free surface for the pool type LMR. DDS provides explicitly an analysis procedure of ratcheting due to moving thermal gradients near hot free surface. A comparison study was carried out between the results by the design code of DDS and by the structural test to investigate the conservativeness of DDS code, which showed that the evaluation results by DDS were in good agreement with those of the structural test

  18. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

  19. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, B. D; Kim, K. D.

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

  20. Simulation of Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes

    Science.gov (United States)

    2015-11-01

    Memorandum Simulation of Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes...Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes by Charles R. Fisher...Welding- Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes 5a. CONTRACT NUMBER N/A 5b. GRANT NUMBER N/A 5c

  1. A code for structural analysis of fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, I.M.V.; Perrotta, J.A.

    1988-08-01

    It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each tubular structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equilavent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt

  2. 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)

  3. 76 FR 57982 - Building Energy Codes Cost Analysis

    Science.gov (United States)

    2011-09-19

    ... DEPARTMENT OF ENERGY Office of Energy Efficiency and Renewable Energy [Docket No. EERE-2011-BT-BC-0046] Building Energy Codes Cost Analysis Correction In notice document 2011-23236 beginning on page...-23236 Filed 9-16-11; 8:45 am] BILLING CODE 1505-01-P ...

  4. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  5. 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)

  6. 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).

  7. User's manual for seismic analysis code 'SONATINA-2V'

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Iyoku, Tatsuo

    2001-08-01

    The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

  8. The Role of Non-­Coding RNA in Plant Stress

    KAUST Repository

    MacPherson, Cameron R.

    2012-12-01

    Post-transcriptional gene silencing (PTGS) is a powerful mechanism that can be adapted to genetically modify crop plants. PTGS operates in many plant signaling pathways including those mediating stress responses. Given the small number of miRNAs known, research on the characterization of stress-related micro-RNA (miRNA) and their targets could provide the basis for engineering stress tolerant traits in crops. Indeed, several examples of miRNA mediated crop tolerance have been reported. In the research presented here, we aimed to analyze the role of small non-coding RNA (smRNA) pathways involved in plant stress. In particular, we focused on miRNA-mediated PTGS in phosphate (Pi) starvation. The analysis was split into two research projects. First, to identify potential miRNA targets we began by analyzing the response and recovery of coding and long non-coding RNAs (lncRNA) to Pi starvation in shoot and root. The results obtained were the first genome-wide description of the root’s Pi starvation response and recovery. We found that the root\\'s response involved a widely different set of genes than that of the shoot. In the second research project, the results of the first project were correlated with the responses of miRNA and trans-acting small-interfering RNA (tasiRNA) during Pi starvation. Many miRNA circuits have been predicted before, however, tasiRNA circuits are not as well defined. Therefore, we made use of the double-stranded RNA-binding protein 4 (DRB4) smRNA libraries to enhance our prediction of tasiRNAs. Altogether, we provided evidence to support the following miRNA-mRNA pairs that may function in Pi starvation: IPS1:miR399:PHO2; miR399:RS4; miR399:NF-YA10; miR398:CSD1/2; miR2111:TPS11; miR164:NAC6; miR157:TMO7; miR157:PSB28; RPS2:miR169:IPS2; miR397:LAC2; TAS4:PAP1; NR1:PAP1; and Chr3_1967672:TMO7. In general, we found that non-miR399 related circuits were active only during the root’s recovery from Pi starvation. The functional roles of the genes

  9. Production of analysis code for 'JOYO' dosimetry experiment

    International Nuclear Information System (INIS)

    Sasaki, Makoto; Nakazawa, Masaharu.

    1981-01-01

    As part of the measurement and analysis plan for the Dosimetry Experiment at the ''JOYO'' experimental fast reactor, neutron flux spectra analysis is performed using the NEUPAC (Neutron Unfolding Code Package) computer program. The code calculates the neutron flux spectra and other integral quantities from the activation data of the dosimeter foils. The NEUPAC code is based on the J1-type unfolding method, and the estimated neutron flux spectra is obtained as its solution. The program is able to determine the integral quantities and their sensitivities, together with an error estimate of the unfolded spectra and integral quantities. The code also performs a chi-square test of the input/output data, and contains many options for the calculational routines. This report presents the analytic theory, the program algorithms, and a description of the functions and use of the NEUPAC code. (author)

  10. 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)

  11. Transient and fuel performance analysis with VTT's coupled code system

    International Nuclear Information System (INIS)

    Daavittila, A.; Hamalainen, A.; Raty, H.

    2005-01-01

    VTT (technical research center of Finland) maintains and further develops a comprehensive safety analysis code system ranging from the basic neutronic libraries to 3-dimensional transient analysis and fuel behaviour analysis codes. The code system is based on various types of couplings between the relevant physical phenomena. The main tools for analyses of reactor transients are presently the 3-dimensional reactor dynamics code HEXTRAN for cores with a hexagonal fuel assembly geometry and TRAB-3D for cores with a quadratic fuel assembly geometry. HEXTRAN has been applied to safety analyses of VVER type reactors since early 1990's. TRAB-3D is the latest addition to the code system, and has been applied to BWR and PWR analyses in recent years. In this paper it is shown that TRAB-3D has calculated accurately the power distribution during the Olkiluoto-1 load rejection test. The results from the 3-dimensional analysis can be used as boundary conditions for more detailed fuel rod analysis. For this purpose a general flow model GENFLO, developed at VTT, has been coupled with USNRC's FRAPTRAN fuel accident behaviour model. The example case for FRAPTRAN-GENFLO is for an ATWS at a BWR plant. The basis for the analysis is an oscillation incident in the Olkiluoto-1 BWR during reactor startup on February 22, 1987. It is shown that the new coupled code FRAPTRAN/GENFLO is quite a promising tool that can handle flow situations and give a detailed analysis of reactor transients

  12. Sensitivity analysis of the RESRAD, a dose assessment code

    International Nuclear Information System (INIS)

    Yu, C.; Cheng, J.J.; Zielen, A.J.

    1991-01-01

    The RESRAD code is a pathway analysis code that is designed to calculate radiation doses and derive soil cleanup criteria for the US Department of Energy's environmental restoration and waste management program. the RESRAD code uses various pathway and consumption-rate parameters such as soil properties and food ingestion rates in performing such calculations and derivations. As with any predictive model, the accuracy of the predictions depends on the accuracy of the input parameters. This paper summarizes the results of a sensitivity analysis of RESRAD input parameters. Three methods were used to perform the sensitivity analysis: (1) Gradient Enhanced Software System (GRESS) sensitivity analysis software package developed at oak Ridge National Laboratory; (2) direct perturbation of input parameters; and (3) built-in graphic package that shows parameter sensitivities while the RESRAD code is operational

  13. PACC information management code for common cause failures analysis

    International Nuclear Information System (INIS)

    Ortega Prieto, P.; Garcia Gay, J.; Mira McWilliams, J.

    1987-01-01

    The purpose of this paper is to present the PACC code, which, through an adequate data management, makes the task of computerized common-mode failure analysis easier. PACC processes and generates information in order to carry out the corresponding qualitative analysis, by means of the boolean technique of transformation of variables, and the quantitative analysis either using one of several parametric methods or a direct data-base. As far as the qualitative analysis is concerned, the code creates several functional forms for the transformation equations according to the user's choice. These equations are subsequently processed by boolean manipulation codes, such as SETS. The quantitative calculations of the code can be carried out in two different ways: either starting from a common cause data-base, or through parametric methods, such as the Binomial Failure Rate Method, the Basic Parameters Method or the Multiple Greek Letter Method, among others. (orig.)

  14. FEAST: a two-dimensional non-linear finite element code for calculating stresses

    International Nuclear Information System (INIS)

    Tayal, M.

    1986-06-01

    The computer code FEAST calculates stresses, strains, and displacements. The code is two-dimensional. That is, either plane or axisymmetric calculations can be done. The code models elastic, plastic, creep, and thermal strains and stresses. Cracking can also be simulated. The finite element method is used to solve equations describing the following fundamental laws of mechanics: equilibrium; compatibility; constitutive relations; yield criterion; and flow rule. FEAST combines several unique features that permit large time-steps in even severely non-linear situations. The features include a special formulation for permitting many finite elements to simultaneously cross the boundary from elastic to plastic behaviour; accomodation of large drops in yield-strength due to changes in local temperature and a three-step predictor-corrector method for plastic analyses. These features reduce computing costs. Comparisons against twenty analytical solutions and against experimental measurements show that predictions of FEAST are generally accurate to ± 5%

  15. Adaptable Value-Set Analysis for Low-Level Code

    OpenAIRE

    Brauer, Jörg; Hansen, René Rydhof; Kowalewski, Stefan; Larsen, Kim G.; Olesen, Mads Chr.

    2012-01-01

    This paper presents a framework for binary code analysis that uses only SAT-based algorithms. Within the framework, incremental SAT solving is used to perform a form of weakly relational value-set analysis in a novel way, connecting the expressiveness of the value sets to computational complexity. Another key feature of our framework is that it translates the semantics of binary code into an intermediate representation. This allows for a straightforward translation of the program semantics in...

  16. Improvement of QR Code Recognition Based on Pillbox Filter Analysis

    Directory of Open Access Journals (Sweden)

    Jia-Shing Sheu

    2013-04-01

    Full Text Available The objective of this paper is to perform the innovation design for improving the recognition of a captured QR code image with blur through the Pillbox filter analysis. QR code images can be captured by digital video cameras. Many factors contribute to QR code decoding failure, such as the low quality of the image. Focus is an important factor that affects the quality of the image. This study discusses the out-of-focus QR code image and aims to improve the recognition of the contents in the QR code image. Many studies have used the pillbox filter (circular averaging filter method to simulate an out-of-focus image. This method is also used in this investigation to improve the recognition of a captured QR code image. A blurred QR code image is separated into nine levels. In the experiment, four different quantitative approaches are used to reconstruct and decode an out-of-focus QR code image. These nine reconstructed QR code images using methods are then compared. The final experimental results indicate improvements in identification.

  17. 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

  18. 14 CFR 33.62 - Stress analysis.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Stress analysis. 33.62 Section 33.62... STANDARDS: AIRCRAFT ENGINES Design and Construction; Turbine Aircraft Engines § 33.62 Stress analysis. A stress analysis must be performed on each turbine engine showing the design safety margin of each turbine...

  19. Numerical analysis of interacting cracks in biaxial stress field

    International Nuclear Information System (INIS)

    Kovac, M.; Cizelj, L.

    1999-01-01

    The stress corrosion cracks as seen for example in PWR steam generator tubing made of Inconel 600 usually produce highly irregular kinked and branched crack patterns. Crack initialization and propagation depends on stress state underlying the crack pattern. Numerical analysis (such as finite element method) of interacting kinked and branched cracks can provide accurate solutions. This paper discusses the use of general-purpose finite element code ABAQUS for evaluating stress fields at crack tips of interacting complex cracks. The results obtained showed reasonable agreement with the reference solutions and confirmed use of finite elements in such class of problems.(author)

  20. Developments of fuel performance analysis codes in KEPCO NF

    International Nuclear Information System (INIS)

    Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

    2012-01-01

    The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

  1. 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.)

  2. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

    Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

    1997-07-01

    During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

  3. Code Coupling for Multi-Dimensional Core Transient Analysis

    International Nuclear Information System (INIS)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

    2015-01-01

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

  4. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  5. 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

  6. 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.

  7. 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

  8. Multi-dimensional Code Development for Safety Analysis of LMR

    International Nuclear Information System (INIS)

    Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.

    2006-08-01

    A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150

  9. Guide to Using Onionskin Analysis Code (U)

    Energy Technology Data Exchange (ETDEWEB)

    Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group; Morzinski, Jerome Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group

    2016-09-15

    This document is a guide to using R-code written for the purpose of analyzing onionskin experiments. We expect the user to be very familiar with statistical methods and the R programming language. For more details about onionskin experiments and the statistical methods mentioned in this document see Storlie, Fugate, et al. (2013). Engineers at LANL experiment with detonators and high explosives to assess performance. The experimental unit, called an onionskin, is a hemisphere consisting of a detonator and a booster pellet surrounded by explosive material. When the detonator explodes, a streak camera mounted above the pole of the hemisphere records when the shock wave arrives at the surface. The output from the camera is a two-dimensional image that is transformed into a curve that shows the arrival time as a function of polar angle. The statistical challenge is to characterize a baseline population of arrival time curves and to compare the baseline curves to curves from a new, so-called, test series. The hope is that the new test series of curves is statistically similar to the baseline population.

  10. Impact testing and analysis for structural code benchmarking

    International Nuclear Information System (INIS)

    Glass, R.E.

    1989-01-01

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

  11. Experimental transport analysis code system in JT-60

    International Nuclear Information System (INIS)

    Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru

    1988-03-01

    Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)

  12. Color Coding of Circuit Quantities in Introductory Circuit Analysis Instruction

    Science.gov (United States)

    Reisslein, Jana; Johnson, Amy M.; Reisslein, Martin

    2015-01-01

    Learning the analysis of electrical circuits represented by circuit diagrams is often challenging for novice students. An open research question in electrical circuit analysis instruction is whether color coding of the mathematical symbols (variables) that denote electrical quantities can improve circuit analysis learning. The present study…

  13. Multi-pressure boiler thermodynamics analysis code

    International Nuclear Information System (INIS)

    Lorenzoni, G.

    1992-01-01

    A new method and the relative FORTRAN program for the thermodynamics design analysis of a multipressure boiler are reported. This method permits the thermodynamics design optimization with regard to total exergy production and a preliminary costs

  14. Advanced Techniques of Stress Analysis

    Directory of Open Access Journals (Sweden)

    Simion TATARU

    2013-12-01

    Full Text Available This article aims to check the stress analysis technique based on 3D models also making a comparison with the traditional technique which utilizes a model built directly into the stress analysis program. This comparison of the two methods will be made with reference to the rear fuselage of IAR-99 aircraft, structure with a high degree of complexity which allows a meaningful evaluation of both approaches. Three updated databases are envisaged: the database having the idealized model obtained using ANSYS and working directly on documentation, without automatic generation of nodes and elements (with few exceptions, the rear fuselage database (performed at this stage obtained with Pro/ ENGINEER and the one obtained by using ANSYS with the second database. Then, each of the three databases will be used according to arising necessities.The main objective is to develop the parameterized model of the rear fuselage using the computer aided design software Pro/ ENGINEER. A review of research regarding the use of virtual reality with the interactive analysis performed by the finite element method is made to show the state- of- the-art achieved in this field.

  15. Analysis of Iterated Hard Decision Decoding of Product Codes with Reed-Solomon Component Codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2007-01-01

    Products of Reed-Solomon codes are important in applications because they offer a combination of large blocks, low decoding complexity, and good performance. A recent result on random graphs can be used to show that with high probability a large number of errors can be corrected by iterating...... minimum distance decoding. We present an analysis related to density evolution which gives the exact asymptotic value of the decoding threshold and also provides a closed form approximation to the distribution of errors in each step of the decoding of finite length codes....

  16. Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

    International Nuclear Information System (INIS)

    Hong, S.Y.; Yeater, M.L.

    1985-01-01

    This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A ''analysis of flaw indication'' for the application to a PWR primary piping. Results of the analysis are discussed in detail. (orig.)

  17. 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.)

  18. Stress analysis of HLW containers. Compas project

    International Nuclear Information System (INIS)

    1989-01-01

    This document reports the work carried out for the Compas project which looked at the performance of various computer codes in a selected benchmark exercise. This exercise consisted of several analyses on simplified models which have features typical of HLW containers. These analyses comprise two groups; one related to thick walled, stressed shell overpacks, the other related to thin walled, supported shell overpacks with a lead filler. The first set of analyses looked at an elastic-plastic behaviour and large deformation of a cylinder representative of the main body of thick walled containers). The second set looked at creep behaviour of the lead filler, and the shape the base of thin walled containers will take up, after hundreds of years in the repository. On the thick walled analyses with the cylinder subject to an external pressure all the codes gave consistent results in the elastic region and there is good agreement in the yield pressures. Once in the plastic region there is more divergence in the results although a consistent trend is predicted. One of the analyses predicted a non-axisymmetric mode of deformation as would be expected in reality. Fewer results were received for the creep analysis, however the transient creep results showed consistency, and were bounded by the final-state results

  19. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  20. Development and validation of sodium fire analysis code ASSCOPS

    International Nuclear Information System (INIS)

    Ohno, Shuji

    2001-01-01

    A version 2.1 of the ASSCOPS sodium fire analysis code was developed to evaluate the thermal consequences of a sodium leak and consequent fire in LMFBRs. This report describes the computational models and the validation studies using the code. The ASSCOPS calculates sodium droplet and pool fire, and consequential heat/mass transfer behavior. Analyses of sodium pool or spray fire experiments confirmed that this code and parameters used in the validation studies gave valid results on the thermal consequences of sodium leaks and fires. (author)

  1. A static analysis tool set for assembler code verification

    International Nuclear Information System (INIS)

    Dhodapkar, S.D.; Bhattacharjee, A.K.; Sen, Gopa

    1991-01-01

    Software Verification and Validation (V and V) is an important step in assuring reliability and quality of the software. The verification of program source code forms an important part of the overall V and V activity. The static analysis tools described here are useful in verification of assembler code. The tool set consists of static analysers for Intel 8086 and Motorola 68000 assembly language programs. The analysers examine the program source code and generate information about control flow within the program modules, unreachable code, well-formation of modules, call dependency between modules etc. The analysis of loops detects unstructured loops and syntactically infinite loops. Software metrics relating to size and structural complexity are also computed. This report describes the salient features of the design, implementation and the user interface of the tool set. The outputs generated by the analyser are explained using examples taken from some projects analysed by this tool set. (author). 7 refs., 17 figs

  2. Fault tree analysis. Implementation of the WAM-codes

    International Nuclear Information System (INIS)

    Bento, J.P.; Poern, K.

    1979-07-01

    The report describes work going on at Studsvik at the implementation of the WAM code package for fault tree analysis. These codes originally developed under EPRI contract by Sciences Applications Inc, allow, in contrast with other fault tree codes, all Boolean operations, thus allowing modeling of ''NOT'' conditions and dependent components. To concretize the implementation of these codes, the auxiliary feed-water system of the Swedish BWR Oskarshamn 2 was chosen for the reliability analysis. For this system, both the mean unavailability and the probability density function of the top event - undesired event - of the system fault tree were calculated, the latter using a Monte-Carlo simulation technique. The present study is the first part of a work performed under contract with the Swedish Nuclear Power Inspectorate. (author)

  3. Code conversion for system design and safety analysis of NSSS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae Cho; Kim, Young Tae; Choi, Young Gil; Kim, Hee Kyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    This report describes overall project works related to conversion, installation and validation of computer codes which are used in NSSS design and safety analysis of nuclear power plants. Domain/os computer codes for system safety analysis are installed and validated on Apollo DN10000, and then Apollo version are converted and installed again on HP9000/700 series with appropriate validation. Also, COOLII and COAST which are cyber version computer codes are converted into versions of Apollo DN10000 and HP9000/700, and installed with validation. This report details whole processes of work involved in the computer code conversion and installation, as well as software verification and validation results which are attached to this report. 12 refs., 8 figs. (author)

  4. Practical stress analysis in engineering design

    CERN Document Server

    Huston, Ronald

    2008-01-01

    Presents the application of engineering design and analysis based on the approach of understanding the physical characteristics of a given problem and then modeling the important aspects of the physical system. This book covers such topics as contact stress analysis, singularity functions, gear stresses, fasteners, shafts, and shaft stresses.

  5. Rotors stress analysis and design

    CERN Document Server

    Vullo, Vincenzo

    2013-01-01

    Stress and strain analysis of rotors subjected to surface and body loads, as well as to thermal loads deriving from temperature variation along the radius, constitutes a classic subject of machine design. Nevertheless attention is limited to rotor profiles for which governing equations are solvable in closed form. Furthermore very few actual engineering issues may relate to structures for which stress and strain analysis in the linear elastic field and, even more, under non-linear conditions (i.e. plastic or viscoelastic conditions) produces equations to be solved in closed form. Moreover, when a product is still in its design stage, an analytical formulation with closed-form solution is of course simpler and more versatile than numerical methods, and it allows to quickly define a general configuration, which may then be fine-tuned using such numerical methods. In this view, all subjects are based on analytical-methodological approach, and some new solutions in closed form are presented. The analytical formul...

  6. SALT4: a two-dimensional displacement discontinuity code for thermomechanical analysis in bedded salt deposits

    International Nuclear Information System (INIS)

    1983-04-01

    SALT4 is a two-dimensional analytical/displacement-discontinuity code designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. This code was developed by the University of Minnesota. This documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of the computer code, SALT4. The SALT4 code takes into account: (1) viscoelastic behavior in the pillars adjacent to excavations; (2) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (2) excavation sequence. Major advantages of the SALT4 code are: (1) computational efficiency; (2) the small amount of input data required; and (3) a creep law consistent with laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of SALT4, i.e., temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT4 code can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermal and thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT4 can also be used to verify fully numerical codes. This is similar to the use of analytic solutions for code verification. Although SALT4 was designed for analysis of bedded salt, it is also applicable to crystalline rock if the creep calculation is suppressed. In Section 1.5 of this document the code custodianship and control is described along with the status of verification, validation and peer review of this report

  7. Development and improvement of safety analysis code for geological disposal

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)

  8. Pipe whip analysis using the TEDEL code

    International Nuclear Information System (INIS)

    Millard, D.; Hoffmann, A.

    1985-02-01

    In view of their abundance, piping systems are one of the main components in power industries and in particular in nuclear power plants. They must be designed for normal as well as faulted conditions, for safety requirements. The prediction of the dynamic behaviour of the free pipe requires accounting for several nonlinearities. For this purpose, a beam type finite element program (TEDEL) has been used. The aim of this paper is to enlight the main features of this program, when applied to pipe whip analysis. An example of application to a real case will also be presented

  9. Development of the versatile reactor analysis code system, MARBLE2

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi

    2015-07-01

    The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)

  10. FEA stress analysis for SAFKEG 2863B

    International Nuclear Information System (INIS)

    Puckett, A.

    1997-01-01

    This report covers the evaluation of the structural design of the two stainless steel containment vessels in CROFT SAFKEG Model Number 2863B, for conformance to the design criteria of the NRC Regulatory Guide 7.6, NRC Regulatory Guide 7.8, and the applicable requirements of the ASME Boiler and Pressure Vessel Code, Section 3, and Section 8. The two containment vessels are designated Cans 2870 and 2871. Each of these containment vessels was analyzed for the loadings specified in chapter 2, Section 2.1.2 of the SARP. Structural assessment of Cans 2870 and 2871 due to loading considerations beyond the evaluation of pressure and temperature are presented. This report is organized as follows: (1) overview of the design of each containment vessel and pressure boundary; (2) brief description of both containment vessels; (3) discussion of normal and accident conditions; (4) analysis assumptions; (5) detailed structural evaluation of each component of each containment vessel; (6) demonstration of compliance to Regulatory Guide 7.6 stress evaluations; (7) demonstration of compliance to Regulatory Guide 7.8 loading combinations; and (8) summary of the calculated stresses, comparison with design allowables, estimates of margins of safety and a summary of results and conclusions

  11. Joining U.S. NRC international round robin for weld residual stress analysis. Stress analysis and validation in PWSCC mitigation program

    International Nuclear Information System (INIS)

    Maekawa, Akira; Serizawa, Hisashi; Murakawa, Hidekazu

    2012-01-01

    It is necessary to establish properly reliable weld residual stress analysis methods for accurate crack initiation and growth assessment of primary water stress corrosion cracking (PWSCC), which may occur in nickel-based dissimilar metal welds in pressurized water reactors. The U.S. Nuclear Regulatory Commission conducted an international round robin for weld residual stress analysis to improve stress analysis methods and to examine the uncertainties involved in the calculated stress values. In this paper, the results from the authors' participation in the round robin were reported. In the round robin, the weld residual stress in a nickel-based dissimilar metal weld of a pressurizer surge nozzle mock-up was computed under various analysis conditions. Based on these residual stress analysis results, a welding simulation code currently being developed that uses the iterative substructure method was validated and affecting factors on the analysis results were identified. (author)

  12. 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

  13. PCLOOK: an interactive code for spectral analysis

    International Nuclear Information System (INIS)

    Macchiavelli, A.O.; Tomasi, D.

    1993-01-01

    The present work describes an interactive programme for the analysis of spectra developed to run in a PC platform. PCLOOK has a graphic interface that allows the user to get access to different functions using the mouse or directly typing commands. In this way one can switch to a suitable required environment to manage the histograms reassembling in this way a spectrum calculator.The PCLOOK programme was mainly developed to use in nuclear physics applications, but it is also possible to modify it with relative little effort to adapt it to other applications. It was written in Microsoft's BASIC 7.1 installed in a 33MHz 486 Everex PC. For a proper operation an ordinary VGA display and mouse are needed. The memory requirements depend on the size and number of the user defined spectra; for instance, for twenty 2048 channels spectra the available memory space must be 320 KBytes. (author). 5 figs

  14. Impact testing and analysis for structural code benchmarking

    International Nuclear Information System (INIS)

    Glass, R.E.

    1989-01-01

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

  15. 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

  16. NACOM - a code for sodium spray fire analysis

    International Nuclear Information System (INIS)

    Rao, P.M.; Kannan, S.E.

    2002-01-01

    Full text: In liquid metal fast breeder reactors (LMFBR), leakage of sodium can result in a spray fire. Because of higher burning rates in droplet form combustion of sodium in spray fire, thermal consequences are more severe than that in a sodium pool fire. The code NACOM was developed for the analysis of sodium spray fires in LMFBRs facilities. The code uses the validated model for estimating the falling droplet burning rates in pre-ignition and vapour phase combustion stages. It uses a distribution system to generate the droplet groups of different diameters that represent the spray. The code requires about 20 input parameters like sodium leak rates, sodium temperature, initial cell conditions like oxygen concentration, temperature and dimensions. NACOM is a validated code based on experiments with sodium inventory up to 650 kg in 0 to 21 % O 2 atmospheres. The paper brings out the salient features of the code along with the sensitivity analysis of the main input parameters like spray volume mean diameter, oxygen concentration etc. based on the results obtained. The limitations of the code and the confidence margins applicable to results obtained are also brought out

  17. NORTICA—a new code for cyclotron analysis

    Science.gov (United States)

    Gorelov, D.; Johnson, D.; Marti, F.

    2001-12-01

    The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER [1] developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state.

  18. Demonstration study on shielding safety analysis code (VI)

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  19. Two-phase flow characteristics analysis code: MINCS

    International Nuclear Information System (INIS)

    Watanabe, Tadashi; Hirano, Masashi; Akimoto, Masayuki; Tanabe, Fumiya; Kohsaka, Atsuo.

    1992-03-01

    Two-phase flow characteristics analysis code: MINCS (Modularized and INtegrated Code System) has been developed to provide a computational tool for analyzing two-phase flow phenomena in one-dimensional ducts. In MINCS, nine types of two-phase flow models-from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model-can be used under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected. It is, therefore, easy to calculate physical or numerical benchmark problems. (author)

  20. A development of containment performance analysis methodology using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. C.; Yoon, J. I. [Future and Challenge Company, Seoul (Korea, Republic of); Byun, C. S.; Lee, J. Y. [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Lee, J. Y. [Seoul National University, Seoul (Korea, Republic of)

    2003-10-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code.

  1. A development of containment performance analysis methodology using GOTHIC code

    International Nuclear Information System (INIS)

    Lee, B. C.; Yoon, J. I.; Byun, C. S.; Lee, J. Y.; Lee, J. Y.

    2003-01-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code

  2. FARO base case post-test analysis by COMETA code

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Addabbo, C. [Joint Research Centre, Ispra (Italy)

    1995-09-01

    The paper analyzes the COMETA (Core Melt Thermal-Hydraulic Analysis) post test calculations of FARO Test L-11, the so-called Base Case Test. The FARO Facility, located at JRC Ispra, is used to simulate the consequences of Severe Accidents in Nuclear Power Plants under a variety of conditions. The COMETA Code has a 6 equations two phase flow field and a 3 phases corium field: the jet, the droplets and the fused-debris bed. The analysis shown that the code is able to pick-up all the major phenomena occurring during the fuel-coolant interaction pre-mixing phase.

  3. Severe accident analysis code Sampson for impact project

    International Nuclear Information System (INIS)

    Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

    2001-01-01

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

  4. Transient electromagnetic analysis in tokamaks using TYPHOON code

    International Nuclear Information System (INIS)

    Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.

    1996-01-01

    The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)

  5. Development of a code for the isotopic analysis of Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)

    2013-05-15

    To strengthen the national nuclear nonproliferation regime by an establishment of nuclear forensic system, the techniques for nuclear material analysis and the categorization of important domestic nuclear materials are being developed. MGAU and FRAM are commercial software for the isotopic analysis of Uranium by using γ-spectroscopy, but the diversity of detection geometry and some effects - self attenuation, coincidence summing, etc. - suggest an analysis tool under continual improvement and modification. Hence, developing another code for HPGe γ- and x-ray spectrum analysis is started in this study. The analysis of the 87-101 keV region of Uranium spectrum is attempted based on the isotopic responses similar to those developed in MGAU. The code for isotopic analysis of Uranium is started from a fitting.

  6. QR Codes in the Library: Are They Worth the Effort? Analysis of a QR Code Pilot Project

    OpenAIRE

    Wilson, Andrew M.

    2012-01-01

    The literature is filled with potential uses for Quick Response (QR) codes in the library. Setting, but few library QR code projects have publicized usage statistics. A pilot project carried out in the Eda Kuhn Loeb Music Library of the Harvard College Library sought to determine whether library patrons actually understand and use QR codes. Results and analysis of the pilot project are provided, attempting to answer the question as to whether QR codes are worth the effort for libraries.

  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. Java Source Code Analysis for API Migration to Embedded Systems

    Energy Technology Data Exchange (ETDEWEB)

    Winter, Victor [Univ. of Nebraska, Omaha, NE (United States); McCoy, James A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Guerrero, Jonathan [Univ. of Nebraska, Omaha, NE (United States); Reinke, Carl Werner [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Perry, James Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-02-01

    Embedded systems form an integral part of our technological infrastructure and oftentimes play a complex and critical role within larger systems. From the perspective of reliability, security, and safety, strong arguments can be made favoring the use of Java over C in such systems. In part, this argument is based on the assumption that suitable subsets of Java’s APIs and extension libraries are available to embedded software developers. In practice, a number of Java-based embedded processors do not support the full features of the JVM. For such processors, source code migration is a mechanism by which key abstractions offered by APIs and extension libraries can made available to embedded software developers. The analysis required for Java source code-level library migration is based on the ability to correctly resolve element references to their corresponding element declarations. A key challenge in this setting is how to perform analysis for incomplete source-code bases (e.g., subsets of libraries) from which types and packages have been omitted. This article formalizes an approach that can be used to extend code bases targeted for migration in such a manner that the threats associated the analysis of incomplete code bases are eliminated.

  9. Development of the vacuum system pressure responce analysis code PRAC

    International Nuclear Information System (INIS)

    Horie, Tomoyoshi; Kawasaki, Kouzou; Noshiroya, Shyoji; Koizumi, Jun-ichi.

    1985-03-01

    In this report, we show the method and numerical results of the vacuum system pressure responce analysis code. Since fusion apparatus is made up of many vacuum components, it is required to analyze pressure responce at any points of the system when vacuum system is designed or evaluated. For that purpose evaluating by theoretical solution is insufficient. Numerical analysis procedure such as finite difference method is usefull. In the PRAC code (Pressure Responce Analysis Code), pressure responce is obtained solving derivative equations which is obtained from the equilibrium relation of throughputs and contain the time derivative of pressure. As it considers both molecular and viscous flows, the coefficients of the equation depend on the pressure and the equations become non-linear. This non-linearity is treated as piece-wise linear within each time step. Verification of the code is performed for the simple problems. The agreement between numerical and theoretical solutions is good. To compare with the measured results, complicated model of gas puffing system is analyzed. The agreement is well for practical use. This code will be a useful analytical tool for designing and evaluating vacuum systems such as fusion apparatus. (author)

  10. Residual stress analysis in thick uranium films

    International Nuclear Information System (INIS)

    Hodge, A.M.; Foreman, R.J.; Gallegos, G.F.

    2005-01-01

    Residual stress analysis was performed on thick, 1-25 μm, depleted uranium (DU) films deposited on an Al substrate by magnetron sputtering. Two distinct characterization techniques were used to measure substrate curvature before and after deposition. Stress evaluation was performed using the Benabdi/Roche equation, which is based on beam theory of a bi-layer material. The residual stress evolution was studied as a function of coating thickness and applied negative bias voltage (0, -200, -300 V). The stresses developed were always compressive; however, increasing the coating thickness and applying a bias voltage presented a trend towards more tensile stresses and thus an overall reduction of residual stresses

  11. Thermoelastic stress analysis system developed for industrial applications

    DEFF Research Database (Denmark)

    Haldorsen, Lars Magne

    The thesis is divided into three parts. The first part describes an extensive evaluation of the existing thermoelastic theory. The second part describes the development and results af a reliable numerical simulation code of the thermoelastic effect and the associated heat transfer effects. Finall......, theories, methods and additional equipment are developed in order to adopt a commercial IR-imaging system to preform Termoelastic Stress Analysis (TSA)....

  12. Stress analysis of shear/compression test

    International Nuclear Information System (INIS)

    Nishijima, S.; Okada, T.; Ueno, S.

    1997-01-01

    Stress analysis has been made on the glass fiber reinforced plastics (GFRP) subjected to the combined shear and compression stresses by means of finite element method. The two types of experimental set up were analyzed, that is parallel and series method where the specimen were compressed by tilted jigs which enable to apply the combined stresses, to the specimen. Modified Tsai-Hill criterion was employed to judge the failure under the combined stresses that is the shear strength under the compressive stress. The different failure envelopes were obtained between the two set ups. In the parallel system the shear strength once increased with compressive stress then decreased. On the contrary in the series system the shear strength decreased monotonicly with compressive stress. The difference is caused by the different stress distribution due to the different constraint conditions. The basic parameters which control the failure under the combined stresses will be discussed

  13. Comparison study of inelastic analysis codes for high temperature structure

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-02-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  15. U.F.F.A.: A numerical procedure for fatigue analysis according to ASME code

    International Nuclear Information System (INIS)

    Bellettato, W.; Ticozzi, C.; Zucchini, C.

    1981-01-01

    A new procedure is developed, which employs some already used methodologies and brings some new concepts. The computer code UFFA employs the so obtained procedure. This paper in the first part describes the methodology used for the usage factor calculation, in the second part carries a general description of the code and in the third part shows some example and their respective results. We suppose an elastic behaviour of the materials and we do not consider the effect of the application order of the loads. Moreover, we suppose valid the hypothesis of cumulative damage, that is we apply the Miner's rule. One of the problems in the nuclear components fatigue analysis is that in the load histories there is a high number of operational cycles for which we cannot specify a succession in the time. Therefore, it was introduced the concept of 'level' (or steady working status) by which we can approximate the load conditions in realistic way. As regard the problem of multiaxial cases, it is possible to show that it is not right, an neither conservative, to make a distinguished analysis of the 3 stress differences and then take the maximum of the 3 compuoted usage factors as component usage factor. Indeed, as the stresses act on the structure at the same time, it is necessary a contemporary analysis of the 3 stress difference curves. The computer code can deal as well with the case of sher stresses (varying principal stress directions) through the ASME 'normalization' procedure. The results of the UFFA program, compared with the results of other programs used at present, come up to the expectations. (orig./HP)

  16. Beacon: A three-dimensional structural analysis code for bowing history of fast breeder reactor cores

    International Nuclear Information System (INIS)

    Miki, K.

    1979-01-01

    The core elements of an LMFBR are bowed due to radial gradients of both temperature and neutron flux in the core. Since all hexagonal elements are multiply supported by adjacent elements or the restraint system, restraint forces and bending stresses are induced. In turn, these forces and stresses are relaxed by irradiation enhanced creep of the material. The analysis of the core bowing behavior requires a three-dimensional consideration of the mechanical interactions among the core elements, because the core consists of different kinds of elements and of fuel assemblies with various burnup histories. A new computational code BEACON has been developed for analyzing the bowing behavior of an LMFBR's core in three dimensions. To evaluate mechanical interactions among core elements, the code uses the analytical method of the earlier SHADOW code. BEACON analyzes the mechanical interactions in three directions, which form angles of 60 0 with one another. BEACON is applied to the 60 0 sector of a typical LMFBR's core for analyzing the bowing history during one equilibrium cycle. 120 core elements are treated, assuming the boundary condition of rotational symmetry. The application confirms that the code can be an effective tool for parametric studies as well as for detailed structural analysis of LMFBR's core. (orig.)

  17. Easy web interfaces to IDL code for NSTX Data Analysis

    International Nuclear Information System (INIS)

    Davis, W.M.

    2012-01-01

    Highlights: ► Web interfaces to IDL code can be developed quickly. ► Dozens of Web Tools are used effectively on NSTX for Data Analysis. ► Web interfaces are easier to use than X-window applications. - Abstract: Reusing code is a well-known Software Engineering practice to substantially increase the efficiency of code production, as well as to reduce errors and debugging time. A variety of “Web Tools” for the analysis and display of raw and analyzed physics data are in use on NSTX [1], and new ones can be produced quickly from existing IDL [2] code. A Web Tool with only a few inputs, and which calls an IDL routine written in the proper style, can be created in less than an hour; more typical Web Tools with dozens of inputs, and the need for some adaptation of existing IDL code, can be working in a day or so. Efficiency is also increased for users of Web Tools because of the familiar interface of the web browser, and not needing X-windows, or accounts and passwords, when used within our firewall. Web Tools were adapted for use by PPPL physicists accessing EAST data stored in MDSplus with only a few man-weeks of effort; adapting to additional sites should now be even easier. An overview of Web Tools in use on NSTX, and a list of the most useful features, is also presented.

  18. [Bioethical analysis of the Brazilian Dentistry Code of Ethics].

    Science.gov (United States)

    Pyrrho, Monique; do Prado, Mauro Machado; Cordón, Jorge; Garrafa, Volnei

    2009-01-01

    The Brazilian Dentistry Code of Ethics (DCE), Resolution CFO-71 from May 2006, is an instrument created to guide dentists' behavior in relation to the ethical aspects of professional practice. The purpose of the study is to analyze the above mentioned code comparing the deontological and bioethical focuses. In order to do so, an interpretative analysis of the code and of twelve selected texts was made. Six of the texts were about bioethics and six on deontology, and the analysis was made through the methodological classification of the context units, textual paragraphs and items from the code in the following categories: the referentials of bioethical principlism--autonomy, beneficence, nonmaleficence and justice -, technical aspects and moral virtues related to the profession. Together the four principles represented 22.9%, 39.8% and 54.2% of the content of the DCE, of the deontological texts and of the bioethical texts respectively. In the DCE, 42% of the items referred to virtues, 40.2% were associated to technical aspects and just 22.9% referred to principles. The virtues related to the professionals and the technical aspects together amounted to 70.1% of the code. Instead of focusing on the patient as the subject of the process of oral health care, the DCE focuses on the professional, and it is predominantly turned to legalistic and corporate aspects.

  19. Development of a subchannel analysis code MATRA (Ver. α)

    International Nuclear Information System (INIS)

    Yoo, Y. J.; Hwang, D. H.

    1998-04-01

    A subchannel analysis code MATRA-α, an interim version of MATRA, has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA-α has been provided with an improved structure, various functions, and models to give the more convenient user environment and to increase the code accuracy, various functions, and models to give the more convenient user environment and to increase the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the lateral transport models between adjacent subchannels have been improved to increase the accuracy in predicting two-phase flow phenomena. Also included in this report are the detailed instructions for input data preparation and for auxiliary pre-processors to serve as a guide to those who want to use MATRA-α. In addition, we compared the predictions of MATRA-α with the experimental data on the flow and enthalpy distribution in three sample rod-bundle cases to evaluate the performance of MATRA-α. All the results revealed that the prediction of MATRA-α were better than those of COBRA-IV-I. (author). 16 refs., 1 tab., 13 figs

  20. Factor analysis for exercise stress radionuclide ventriculography

    International Nuclear Information System (INIS)

    Hirota, Kazuyoshi; Yasuda, Mitsutaka; Oku, Hisao; Ikuno, Yoshiyasu; Takeuchi, Kazuhide; Takeda, Tadanao; Ochi, Hironobu

    1987-01-01

    Using factor analysis, a new image processing in exercise stress radionuclide ventriculography, changes in factors associated with exercise were evaluated in 14 patients with angina pectoris or old myocardial infarction. The patients were imaged in the left anterior oblique projection, and three factor images were presented on a color coded scale. Abnormal factors (AF) were observed in 6 patients before exercise, 13 during exercise, and 4 after exercise. In 7 patients, the occurrence of AF was associated with exercise. Five of them became free from AF after exercise. Three patients showing AF before exercise had aggravation of AF during exercise. Overall, the occurrence or aggravation of AF was associated with exercise in ten (71 %) of the patients. The other three patients, however, had disappearance of AF during exercise. In the last patient, none of the AF was observed throughout the study. In view of a high incidence of AF associated with exercise, the factor analysis may have the potential in evaluating cardiac reverse from the viewpoint of left ventricular wall motion abnormality. (Namekawa, K.)

  1. ISODEP, A Fuel Depletion Analysis Code for Predicting Isotopic ...

    African Journals Online (AJOL)

    The trend of results was found to be consistent with those obtained by analytical and other numerical methods. Discovery and Innovation Vol. 13 no. 3/4 December (2001) pp. 184-195. KEY WORDS: depletion analysis, code, research reactor, simultaneous equations, decay of nuclides, radionuclitides, isotope. Résumé

  2. Thermal stratification and fatigue stress analysis for pressurizer surge line

    International Nuclear Information System (INIS)

    Yu Xiaofei; Zhang Yixiong

    2011-01-01

    Thermal stratification of pressurizer surge line induced by the inside fluid results in the global bending moments, local thermal stresses, unexpected displacements and support loadings of the pipe system. In order to avoid a costly three-dimensional computation, a combined 1D/2D technique has been developed and implemented to analyze the thermal stratification and fatigue stress of pressurize surge line of QINSHAN Phase II Extension Nuclear Power Project in this paper, using the computer codes SYSTUS and ROCOCO. According to the mechanical analysis results of stratification, the maximum stress and cumulative usage factor are obtained. The results indicate that the stress and fatigue intensity considering thermal stratification satisfies RCC-M criterion. (authors)

  3. Analysis of pressure wave transients and seismic response in LMFBR piping systems using the SHAPS code

    International Nuclear Information System (INIS)

    Zeuch, W.R.; Wang, C.Y.

    1985-01-01

    This paper presents some of the current capabilities of the three-dimensional piping code SHAPS and demonstrates their usefulness in handling analyses encountered in typical LMFBR studies. Several examples demonstrate the utility of the SHAPS code for problems involving fluid-structure interactions and seismic-related events occurring in three-dimensional piping networks. Results of two studies of pressure wave propagation demonstrate the dynamic coupling of pipes and elbows producing global motion and rigorous treatment of physical quantities such as changes in density, pressure, and strain energy. Results of the seismic analysis demonstrate the capability of SHAPS to handle dynamic structural response within a piping network over an extended transient period of several seconds. Variation in dominant stress frequencies and global translational frequencies were easily handled with the code. 4 refs., 10 figs

  4. Description and user's manual of light water reactor fuel analysis code FEMAXI-IV (Ver.2)

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Saitou, Hiroaki.

    1997-03-01

    FEMAXI-IV is an advanced version of FEMAXI-III, the analysis code of light water reactor fuel behavior in which various functions and improvements have been incorporated. The present report describes in detail the basic theories and structure, the models and numerical solutions applied, and the material properties adopted in the version 2 which is an improved version of the first version of FEMAXI-IV. In FEMAXI-IV (Ver.2), bugs have been fixed, pellet thermal conductivity properties have been updated, and thermal-stress-induced FP gas release model have been incorporated. In order to facilitate effective and wide-ranging application of the code, types and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

  5. Notes on nuclear reactor core analysis code: CITATION

    International Nuclear Information System (INIS)

    Cepraga, D.G.

    1980-01-01

    The method which has evolved over the years for making power reactor calculations is the multigroup diffusion method. The CITATION code is designed to solve multigroup neutronics problems with application of the finite-difference diffusion theory approximation to neutron transport in up to three-dimensional geometry. The first part of this paper presents information about the mathematical equations programmed along with background material and certain displays to convey the nature of some of the formulations. The results obtained with the CITATION code regarding the neutron and burnup core analysis for a typical PWR reactor are presented in the second part of this paper. (author)

  6. FARO and KROTOS code simulation and analysis at JRC Ispra

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Yerkess, A.; Addabbo, C. [European Commission-Joint Research Centre, Inst. for Systems, Informatics and Safety, 21020 Ispra (Italy)

    1998-01-01

    The paper summarizes relevant results from the pre and post test calculations of fuel coolant interaction and quenching tests performed in the FARO and KROTOS test facilities. The main analytical tools adopted at JRC Ispra are the COMETA and the TEXAS codes. COMETA pre and post test calculations of FARO Test L-20 as well as an application of the code to KROTOS test facility are presented. The analysis provides the need to account for H{sub 2} generation models into the pre-mixing calculations. In addition salient results from the application of TEXAS to FARO and KROTOS tests are shown. (author)

  7. 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

  8. 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.

  9. Room Heat-Up Analysis with GOTHIC code

    International Nuclear Information System (INIS)

    Jimenez, G.; Olza, J. M.

    2010-01-01

    The GOTHIC T M computer code is a state-of-the art program for modeling multiphase, multicomponent fluid flow. GOTHIC is rapidly becoming the industry-standard code for performing both containment design basis accident (DBA) analyses and analyses to support equipment qualification. GOTHIC has a flexible nodding structure that allows both lumped parameter and 3-D modeling capabilities. Multidimensional analysis capabilities greatly enhance the study of noncondensable gases and stratification and permit the calculation of flow field details within any given volume.

  10. 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

  11. Organization of Risk Analysis Codes for Living Evaluations (ORACLE)

    International Nuclear Information System (INIS)

    Batt, D.L.; MacDonald, P.E.; Sattison, M.B.; Vesely, E.

    1987-01-01

    ORACLE (Organization of Risk Analysis Codes for Living Evaluations) is an integration concept for using risk-based information in United States Nuclear Regulatory Commission (USNRC) applications. Portions of ORACLE are being developed at the Idaho Nationale Engineering Laboratory for the USNRC. The ORACLE concept consists of related databases, software, user interfaces, processes, and quality control checks allowing a wide variety of regulatory problems and activities to be addressed using current, updated PRA information. The ORACLE concept provides for smooth transitions between one code and the next without pre- or post-processing. (orig.)

  12. Probabilistic analysis of crack containing structures with the PARIS code

    International Nuclear Information System (INIS)

    Brueckner-Foit, A.

    1987-10-01

    The basic features of the PARIS code which has been developed for the calculation of failure probabilities of crack containing structures are explained. An important issue in the reliability analysis of cracked components is the probabilistic leak-before-break behaviour. Formulae for the leak and break probabilities are derived and it is shown how a leak detection system influences the results. An example taken from nuclear applications illustrates the details of the probabilistic leak-before-break analysis. (orig.) [de

  13. Development and assessment of best estimate integrated safety analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu (and others)

    2007-03-15

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published.

  14. Development and assessment of best estimate integrated safety analysis code

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu

    2007-03-01

    Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published

  15. Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Kwi Hyun; Song, Wan Jung [ENERGEO Inc. Sungnam, (Korea, Republic of); Song, Dong Soo; Byun, Choong Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences.

  16. Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

    International Nuclear Information System (INIS)

    Seo, Kwi Hyun; Song, Wan Jung; Song, Dong Soo; Byun, Choong Sup

    2006-01-01

    The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences

  17. Analysis of ATLAS Cold Leg SBLOCA Using SPACE Code

    International Nuclear Information System (INIS)

    Kang, Doo Hyuk; Suh, Jae Seung; Kim, Se Yun

    2012-01-01

    SPACE Code has been developed to predict the thermal-hydraulic responses of nuclear steam supply system to the anticipated transients and postulated accidents and adopted advanced physical modeling of two-phase flows, mainly two-fluid, three-field models that comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or non-structured meshes. In this paper, a cold-leg SBLOCA which is the experiment, SB-CL-09, of the ATLAS integral effect test facility during the second domestic stand problem (DSP-02) was analyzed. The results were compared with those of MARS-KS code simulations. The SPACE code with a 1.0 version was released by KHNP in 2012. The analysis has been performed in a desktop PC with Windows 7 environment

  18. FAST: An advanced code system for fast reactor transient analysis

    International Nuclear Information System (INIS)

    Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

    2005-01-01

    One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

  19. Stress analysis in FRP composites

    Directory of Open Access Journals (Sweden)

    Nitin Jauhari

    2016-09-01

    Full Text Available A composite material, in mechanics sense, is a structure with the ingredients as element transferring forces to adjacent members. The issue of defects and their effect on the mechanical properties of composites is of great concern among high end users. Experimental investigation of failure modes of composite materials requires correlating the fundamentals of composite materials, their mechanical properties as well as their failure characteristics in the presence of defects. In this paper, three formats of defects of hole (single, double and quadruple as a discontinuity were incorporated along with tensile testing. Unique failure modes of these specimens provided overview regarding mechanical behaviour of composite materials containing defects. Certain correlations were observed between defects and resulting properties. Results are in agreement with general behaviour of FRP composite laminates and it can be concluded that for low deformation in composite laminates, number of layers must be increased, which at the same time results in increase of von-Mises stress. Fibres are the main constituents which are responsible for strength of a composite laminate and they along with fibre orientation, play an important role on its load bearing capacity. It can be inferred based on the analysis that cross-ply configuration [0°/90°] has good load bearing capacity as well as least deflection emphasizing more strength.

  20. Analysis of the anisotropy effects with the AZTRAN code

    International Nuclear Information System (INIS)

    Xolocostli, V.; Vargas, S.; Gomez, A.; Del Valle, E.

    2017-09-01

    Among the improvements that are made for the deterministic codes with which nuclear reactors are analyzed, is the implementation of the dispersion anisotropic dispersion section, which can obtain better results. With the current computing technology is possible to carry out these implementations, since the computation time is no longer a considerable problem as in the past. In this paper we analyze some effects of anisotropy in the AZTRAN code, a code that solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multigroup technique, the nodal method RTN-0 and ordered discrete, which is part of the AZTLAN platform for analysis of nuclear reactors, which is currently under development. The implementation of the anisotropy in the AZTRAN code is one of the latest improvements that have been made to the code, leading to different tests and analyzes regarding the anisotropic dispersion, some as a test with homogeneous fuel assemblies. In the case presented here, the benchmark problem of a fuel assembly type BWR is analyzed, which is part of the Benchmark problem suite for reactor physics study of LWR next generation fuels, proposed by the Committee on Reactor Physics organized by the Japan Atomic Energy Research Institute (JAERI). In this problem the behavior of the infinite multiplication factor (k inf ) is analyzed, as well as the behavior of using odd and even anisotropy approximation with respect to the symmetry in the radial power of the assembly. (Author)

  1. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

    Togawa, O.; Homma, T.

    1992-01-01

    As part of a computer code system for nuclear reactor accident consequence analysis, two computer codes have been developed for estimating health effects expected to occur following an accident. Health effects models used in the codes are based on the models of NUREG/CR-4214 and are revised for the Japanese population on the basis of the data from the reassessment of the radiation dosimetry and information derived from epidemiological studies on atomic bomb survivors of Hiroshima and Nagasaki. The health effects models include early and continuing effects, late somatic effects and genetic effects. The values of some model parameters are revised for early mortality. The models are modified for predicting late somatic effects such as leukemia and various kinds of cancers. The models for genetic effects are the same as those of NUREG. In order to test the performance of one of these codes, it is applied to the U.S. and Japanese populations. This paper provides descriptions of health effects models used in the two codes and gives comparisons of the mortality risks from each type of cancer for the two populations. (author)

  2. Seismic Analysis Code (SAC): Development, porting, and maintenance within a legacy code base

    Science.gov (United States)

    Savage, B.; Snoke, J. A.

    2017-12-01

    The Seismic Analysis Code (SAC) is the result of toil of many developers over almost a 40-year history. Initially a Fortran-based code, it has undergone major transitions in underlying bit size from 16 to 32, in the 1980s, and 32 to 64 in 2009; as well as a change in language from Fortran to C in the late 1990s. Maintenance of SAC, the program and its associated libraries, have tracked changes in hardware and operating systems including the advent of Linux in the early 1990, the emergence and demise of Sun/Solaris, variants of OSX processors (PowerPC and x86), and Windows (Cygwin). Traces of these systems are still visible in source code and associated comments. A major concern while improving and maintaining a routinely used, legacy code is a fear of introducing bugs or inadvertently removing favorite features of long-time users. Prior to 2004, SAC was maintained and distributed by LLNL (Lawrence Livermore National Lab). In that year, the license was transferred from LLNL to IRIS (Incorporated Research Institutions for Seismology), but the license is not open source. However, there have been thousands of downloads a year of the package, either source code or binaries for specific system. Starting in 2004, the co-authors have maintained the SAC package for IRIS. In our updates, we fixed bugs, incorporated newly introduced seismic analysis procedures (such as EVALRESP), added new, accessible features (plotting and parsing), and improved the documentation (now in HTML and PDF formats). Moreover, we have added modern software engineering practices to the development of SAC including use of recent source control systems, high-level tests, and scripted, virtualized environments for rapid testing and building. Finally, a "sac-help" listserv (administered by IRIS) was setup for SAC-related issues and is the primary avenue for users seeking advice and reporting bugs. Attempts are always made to respond to issues and bugs in a timely fashion. For the past thirty-plus years

  3. Stress analysis of HTR-10 steam generator heat exchanging tubes

    International Nuclear Information System (INIS)

    Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

    2001-01-01

    Steam Generator (SG) heat exchanging tubes of 10 MW High Temperature Gas Cooled Reactor (HTR-10) are protective screens between the primary loop of helium with radioactivity and the secondary loop of feeding water and steam without radioactivity. Water and steam will enter into the primary loop when rupture of the heat exchanging tubes occurs, which lead to increase of the primary loop pressure and discharge of radioactive materials. Therefore it is important to guarantee the integrity of the tubes. The tube structure is spiral tube with small bending radius, which make it impossible to test with volumetric in-service detection. For such kind of spiral tube, using LBB concept to guarantee the integrity of the tubes is an important option. The author conducts stress analysis and calculation of HTR-10 SG heat exchanging tubes using the FEM code of piping stress analysis, PIPESTRESS. The maximum stress and the dangerous positions are obtained

  4. SATURN-FS 1: A computer code for thermo-mechanical fuel rod analysis

    International Nuclear Information System (INIS)

    Ritzhaupt-Kleissl, H.J.; Heck, M.

    1993-09-01

    The SATURN-FS code was written as a general revision of the SATURN-2 code. SATURN-FS is capable to perform a complete thermomechanical analysis of a fuel pin, with all thermal, mechanical and irradiation-based effects. Analysis is possible for LWR and for LMFBR fuel pins. The thermal analysis consists of calculations of the temperature profile in fuel, gap and in the cladding. Pore migration, stoichiometry change of oxide fuel, gas release and diffusion effects are taken into account. The mechanical modeling allows the non steady-state analysis of elastic and nonelastic fuel pin behaviour, such as creep, strain hardening, recovery and stress relaxation. Fuel cracking and healing is taken into account as well as contact and friction between fuel and cladding. The modeling of the irradiation effects comprises swelling and fission gas production, Pu-migration and irradiation induced creep. The code structure, the models and the requirements for running the code are described in the report. Recommendations for the application are given. Program runs for verification and typical examples of application are given in the last part of this report. (orig.) [de

  5. Performance Analysis of CRC Codes for Systematic and Nonsystematic Polar Codes with List Decoding

    Directory of Open Access Journals (Sweden)

    Takumi Murata

    2018-01-01

    Full Text Available Successive cancellation list (SCL decoding of polar codes is an effective approach that can significantly outperform the original successive cancellation (SC decoding, provided that proper cyclic redundancy-check (CRC codes are employed at the stage of candidate selection. Previous studies on CRC-assisted polar codes mostly focus on improvement of the decoding algorithms as well as their implementation, and little attention has been paid to the CRC code structure itself. For the CRC-concatenated polar codes with CRC code as their outer code, the use of longer CRC code leads to reduction of information rate, whereas the use of shorter CRC code may reduce the error detection probability, thus degrading the frame error rate (FER performance. Therefore, CRC codes of proper length should be employed in order to optimize the FER performance for a given signal-to-noise ratio (SNR per information bit. In this paper, we investigate the effect of CRC codes on the FER performance of polar codes with list decoding in terms of the CRC code length as well as its generator polynomials. Both the original nonsystematic and systematic polar codes are considered, and we also demonstrate that different behaviors of CRC codes should be observed depending on whether the inner polar code is systematic or not.

  6. Qualification of ARROTTA code for LWR accident analysis

    International Nuclear Information System (INIS)

    Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.

    2004-01-01

    This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)

  7. Analysis of the Behavior of CAREM-25 Fuel Rods Using Computer Code BACO

    International Nuclear Information System (INIS)

    Estevez, Esteban; Markiewicz, Mario; Marino, Armando

    2000-01-01

    The thermo-mechanical behavior of a fuel rod subjected to irradiation is a complex process, on which a great quantity of interrelated physical-chemical phenomena are coupled.The code BACO simulates the thermo-mechanical behavior and the evolution of fission gases of a cylindrical rod in operation.The power history of fuel rods, arising from neutronic calculations, is the program input.The code calculates, among others, the temperature distribution and the principal stresses in the pellet and cladding, changes in the porosity and restructuring of pellet, the fission gases release, evolution of the internal gas pressure.In this work some of design limits of CAREM-25's fuel rods are analyzed by means of the computer code BACO.The main variables directly related with the integrity of the fuel rod are: Maximum temperature of pellet; Cladding hoop stresses; Gases pressure in the fuel rod; Cladding axial and radial strains, etc.The analysis of results indicates that, under normal operation conditions, the maximum fuel pellet temperature, cladding stresses, pressure of gases at end of life, etc, are below the design limits considered for the fuel rod of CAREM-25 reactor

  8. Code portability and data management considerations in the SAS3D LMFBR accident-analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1981-01-01

    The SAS3D code was produced from a predecessor in order to reduce or eliminate interrelated problems in the areas of code portability, the large size of the code, inflexibility in the use of memory and the size of cases that can be run, code maintenance, and running speed. Many conventional solutions, such as variable dimensioning, disk storage, virtual memory, and existing code-maintenance utilities were not feasible or did not help in this case. A new data management scheme was developed, coding standards and procedures were adopted, special machine-dependent routines were written, and a portable source code processing code was written. The resulting code is quite portable, quite flexible in the use of memory and the size of cases that can be run, much easier to maintain, and faster running. SAS3D is still a large, long running code that only runs well if sufficient main memory is available

  9. Development of the criticality accident analysis code, AGNES

    International Nuclear Information System (INIS)

    Nakajima, Ken

    1989-01-01

    In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

  10. Stress analysis of closure bolts for shipping casks

    International Nuclear Information System (INIS)

    Mok, G.C.; Fischer, L.E.; Hsu, S.T.

    1993-01-01

    This report specifies the requirements and criteria for stress analysis of closure bolts for shipping casks containing nuclear spent fuels or high level radioactive materials. The specification is based on existing information conceming the structural behavior, analysis, and design of bolted joints. The approach taken was to extend the ASME Boiler and Pressure Vessel Code requirements and criteria for bolting analysis of nuclear piping and pressure vessels to include the appropriate design and load characteristics of the shipping cask. The characteristics considered are large, flat, closure lids with metal-to-metal contact within the bolted joint; significant temperature and impact loads; and possible prying and bending effects. Specific formulas and procedures developed apply to the bolt stress analysis of a circular, flat, bolted closure. The report also includes critical load cases and desirable design practices for the bolted closure, an in-depth review of the structural behavior of bolted joints, and a comprehensive bibliography of current information on bolted joints

  11. Dynamic analysis of aircraft impact using the linear elastic finite element codes FINEL, SAP and STARDYNE

    International Nuclear Information System (INIS)

    Lundsager, P.; Krenk, S.

    1975-08-01

    The static and dynamic response of a cylindrical/ spherical containment to a Boeing 720 impact is computed using 3 different linear elastic computer codes: FINEL, SAP and STARDYNE. Stress and displacement fields are shown together with time histories for a point in the impact zone. The main conclusions from this study are: - In this case the maximum dynamic load factors for stress and displacements were close to 1, but a static analysis alone is not fully sufficient. - More realistic load time histories should be considered. - The main effects seem to be local. The present study does not indicate general collapse from elastic stresses alone. - Further study of material properties at high rates is needed. (author)

  12. On the concept of elasticity used in some fast reactor accident analysis codes

    International Nuclear Information System (INIS)

    Malmberg, T.

    1975-01-01

    The analysis to be presented will restrict attention to the elastic part of the elastic-plastic constitutive equation used in several Fast Reactor Accident Analysis Codes and originally applied by M.L. Wilkins: Calculation of Elastic-Plastic Flow, UCRL-7322, Rev. 1, Jan. 1969. It is shown that the used elasticity concept is within the frame of hypo-elasticity. On the basis of a test found by Bernstein it is proven that the state of stress is generally depending on the path of deformation. Therefore this concept of elasticity is not compatible with finite elasticity. For several simple deformation processes this special hypo-elastic constitutive equation is integrated to give a stress-strain relation. The path-dependence of this relation is demonstrated. Further the phenomenon of hypo-elastic yield under shear deformation is pointed out. The relevance to modelling material behaviour in primary containment analysis is discussed

  13. On the concept of elasticity used in some fast reactor accident analysis codes

    International Nuclear Information System (INIS)

    Malmberg, T.

    1975-01-01

    The analysis presented restricts attention to the elastic part of the elastic-plastic equation used in several Fast Reactor Accident Analysis Codes and originally applied by M.L. Wilkins: Calculation of Elastic-Plastic Flow, UCRL-7322, Rev. 1, Jan 1969. It is shown that the used elasticity concept is within the frame of hypo-elasticity. On the basis of a test found by Bernstein it is proven that the state of stress is generally depending on the path of deformation. Therefore this concept of elasticity is not compatible with finite elasticity. For several deformation processes this special hypo-elastic constitutive equation is integrated to give a stress-strain relation. The path-dependence of this relation is demonstrated. Further the phenomenon of hypo-elastic yield under shear deformation is pointed out. The relevance to modelling material behaviour in primary containment analysis is discussed. (Auth.)

  14. THYDE-P2 code: RCS (reactor-coolant system) analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro; Hirano, Masashi; Sato, Kazuo

    1986-12-01

    THYDE-P2, being characterized by the new thermal-hydraulic network model, is applicable to analysis of RCS behaviors in response to various disturbances including LB (large break)-LOCA(loss-of-coolant accident). In LB-LOCA analysis, THYDE-P2 is capable of through calculation from its initiation to complete reflooding of the core without an artificial change in the methods and models. The first half of the report is the description of the methods and models for use in the THYDE-P2 code, i.e., (1) the thermal-hydraulic network model, (2) the various RCS components models, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the user's mannual for the THYDE-P2 code (version SV04L08A) containing items; (1) the program control (2) the input requirements, (3) the execution of THYDE-P2 job, (4) the output specifications and (5) the sample problem to demonstrate capability of the thermal-hydraulic network model, among other things. (author)

  15. Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corporation, Tokyo (Japan)

    2013-10-15

    A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

  16. Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

    2013-10-01

    A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

  17. 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

  18. A Novel Non-coding RNA Regulates Drought Stress Tolerance in Arabidopsis thaliana

    KAUST Repository

    Albesher, Nour H.

    2014-05-01

    Drought (soil water deficit) as a major adverse environmental condition can result in serious reduction in plant growth and crop production. Plants respond and adapt to drought stresses by triggering various signalling pathways leading to physiological, metabolic and developmental changes that may ultimately contribute to enhanced tolerance to the stress. Here, a novel non-coding RNA (ncRNA) involved in plant drought stress tolerance was identified. We showed that increasing the expression of this ncRNA led to enhanced sensitivity during seed germination and seedling growth to the phytohormone abscisic acid. The mutant seedlings are also more sensitive to osmotic stress inhibition of lateral root growth. Consistently, seedlings with enhanced expression of this ncRNA exhibited reduced transiprational water loss and were more drought-tolerant than the wild type. Future analyses of the mechanism for its role in drought tolerance may help us to understand how plant drought tolerance could be further regulated by this novel ncRNA.

  19. A Stress-Induced Bias in the Reading of the Genetic Code in Escherichia coli

    Directory of Open Access Journals (Sweden)

    Adi Oron-Gottesman

    2016-11-01

    Full Text Available Escherichia coli mazEF is an extensively studied stress-induced toxin-antitoxin (TA system. The toxin MazF is an endoribonuclease that cleaves RNAs at ACA sites. Thereby, under stress, the induced MazF generates a stress-induced translation machinery (STM, composed of MazF-processed mRNAs and selective ribosomes that specifically translate the processed mRNAs. Here, we further characterized the STM system, finding that MazF cleaves only ACA sites located in the open reading frames of processed mRNAs, while out-of-frame ACAs are resistant. This in-frame ACA cleavage of MazF seems to depend on MazF binding to an extracellular-death-factor (EDF-like element in ribosomal protein bS1 (bacterial S1, apparently causing MazF to be part of STM ribosomes. Furthermore, due to the in-frame MazF cleavage of ACAs under stress, a bias occurs in the reading of the genetic code causing the amino acid threonine to be encoded only by its synonym codon ACC, ACU, or ACG, instead of by ACA.

  20. Moderator 3-D Thermalhydraulic Analysis Using MODTURCCLAS Code

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Yoon, Churl; Park, Joo Hwan

    2008-12-01

    For the moderator subcooling analysis of the refurbished Wolsong NPP (Nuclear Power Plant) Unit 1, the 3-D moderator thermalhydraulic calculations were preliminarily conducted from September to October in 2008, using the preliminary thermalhydraulic analysis results and the MODTURC C LAS (MODerator TURbulent Circulation Co-Located Advanced Solutions) Ver.2.9-IST, which was developed and validated by OPG (Ontario Power Generation) in Canada. The present report consists of the steady-state calculation and transient calculation. First the grid structure, main input parameters, and boundary conditions needed for the steady-state calculation are produced and the steady-state results are obtained. These steady-state results are used for the initial conditions for the transient analysis during a LOCA. The moderator analysis results during the steady-state calculation show the quasy-steady state behavior, because the thermalhydraulic behavior are fluctuating although all boundary conditions are constant. In the transient calculations, based on the present thermalhydraulic analysis results, 3-D thermalhydraulic behavior and moderator subcooling are predicted for the accident scenarios of reactor inlet header 35% and 40% breaks, outlet header 100% break, and pump suction 80% break, subsequent with loss of Class IV power. In the previous moderator analysis for Wolsong NPP Unit 2,3,4 a PHOENICS code was used, which is different from the MODTURC C LAS code used for the analysis of Wolsong NPP Unit 1. However, the moderator subcooling analysis results by these two codes are qualitatively similar. The minimum subcooling for RIH 40% break of Wolsong NPP Unit 1 is 17 .deg. C which is larger than 13 .deg. C for RIH 35% break of Wolsong NPP Unit 2,3,4. Therefore, it is concluded that the refurbished Wolsong NPP Unit 1 satisfies the channel integrity criteria based on the higher subcooling margin compared with that of Wolsong NPP Unit 2,3,4

  1. Development of thermal hydraulic analysis code for IHX of FBR

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Naohara, Nobuyuki

    1991-01-01

    In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)

  2. Research on the improvement of nuclear safety -The development of a severe accident analysis code-

    International Nuclear Information System (INIS)

    Kim, Heui Dong; Cho, Sung Won; Park, Jong Hwa; Hong, Sung Wan; Yoo, Dong Han; Hwang, Moon Kyoo; Noh, Kee Man; Song, Yong Man

    1995-07-01

    For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity and is intended to improve existing models and develop analytical tools for the assessment of severe accidents. A correlation equation of the flame velocity of pre mixture gas of H 2 /air/steam has been suggested and combustion flame characteristic was analyzed using a developed computer code. For the analysis of the expansion phase of vapor explosion, the mechanical model has been developed. The development of a debris entrainment model in a reactor cavity with captured volume has been continued to review and examine the limitation and deficiencies of the existing models. Pre-test calculation was performed to support the severe accident experiment for molten corium concrete interaction study and the crust formation process and heat transfer characteristics of the crust have been carried out. A stress analysis code was developed using finite element method for the reactor vessel lower head failure analysis. Through international program of PHEBUS-FP and participation in the software development, the research on the core degradation process and fission products release and transportation are undergoing. CONTAIN and MELCOR codes were continuously updated under the cooperation with USNRC and French developed computer codes such as ICARE2, ESCADRE, SOPHAEROS were also installed into the SUN workstation. 204 figs, 61 tabs, 87 refs. (Author)

  3. Source Code Analysis Laboratory (SCALe) for Energy Delivery Systems

    Science.gov (United States)

    2010-12-01

    technical competence for the type of tests and calibrations SCALe undertakes. Testing and calibration laboratories that comply with ISO / IEC 17025 ...and exec t [ ISO / IEC 2005]. f a software system indicates that the SCALe analysis di by a CERT secure coding standard. Successful conforma antees that...to be more secure than non- systems. However, no study has yet been performed to p t ssment in accordance with ISO / IEC 17000: “a demonstr g to a

  4. Validation of OPERA3D PCMI Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeun, Ji Hoon; Choi, Jae Myung; Yoo, Jong Sung [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of); Cheng, G.; Sim, K. S.; Chassie, Girma [Candu Energy INC.,Ontario (Canada)

    2013-10-15

    This report will describe introduction of validation of OPERA3D code, and validation results that are directly related with PCMI phenomena. OPERA3D was developed for the PCMI analysis and validated using the in-pile measurement data. Fuel centerline temperature and clad strain calculation results shows close expectations with measurement data. Moreover, 3D FEM fuel model of OPERA3D shows slight hour glassing behavior of fuel pellet in contact case. Further optimization will be conducted for future application of OPERA3D code. Nuclear power plant consists of many complicated systems, and one of the important objects of all the systems is maintaining nuclear fuel integrity. However, it is inevitable to experience PCMI (Pellet Cladding Mechanical Interaction) phenomena at current operating reactors and next generation reactors for advanced safety and economics as well. To evaluate PCMI behavior, many studies are on-going to develop 3-dimensional fuel performance evaluation codes. Moreover, these codes are essential to set the safety limits for the best estimated PCMI phenomena aimed for high burnup fuel.

  5. LACEwING: A New Moving Group Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Riedel, Adric R. [Department of Astronomy, California Institute of Technology, Pasadena, CA 91125 (United States); Blunt, Sarah C.; Faherty, Jacqueline K. [Department of Astrophysics, American Museum of Natural History, New York, NY 10024 (United States); Lambrides, Erini L. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Rice, Emily L. [Department of Engineering Science and Physics, The College of Staten Island, Staten Island, NY 10314 (United States); Cruz, Kelle L., E-mail: arr@astro.caltech.edu [Department of Physics and Astronomy, Hunter College, New York, NY 10065 (United States)

    2017-03-01

    We present a new nearby young moving group (NYMG) kinematic membership analysis code, LocAting Constituent mEmbers In Nearby Groups (LACEwING), a new Catalog of Suspected Nearby Young Stars, a new list of bona fide members of moving groups, and a kinematic traceback code. LACEwING is a convergence-style algorithm with carefully vetted membership statistics based on a large numerical simulation of the Solar Neighborhood. Given spatial and kinematic information on stars, LACEwING calculates membership probabilities in 13 NYMGs and three open clusters within 100 pc. In addition to describing the inputs, methods, and products of the code, we provide comparisons of LACEwING to other popular kinematic moving group membership identification codes. As a proof of concept, we use LACEwING to reconsider the membership of 930 stellar systems in the Solar Neighborhood (within 100 pc) that have reported measurable lithium equivalent widths. We quantify the evidence in support of a population of young stars not attached to any NYMGs, which is a possible sign of new as-yet-undiscovered groups or of a field population of young stars.

  6. Stress analysis program system for nuclear vessel: STANSAS

    International Nuclear Information System (INIS)

    Okamoto, Asao; Michikami, Shinsuke

    1979-01-01

    IHI has developed a computer system of stress analysis and evaluation for nuclear vessels: STANSAS (STress ANalysis System for Axi-symmetric Structure). The system consists of more than twenty independent programs divided into the following six parts. 1. Programs for opening design by code rule. 2. Calculation model generating programs. 3. Load defining programs. 4. Structural analysis programs. 5. Load data/calculation results plotting programs. 6. Stress evaluation programs. Each program is connected with its pre- or post-processor through three data-bases which enable automatic data transfer. The user can make his choice of structural analysis programs in accordance with the problem to be solved. The interface to STANSAS can be easily installed in generalized structural analysis programs such as NASTRAN and MARC. For almost all tables and figures in the stress report, STANSAS has the function to print or plot out. The complicated procedures of ''Design by Analysis'' for pressure vessels have been well standardized by STANSAS. The system will give a high degree of efficiency and confidence to the design work. (author)

  7. SHEAN (Simplified Human Error Analysis code) and automated THERP

    International Nuclear Information System (INIS)

    Wilson, J.R.

    1993-01-01

    One of the most widely used human error analysis tools is THERP (Technique for Human Error Rate Prediction). Unfortunately, this tool has disadvantages. The Nuclear Regulatory Commission, realizing these drawbacks, commissioned Dr. Swain, the author of THERP, to create a simpler, more consistent tool for deriving human error rates. That effort produced the Accident Sequence Evaluation Program Human Reliability Analysis Procedure (ASEP), which is more conservative than THERP, but a valuable screening tool. ASEP involves answering simple questions about the scenario in question, and then looking up the appropriate human error rate in the indicated table (THERP also uses look-up tables, but four times as many). The advantages of ASEP are that human factors expertise is not required, and the training to use the method is minimal. Although not originally envisioned by Dr. Swain, the ASEP approach actually begs to be computerized. That WINCO did, calling the code SHEAN, for Simplified Human Error ANalysis. The code was done in TURBO Basic for IBM or IBM-compatible MS-DOS, for fast execution. WINCO is now in the process of comparing this code against THERP for various scenarios. This report provides a discussion of SHEAN

  8. Development of the SCHAMBETA code for scoping analysis of HCDA

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Hahn, D. H

    2000-06-01

    A computer code, SCHAMBETA(Scoping Code for HCDA Analysis using Modified Bethe-Tait Method), is developed to investigate the core disassembly process following a meltdown accident in the framework of a mofified Bethe-Tait method as part of the scoping analysis work to demonstrate the inherent safety of conceptual designs of Korea Advanced Liquid Metal Reactor(KALIMER), A 150 Mwe pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. The methodologies adopted in the code ared particularly useful to perform various parametric studies for better understanding of core disassembly process of liquid metal fast reactors as well as to estimate upper-limit values of the energy release resulting from a power excursion. In the SCHAMBETA code, the core kinetics and hydraulic behavior of the KALIMER is followed over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion, starting at the time that the sodium-voided core reaches the melting temperature of the metallic fuels. For this purpose, the equations of state of pressure-energy density relationship are derived for the saturated-vapor as well as the solid liquid of metallic uranium fuel, and implemenmted into the formulations of the disassembly reactivity. Mathematical formulations are then developed, in the framework of Modified Bethe-Tait method, in a form relevant to utilize the improved equations of state as well as to consider Doppler effects, for scoping analysis of the super-prompt-critical power excursions driven by a specified rate of reactivity insertion.

  9. Site-specific Probabilistic Analysis of DCGLs Using RESRAD Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeongju; Yoon, Suk Bon; Sohn, Wook [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    In general, DCGLs can be conservative (screening DCGL) if they do not take into account site specific factors. Use of such conservative DCGLs can lead to additional remediation that would not be required if the effort was made to develop site-specific DCGLs. Therefore, the objective of this work is to provide an example on the use of the RESRAD 6.0 probabilistic (site-specific) dose analysis to compare with the screening DCGL. Site release regulations state that a site will be considered acceptable for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent (TEDE) to an average member of the critical group of less than the site release criteria, for example 0.25 mSv per year in U.S. Utilities use computer dose modeling codes to establish an acceptable level of contamination, the derived concentration guideline level (DCGL) that will meet this regulatory limit. Since the DCGL value is the principal measure of residual radioactivity, it is critical to understand the technical basis of these dose modeling codes. The objective this work was to provide example on nuclear power plant decommissioning dose analysis in a probabilistic analysis framework. The focus was on the demonstration of regulatory compliance for surface soil contamination using the RESRAD 6.0 code. Both the screening and site-specific probabilistic dose analysis methodologies were examined. Example analyses performed with the screening probabilistic dose analysis confirmed the conservatism of the NRC screening values and indicated the effectiveness of probabilistic dose analysis in reducing the conservatism in DCGL derivation.

  10. Light water reactor fuel analysis code FEMAXI-IV(Ver.2). Detailed structure and user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    A light water reactor fuel behavior analysis code FEMAXI-IV(Ver.2) was developed as an improved version of FEMAXI-IV. Development of FEMAXI-IV has been already finished in 1992, though a detailed structure and input manual of the code have not been open to users yet. Here, the basic theories and structure, the models and numerical solutions applied to FEMAXI-IV(Ver.2), and the material properties adopted in the code are described in detail. In FEMAXI-IV(Ver.2), programming bugs in previous FEMAXI-IV were eliminated, renewal of the pellet thermal conductivity was performed, and a model of thermal-stress restraint on FP gas release was incorporated. For facilitation of effective and wide-ranging application of the code, methods of input/output of the code are also described in detail, and sample output is included. (author)

  11. THYDE-NEU: Nuclear reactor system analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2002-03-01

    THYDE-NEU is applicable not only to transient analyses, but also to steady state analyses of nuclear reactor systems (NRSs). In a steady state analysis, the code generates a solution satisfying the transient equations without external disturbances. In a transient analysis, the code calculates temporal NRS behaviors in response to various external disturbances in such a way that mass and energy of the coolant as well as the number of neutrons conserve. The first half of the report is the description of the methods and models for use in the THYDE-NEU code, i.e., (1) the thermal-hydraulic network model, (2) the spatial kinetics model, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the users' mannual containing the items; (1) the program control, (2) the input requirements, (3) the execution of THYDE-NEU jobs, (4) the output specifications and (5) the sample calculation. (author)

  12. Development of a neutronic analysis code using data from Monju

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van; Yamano, N.; Shimazu, Y.

    2015-01-01

    In recent years three major sets of modern evaluated nuclear data have become available: JENDL-4.0, JEFF-3.1.2 and ENDF/B-VII.1. The authors were involved with a research project to establish analysis method for a future commercial-scale LMFBR. This project focused on JENDL-4.0 and conventional Japanese codes. As a cross check, we decided to also apply the fast reactor code ERANOS. This necessitated to produce nuclear data (cross sections, etc) for the ERANOS code system, discussed in this paper. We developed a nuclear data processing system to produce cross sections, probability tables, delayed neutron data, and covariance data from the evaluated nuclear data files for ERANOS. A benchmark calculation on the MZA/MZB benchmark showed very satisfying results. Subsequently, we analyzed the prototype LMFBR Monju with ERANOS and our own sets of nuclear data. The results are very satisfactory. The results from ERANOS indicate that the target accuracies for nuclear data have not been met, although the three sets of evaluated nuclear data all performed very well in our analysis. In the future, the covariance on nuclear data should be reduced to meet the target accuracies on criticality and feedback coefficients. (author)

  13. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

  14. TRAWA, a transient analysis code for water reactions

    International Nuclear Information System (INIS)

    Rajamaeki, M.

    1976-06-01

    TRAWA is a transient analysis code for water reactors. It solves the two-group neutron diffusion equations simultaneously with the heat conduction equations and the two-phase hydraulic equations for one or more channels. At most one-dimensional submodels are used. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermohydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channel and risers with two-phase flow and of pump lines with incompressible flow. Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. Various transients can be calculated by applying external disturbances. The code is extensively supplied by input and output capabilities. TRAWA is written in FORTRAN V for UNIVAC 1108 computer. (author)

  15. Beam Optics Analysis - An Advanced 3D Trajectory Code

    International Nuclear Information System (INIS)

    Ives, R. Lawrence; Bui, Thuc; Vogler, William; Neilson, Jeff; Read, Mike; Shephard, Mark; Bauer, Andrew; Datta, Dibyendu; Beal, Mark

    2006-01-01

    Calabazas Creek Research, Inc. has completed initial development of an advanced, 3D program for modeling electron trajectories in electromagnetic fields. The code is being used to design complex guns and collectors. Beam Optics Analysis (BOA) is a fully relativistic, charged particle code using adaptive, finite element meshing. Geometrical input is imported from CAD programs generating ACIS-formatted files. Parametric data is inputted using an intuitive, graphical user interface (GUI), which also provides control of convergence, accuracy, and post processing. The program includes a magnetic field solver, and magnetic information can be imported from Maxwell 2D/3D and other programs. The program supports thermionic emission and injected beams. Secondary electron emission is also supported, including multiple generations. Work on field emission is in progress as well as implementation of computer optimization of both the geometry and operating parameters. The principle features of the program and its capabilities are presented

  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. Development of Fuel ROd Behavior Analysis code (FROBA) and its application to AP1000

    International Nuclear Information System (INIS)

    Yu, Hongxing; Tian, Wenxi; Yang, Zhen; SU, G.H.; Qiu, Suizheng

    2012-01-01

    Highlights: ► A Fuel ROd Behavior Analysis code (FROBA) has been developed. ► The effects irradiation and burnup has been considered in FROBA. ► The comparison with INL’s results shows a good agreement. ► The FROBA code was applied to AP1000. ► Peak fuel temperature, gap width, hoop strain, etc. were obtained. -- Abstract: The reliable prediction of nuclear fuel rod behavior is of great importance for safety evaluation of nuclear reactors. In the present study, a thermo-mechanical coupling code FROBA (Fuel ROd Behavior Analysis) has been independently developed with consideration of irradiation and burnup effects. The thermodynamic, geometrical and mechanical behaviors have been predicted and were compared with the results obtained by Idaho National Laboratory to validate the reliability and accuracy of the FROBA code. The validated code was applied to analyze the fuel behavior of AP1000 at different burnup levels. The thermal results show that the predicted peak fuel temperature experiences three stages in the fuel lifetime. The mechanical results indicate that hoop strain at high power is greater than that at low power, which means that gap closure phenomenon will occur earlier at high power rates. The maximum cladding stress meets the requirement of yield strength limitation in the entire fuel lifetime. All results show that there are enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The FROBA code is expected to be applied to deal with more complicated fuel rod scenarios after some modifications.

  18. The Analysis of SBWR Critical Power Bundle Using Cobrag Code

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2013-03-01

    Full Text Available The coolant mechanism of SBWR is similar with the Dodewaard Nuclear Power Plant (NPP in the Netherlands that first went critical in 1968. The similarity of both NPP is cooled by natural convection system. These coolant concept is very related with same parameters on fuel bundle design especially fuel bundle length, core pressure drop and core flow rate as well as critical power bundle. The analysis was carried out by using COBRAG computer code. COBRAG computer code is GE Company proprietary. Basically COBRAG computer code is a tool to solve compressible three-dimensional, two fluid, three field equations for two phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. This code has been applied to analyses model flow and heat transfer within the reactor core. This volume describes the finitevolume equations and the numerical solution methods used to solve these equations. This analysis of same parameters has been done i.e.; inlet sub cooling 20 BTU/lbm and 40 BTU/lbm, 1000 psi pressure and R-factor is 1.038, mass flux are 0.5 Mlb/hr.ft2, 0.75 Mlb/hr.ft2, 1.00 Mlb/hr.ft2 and 1.25 Mlb/hr.ft2. Those conditions based on history operation of some type of the cell fuel bundle line at GE Nuclear Energy. According to the results, it can be concluded that SBWR critical power bundle is 10.5 % less than current BWR critical power bundle with length reduction of 12 ft to 9 ft.

  19. Performance analysis of WS-EWC coded optical CDMA networks with/without LDPC codes

    Science.gov (United States)

    Huang, Chun-Ming; Huang, Jen-Fa; Yang, Chao-Chin

    2010-10-01

    One extended Welch-Costas (EWC) code family for the wavelength-division-multiplexing/spectral-amplitude coding (WDM/SAC; WS) optical code-division multiple-access (OCDMA) networks is proposed. This system has a superior performance as compared to the previous modified quadratic congruence (MQC) coded OCDMA networks. However, since the performance of such a network is unsatisfactory when the data bit rate is higher, one class of quasi-cyclic low-density parity-check (QC-LDPC) code is adopted to improve that. Simulation results show that the performance of the high-speed WS-EWC coded OCDMA network can be greatly improved by using the LDPC codes.

  20. Development status of Severe Accident Analysis Code SAMPSON

    International Nuclear Information System (INIS)

    Iwashita, Tsuyoshi; Ujita, Hiroshi

    2000-01-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology' project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Test data are as follows: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex-Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analyzed and phenomena occurred in scenarios can be simulated qualitatively reasonably considering the physical models used for the situation. The Ministry of International Trade and Industry, Japan sponsors this work. (author)

  1. User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL Ver.2 for reactor pressure vessel (Contract research)

    International Nuclear Information System (INIS)

    Osakabe, Kazuya; Onizawa, Kunio; Shibata, Katsuyuki; Kato, Daisuke

    2006-09-01

    As a part of the aging structural integrity research for LWR components, the probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics and computer performance. PASCAL Ver.1 has functions of optimized sampling in the stratified Monte Carlo simulation, elastic-plastic fracture criterion of the R6 method, crack growth analysis models for a semi-elliptical crack, recovery of fracture toughness due to thermal annealing and so on. Since then, under the contract between the Ministry of Economy, Trading and Industry of Japan and JAEA, we have continued to develop and introduce new functions into PASCAL Ver.2 such as the evaluation method for an embedded crack, K I database for a semi-elliptical crack considering stress discontinuity at the base/cladding interface, PTS transient database, and others. A generalized analysis method is proposed on the basis of the development of PASCAL Ver.2 and results of sensitivity analyses. Graphical user interface (GUI) including a generalized method as default values has been also developed for PASCAL Ver.2. This report provides the user's manual and theoretical background of PASCAL Ver.2. (author)

  2. The stress analysis and stress evaluates of γ-spectrometer-probe station

    International Nuclear Information System (INIS)

    Li Hailong

    2005-01-01

    γ-Spectrometer -Probe Station is used for monitoring the reactor core fuel assemblies. The structural framework of this equipment possessed the massive lead bricks and linear supports. The article uses the finite element method and the conversion density method for processing lead bricks. Using shell element makes the analysis of liberating shape. The rigid supports are proposed and the stacking of the lead-bricks is improved. Meanwhile, the optimized design has been conducted for the equipment component. Using the computed results, the stress evaluate of the equipment is strictly made according to the ASME codes and standards. (author)

  3. A simple in-surge pressure analysis using the SPACE code

    International Nuclear Information System (INIS)

    Youn, Bum Soo; Kim, Yo Han; Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2010-01-01

    Currently, nuclear safety analysis codes used in Korea are developed by all the overseas. These codes are paying huge fee and permission must be obtained for use in the country. In addition, orders for nuclear power plants must ensure the safety analysis code for independent domestic technology. Therefore, Korea Electric Power Research Institute(KEPRI) is developing the domestic nuclear power safety analysis, SPACE(Safety and Performance Analysis Code for nuclear power plants). To determine the computational power of pressurizer model in development SPACE code, it was compared with existing commercial nuclear power safety analysis code, RETRAN

  4. On the development of LWR fuel analysis code (1). Analysis of the FEMAXI code and proposal of a new model

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Suzuki, Motoe

    2000-01-01

    This report summarizes the review on the modeling features of FEMAXI code and proposal of a new theoretical equation model of clad creep on the basis of irradiation-induced microstructure change. It was pointed out that plutonium build-up in fuel matrix and non-uniform radial power profile at high burn-up affect significantly fuel behavior through the interconnected effects with such phenomena as clad irradiation-induced creep, fission gas release, fuel thermal conductivity degradation, rim porous band formation and associated fuel swelling. Therefore, these combined effects should be properly incorporated into the models of the FEMAXI code so that the code can carry out numerical analysis at the level of accuracy and elaboration that modern experimental data obtained in test reactors have. Also, the proposed new mechanistic clad creep model has a general formalism which allows the model to be flexibly applied for clad behavior analysis under normal operation conditions and power transients as well for Zr-based clad materials by the use of established out-of-pile mechanical properties. The model has been tested against experimental data, while further verification is needed with specific emphasis on power ramps and transients. (author)

  5. Residual stress concerns in containment analysis

    International Nuclear Information System (INIS)

    Costantini, F.; Kulak, R. F.; Pfeiffer, P. A.

    1997-01-01

    The manufacturing of steel containment vessels starts with the forming of flat plates into curved plates. A steel containment structure is made by welding individual plates together to form the sections that make up the complex shaped vessels. The metal forming and welding process leaves residual stresses in the vessel walls. Generally, the effect of metal forming residual stresses can be reduced or virtually eliminated by thermally stress relieving the vesseL In large containment vessels this may not be practical and thus the residual stresses due to manufacturing may become important. The residual stresses could possibly tiect the response of the vessel to internal pressurization. When the level of residual stresses is significant it will affect the vessel's response, for instance the yielding pressure and possibly the failure pressure. The paper will address the effect of metal forming residual stresses on the response of a generic pressure vessel to internal pressurization. A scoping analysis investigated the effect of residual forming stresses on the response of an internally pressurized vessel. A simple model was developed to gain understanding of the mechanics of the problem. Residual stresses due to the welding process were not considered in this investigation

  6. Meanline Analysis of Turbines with Choked Flow in the Object-Oriented Turbomachinery Analysis Code

    Science.gov (United States)

    Hendricks, Eric S.

    2016-01-01

    The Object-Oriented Turbomachinery Analysis Code (OTAC) is a new meanline/streamline turbomachinery modeling tool being developed at NASA GRC. During the development process, a limitation of the code was discovered in relation to the analysis of choked flow in axial turbines. This paper describes the relevant physics for choked flow as well as the changes made to OTAC to enable analysis in this flow regime.

  7. 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.)

  8. 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

  9. 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

  10. 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

  11. 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)

  12. Analysis of Non-binary Hybrid LDPC Codes

    OpenAIRE

    Sassatelli, Lucile; Declercq, David

    2008-01-01

    In this paper, we analyse asymptotically a new class of LDPC codes called Non-binary Hybrid LDPC codes, which has been recently introduced. We use density evolution techniques to derive a stability condition for hybrid LDPC codes, and prove their threshold behavior. We study this stability condition to conclude on asymptotic advantages of hybrid LDPC codes compared to their non-hybrid counterparts.

  13. Code structure for U-Mo fuel performance analysis in high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

  14. Stress analysis of CVD diamond window for ECH system

    International Nuclear Information System (INIS)

    Takahashi, Koji

    2001-03-01

    The stress analysis of a chemical vapor deposition (CVD) diamond window for Electron Cyclotron Heating and Current Drive (ECH/ECCD) system of fusion reactors is described. It was found that the real size diamond window (φ aper =70mm, t=2.25mm) withstood 14.5 atm. (1.45 MPa). The calculation results of the diamond window by ABAQUS code agree well with the results of the pressure test. The design parameters of the torus diamond window for a vacuum and a safety barrier were also obtained. (author)

  15. Development of total systems of piping stress analysis and evaluation: ISAPPS

    International Nuclear Information System (INIS)

    Oki, Teizaburo; Koyanagi, Ryoichi; Fukuda, Masanao

    1978-01-01

    IHI has developed the systems of piping stress analysis and evaluation: ISAPPS (IHI Stress Analysis Program for Piping Systems), which are further described in this paper. In addition, the results of structural analysis and heat transfer analysis were confirmed. An example of stress evaluation in accordance with the modified ASME Code Sec. III is shown. ISAPPS consists of the following seven parts, and is designed for easy adoption of other programs by making modifications. 1. Piping design oriented language programs 2. Structural analysis programs 3. Isometric plotting programs 4. Multi-file dumping program 5. Load combination program 6. Heat transfer program 7. Stress evaluation programs As one of the examples of structural analysis programs, IHI make use of the modified SAP IV developed by the University of California. Evaluations of stresses are performed in accordance with: 1. ASME Boiler and Pressure Vessel Code, Sec. III Class 1, 2 and 3 2. ANSI Code, B31.1 and B31.3 3. MITI (Ministry of International Trade and Industry ) Code ISAPPS is very useful for design of nuclear and chemical pipings and so on. (author)

  16. HEFF---A user's manual and guide for the HEFF code for thermal-mechanical analysis using the boundary-element method

    International Nuclear Information System (INIS)

    St John, C.M.; Sanjeevan, K.

    1991-12-01

    The HEFF Code combines a simple boundary-element method of stress analysis with the closed form solutions for constant or exponentially decaying heat sources in an infinite elastic body to obtain an approximate method for analysis of underground excavations in a rock mass with heat generation. This manual describes the theoretical basis for the code, the code structure, model preparation, and step taken to assure that the code correctly performs its intended functions. The material contained within the report addresses the Software Quality Assurance Requirements for the Yucca Mountain Site Characterization Project. 13 refs., 26 figs., 14 tabs

  17. Systemization of burnup sensitivity analysis code (2) (Contract research)

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2008-08-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

  18. Concept Analysis: Alzheimer’s Caregiver Stress

    Science.gov (United States)

    Llanque, Sarah; Savage, Lynette; Rosenburg, Neal; Honor’s, BA; Caserta, Michael

    2015-01-01

    AIM The aim of this article was to analyze the concept of caregiver stress in the context of caring for a person with Alzheimer’s disease and related dementias. BACKGROUND Currently, there are more than 15 million unpaid care-givers for persons suffering from Alzheimer’s disease and related dementias. This unpaid care can be stressful for caregivers due to the chronic nature of the disease process, as well as other factors. METHOD The paper incorporates the modified method of Wilson’s concept analysis procedure to analyze the concept of caregiver stress. DATA SOURCES A review of the literature was undertaken using the Cumulative Index to Nursing and Allied Health Literature, Google Scholar, and PubMed. RESULTS A theoretical definition of caregiver stress is provided, and the defining attributes, related concepts, antecedents, and consequences of caregiver stress are proposed, and case studies are presented. CONCLUSIONS The analysis demonstrates that caregiver stress is the unequal exchange of assistance among people who stand in close relationship to one another, which results in emotional and physical stress on the caregiver. Implications for future nursing research and practice conclude the paper. PMID:24787468

  19. Analysis of Isp-42, panda test with the spectra code

    International Nuclear Information System (INIS)

    Stempniewicz, M.M.

    2001-01-01

    International Standard Problems (ISP) are organized in order to assess the ability of computer codes to predict the outcome of accidents in Nuclear Power Plants. The ISP-42 test was performed at Paul Scherrer Institute in 1998, as a sequence of six phases, Phase A through F Blind and open calculations of ISP-42 were performed with the computer code SPECTRA for each of the six phases. SPECTRA is a general tool for thermal-hydraulic analyses. Results of blind calculations are in good agreement with experiment. For open calculations several modifications were made in the model. These were mainly corrections of some input errors made in the model used for blind analysis. Some small improvements to the nodalization were made. Results of open calculations are generally closer to the experiment than the blind results. For phase D the containment pressure prediction was somewhat worse in the open calculation. Based on comparisons of blind and open results with experiment several conclusions may be drawn: - use of long ID structures, in contact with pool and atmosphere should be avoided, - PCC units are better represented with larger amount of Control Volumes, - two parallel junctions should be used to represent large openings between vessels, like drywell air line, etc., - careful verification of input decks is needed, - stratification models in SPECTRA are useful for cases with light gas injection; for complex cases a complementary SPECTRA-CFD analysis may be performed. (author)

  20. Stress analysis for robot arm version 2

    International Nuclear Information System (INIS)

    Anwar Abdul Rahman; Fikri, A.; Salleh, M. S.; Mohd Arif Hamzah; Azraf Azman; Rosli Darmawan; Mohd Rizal Mamat

    2010-01-01

    The design of a robot needs to be analyzed to ensure the specification and requirement by the user is full filled. Therefore, stress analysis has been performed on the robot arm version 2 after its complete fabrication. This paper discusses the result of the analysis and proposed measures to improve the future design of robot arm. (author)

  1. Investigation of alpha experiment by severe accident analysis code SAMPSON

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi; Naitoh, Masanori

    2006-01-01

    The severe accident analysis code SAMPSON is adopted in this work to evaluate its capability of reproducing the complex gap cooling phenomenon. The ALPHA experiment is adopted for validation, where molten aluminum oxide (Al 2 O 3 ) produced by a thermite reaction is poured into a water filled hemispherical vessel at the ambient pressure of approximately 1.3 MPa. The spreading and cooling of the debris that has relocated into the pressure vessel lower plenum are simulated, including the analysis of the RPV failure. The model included in the core to mimic the water penetration inside the gap is evaluated and improvements are proposed. The importance of the introduction of some mechanistic approach to describe the gap formation and evolution is underlined, where the results show its necessity in order to correctly reproduce the experimental trends. (author)

  2. Optimization and Validation of the Developed Uranium Isotopic Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    γ-ray spectroscopy is a representative non-destructive assay for nuclear material, and less time-consuming and less expensive than the destructive analysis method. The destructive technique is more precise than NDA technique, however, there is some correction algorithm which can improve the performance of γ-spectroscopy. For this reason, an analysis code for uranium isotopic analysis is developed by Applied Nuclear Physics Group in Seoul National University. Overlapped γ- and x-ray peaks in the 89-101 keV X{sub α}-region are fitted with Gaussian and Lorentzian distribution peak functions, tail and background functions. In this study, optimizations for the full-energy peak efficiency calibration and fitting parameters of peak tail and background are performed, and validated with 24 hour acquisition of CRM uranium samples. The optimization of peak tail and background parameters are performed with the validation by using CRM uranium samples. The analysis performance is improved in HEU samples, but more optimization of fitting parameters is required in LEU sample analysis. In the future, the optimization research about the fitting parameters with various type of uranium samples will be performed. {sup 234}U isotopic analysis algorithms and correction algorithms (coincidence effect, self-attenuation effect) will be developed.

  3. Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

    2015-08-15

    Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.

  4. Stress analysis in a non axisymmetric loaded reactor pressure vessel

    International Nuclear Information System (INIS)

    Albuquerque, Levi Barcelos; Assis, Gracia Menezes V. de; Miranda, Carlos Alexandre J.; Cruz, Julio Ricardo B.; Mattar Neto, Miguel

    1995-01-01

    In this work we intend to present the stress analysis of a PWR vessel under postulated concentrated loads. The vessel was modeled with Axisymmetric solid 4 nodes harmonic finite elements with the use of the ANSYS program, version 5.0. The bolts connecting the vessel flanges were modeled with beam elements. Some considerations were made to model the contact between the flanges. The perforated part of the vessel tori spherical head was modeled (with reduced properties due to its holes) to introduce its stiffness and loads but was not within the scope of this work. The loading consists of some usual ones, as pressure, dead weight, bolts preload, seismic load and some postulated ones as concentrated loads, over the vessel, modeled by Fourier Series. The results in the axisymmetric model are taken in terms of linearized stresses, obtained in some circumferential positions and for each position, in some sections along the vessel. Using the ASME Code (Section III, Division 1, Sub-section NB) the stresses are within the allowable limits. In order to draw some conclusions about stress linearization, the membrane plus bending stresses (Pl + Pb) are obtained and compared in some sections, using three different methods. (author)

  5. Analysis of selected Halden overpressure tests using the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G., E-mail: grigori.khvostov@psi.ch [Paul Scherrer Institut, CH 5232 Villigen PSI (Switzerland); Wiesenack, W. [Institute for Energy Technology – OECD Halden Reactor Project, P.O. Box 173, N-1751 Halden (Norway)

    2016-12-15

    Highlights: • We analyse four Halden overpressure tests. • We determine a critical overpressure value for lift-off in a BWR fuel sample. • We show the role of bonding in over-pressurized rod behaviour. • We analytically quantify the degree of bonding via its impact on cladding elongation. • We hypothesize on an effect of circumferential cracks on thermal fuel response to overpressure. • We estimate a thermal effect of circumferential cracks based on interpretation of the data. - Abstract: Four Halden overpressure (lift-off) tests using samples with uranium dioxide fuel pre-irradiated in power reactors to a burnup of 60 MWd/kgU are analyzed. The FALCON code coupled to a mechanistic model, GRSW-A for fission gas release and gaseous-bubble swelling is used for the calculation. The advanced version of the FALCON code is shown to be applicable to best-estimate predictive analysis of overpressure tests using rods without, or weak pellet-cladding bonding, as well as scoping analysis of tests with fuels where stronger pellet-cladding bonding occurs. Significant effects of bonding and fuel cracking/relocation on the thermal and mechanical behaviour of highly over-pressurized rods are shown. The effect of bonding is particularly pronounced in the tests with the PWR samples. The present findings are basically consistent with an earlier analysis based on a direct interpretation of the experimental data. Additionally, in this paper, the specific effects are quantified based on the comparison of the data with the results of calculation. It is concluded that the identified effects are largely beyond the current traditional fuel-rod licensing analysis methods.

  6. 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

  7. 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.

  8. Preliminary Coupling of MATRA Code for Multi-physics Analysis

    International Nuclear Information System (INIS)

    Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

    2014-01-01

    The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described

  9. Corporate governance codes and their contents : An analysis of Eastern European codes

    NARCIS (Netherlands)

    Hermes, Niels; Postma, Theo J. B. M.; Zivkov, Orestis

    2007-01-01

    Existing literature suggests that the contents of corporate governance codes are similar due to external forces, such as increased integration of countries in the global economy, the increased role of foreign institutional investors and recommendations on corporate governance practices of

  10. Development of a safety analysis code for molten salt reactors

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui

    2009-01-01

    The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.

  11. Analysis of visual coding variables on CRT generated displays

    International Nuclear Information System (INIS)

    Blackman, H.S.; Gilmore, W.E.

    1985-01-01

    Cathode ray tube generated safety parameter display systems in a nuclear power plant control room situation have been found to be improved in effectiveness when color coding is employed. Research has indicated strong support for graphic coding techniques particularly in redundant coding schemes. In addition, findings on pictographs, as applied in coding schemes, indicate the need for careful application and for further research in the development of a standardized set of symbols

  12. Fuel analysis code FAIR and its high burnup modelling capabilities

    International Nuclear Information System (INIS)

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

    1995-01-01

    A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs

  13. Ethical Code Effectiveness in Football Clubs: A Longitudinal Analysis

    OpenAIRE

    Constandt, Bram; De Waegeneer, Els; Willem, Annick

    2017-01-01

    As football (soccer) clubs are facing different ethical challenges, many clubs are turning to ethical codes to counteract unethical behaviour. However, both in- and outside the sport field, uncertainty remains about the effectiveness of these ethical codes. For the first time, a longitudinal study design was adopted to evaluate code effectiveness. Specifically, a sample of non-professional football clubs formed the subject of our inquiry. Ethical code effectiveness was...

  14. Micromechanical combined stress analysis: MICSTRAN, a user manual

    Science.gov (United States)

    Naik, R. A.

    1992-01-01

    Composite materials are currently being used in aerospace and other applications. The ability to tailor the composite properties by the appropriate selection of its constituents, the fiber and matrix, is a major advantage of composite materials. The Micromechanical Combined Stress Analysis (MICSTRAN) code provides the materials engineer with a user-friendly personal computer (PC) based tool to calculate overall composite properties given the constituent fiber and matrix properties. To assess the ability of the composite to carry structural loads, the materials engineer also needs to calculate the internal stresses in the composite material. MICSTRAN is a simple tool to calculate such internal stresses with a composite ply under combined thermomechanical loading. It assumes that the fibers have a circular cross-section and are arranged either in a repeating square or diamond array pattern within a ply. It uses a classical elasticity solution technique that has been demonstrated to calculate accurate stress results. Input to the program consists of transversely isotropic fiber properties and isotropic matrix properties such as moduli, Poisson's ratios, coefficients of thermal expansion, and volume fraction. Output consists of overall thermoelastic constants and stresses. Stresses can be computed under the combined action of thermal, transverse, longitudinal, transverse shear, and longitudinal shear loadings. Stress output can be requested along the fiber-matrix interface, the model boundaries, circular arcs, or at user-specified points located anywhere in the model. The MICSTRAN program is Windows compatible and takes advantage of the Microsoft Windows graphical user interface which facilitates multitasking and extends memory access far beyond the limits imposed by the DOS operating system.

  15. Analysis of mixed oxide fuel critical experiments with neutronics analysis codes for boiling water reactors

    International Nuclear Information System (INIS)

    Tamitani, Masashi; Maruyama, Hiromi; Ishii, Kazuya; Izutsu, Sadayuki; Yamaguchi, Masao

    2000-01-01

    Critical experiments of UO 2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were analyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%Δk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO 2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO 2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO 2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP. (author)

  16. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

  17. Severe accident analysis using MARCH 1.0 code

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1987-09-01

    The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt

  18. DYNAVAC: a transient-vacuum-network analysis code

    International Nuclear Information System (INIS)

    Deis, G.A.

    1980-01-01

    This report discusses the structure and use of the program DYNAVAC, a new transient-vacuum-network analysis code implemented on the NMFECC CDC-7600 computer. DYNAVAC solves for the transient pressures in a network of up to twenty lumped volumes, interconnected in any configuration by specified conductances. Each volume can have an internal gas source, a pumping speed, and any initial pressure. The gas-source rates can vary with time in any piecewise-linear manner, and up to twenty different time variations can be included in a single problem. In addition, the pumping speed in each volume can vary with the total gas pumped in the volume, thus simulating the saturation of surface pumping. This report is intended to be both a general description and a user's manual for DYNAVAC

  19. Safety margin evaluation of pre-stressed concrete nuclear containment vessel model with BARC code ULCA

    International Nuclear Information System (INIS)

    Basha, S.M.; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Full text: Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian pressurised heavy water reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results and for prediction of safety margins of Indian PHWRs. The present paper highlights the analysis results for prestressed concrete containment vessel (PCCV) tested at Sandia National Labs, USA in a round robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd

  20. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    International Nuclear Information System (INIS)

    Shi, Chengbin; Cheng, Maosong; Liu, Guimin

    2016-01-01

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  1. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-08-15

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  2. Plant stress analysis technology deployment

    Energy Technology Data Exchange (ETDEWEB)

    Ebadian, M.A.

    1998-01-01

    Monitoring vegetation is an active area of laser-induced fluorescence imaging (LIFI) research. The Hemispheric Center for Environmental Technology (HCET) at Florida International University (FIU) is assisting in the transfer of the LIFI technology to the agricultural private sector through a market survey. The market survey will help identify the key eco-agricultural issues of the nations that could benefit from the use of sensor technologies developed by the Office of Science and Technology (OST). The principal region of interest is the Western Hemisphere, particularly, the rapidly growing countries of Latin America and the Caribbean. The analysis of needs will assure that the focus of present and future research will center on economically important issues facing both hemispheres. The application of the technology will be useful to the agriculture industry for airborne crop analysis as well as in the detection and characterization of contaminated sites by monitoring vegetation. LIFI airborne and close-proximity systems will be evaluated as stand-alone technologies and additions to existing sensor technologies that have been used to monitor crops in the field and in storage.

  3. Development of NONSTA code for the design and analysis of LMR high temperature structure

    International Nuclear Information System (INIS)

    Kim, Jong Bum; Lee, H. Y.; Yoo, B.

    1999-02-01

    Liquid metal reactor(LMR) operates at high temperature (500-550 dg C) and structural materials undergo complex deformation behavior like diffusion, dislocation glide, and dislocation climb due to high temperature environment. And the material life reduces rapidly due to the interaction of cavities created inside structural materials and high temperature fatigue cracks. Thus the establishment of high temperature structure analysis techniques is necessary for the reliability and safety evaluation of such structures. The objectives of this study are to develop NONSTA code as the subprogram of ABAQUS code adopting constitutive equations which can predict high temperature material behavior precisely and to build the systematic analysis procedures. The developed program was applied to the example problems such as the tensile analysis using exponential creep model and the repetitive tensile-compression analysis using Chaboche unified viscoplastic model. In addition, the problem of a plate with a center hole subjected to tensile load was solved to show the applicability of the program to multiaxial problem and the time dependent stress redistribution was observed. (Author). 40 refs., 2 tabs., 24 figs

  4. Behavior Analysis Usage with Behavior Tures Adoption for Malicious Code Detection on JAVASCRIPT Scenarios Example

    Directory of Open Access Journals (Sweden)

    Y. M. Tumanov

    2010-03-01

    Full Text Available The article offers the method of malicious JavaScript code detection, based on behavior analysis. Conceptions of program behavior, program state, an algorithm of malicious code detection are described.

  5. The WEST project mechanical analysis of the divertor structure according to the nuclear construction code

    Energy Technology Data Exchange (ETDEWEB)

    Larroque, S., E-mail: sebastien.larroque@cea.fr [CEA Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France); Portafaix, C. [ITER Organization, 13108 Saint-Paul-lez-Durance (France); Saille, A.; Doceul, L.; Bucalossi, J.; Samaille, F.; Freslon, S. de [CEA Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • Divertor structure is mainly loaded by electromagnetical forces. • A simplified FEM analysis give the stresses in the structure. • RCCM criteria are required for the sizing. • Refined finite element models are used for local overstresses. - Abstract: The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST project, launched in support to the ITER tungsten divertor strategy. The installation of coils inside the vacuum vessel led to the design of a divertor supporting platform able to meet the project requirements and the associated electromagnetic loads. This paper illustrates the design, the method and the results of the thermomechanical elastic stress analyses performed in 2012. The validation of the integrity of the structure is based on the compliance with RCCMR design criteria (even though these Design and Construction rules for Mechanical Components of nuclear installations are not required for such experimental fusion device). Several 3D analyses are performed with the ANSYS code. The major one is a global analysis of half structure which determinates the stresses in the main part of the components. It gives an idea of the areas which needs local analyses. It also provides the interface loads for junction studies or simplified local model.

  6. Heat transfer and thermal stress analysis in grooved tubes

    Indian Academy of Sciences (India)

    ANSYS (1997) computer code has been used to analyse the thermal ... The numerical method is used succesfully to solve the governing equations ... thermal stress is an important criterion for consideration in the design of new compact heat.

  7. Turbine Airfoil Optimization Using Quasi-3D Analysis Codes

    Directory of Open Access Journals (Sweden)

    Sanjay Goel

    2009-01-01

    Full Text Available A new approach to optimize the geometry of a turbine airfoil by simultaneously designing multiple 2D sections of the airfoil is presented in this paper. The complexity of 3D geometry modeling is circumvented by generating multiple 2D airfoil sections and constraining their geometry in the radial direction using first- and second-order polynomials that ensure smoothness in the radial direction. The flow fields of candidate geometries obtained during optimization are evaluated using a quasi-3D, inviscid, CFD analysis code. An inviscid flow solver is used to reduce the execution time of the analysis. Multiple evaluation criteria based on the Mach number profile obtained from the analysis of each airfoil cross-section are used for computing a quality metric. A key contribution of the paper is the development of metrics that emulate the perception of the human designer in visually evaluating the Mach Number distribution. A mathematical representation of the evaluation criteria coupled with a parametric geometry generator enables the use of formal optimization techniques in the design. The proposed approach is implemented in the optimal design of a low-pressure turbine nozzle.

  8. Code of Practice for Protection Against Ionizing Radiation Emitted from X-ray Analysis Equipment (1984)

    International Nuclear Information System (INIS)

    1984-01-01

    Appropriate working rules, safety features and monitoring requirements for general X-ray analysis units or equipment are laid down in this Code which is intended for users of such equipment. The Code advises that establishments draw up their own working procedures based on appropriate legislation and on the recommendations contained in this Code. The Code also describes the requirements for X-ray analysis equipment necessary to ensure safety. (NEA) [fr

  9. Analysis of Biaxially Stressed Bridge Deck Plates

    DEFF Research Database (Denmark)

    Jönsson, Jeppe; Bondum, Tommi Højer

    2012-01-01

    The ultimate state analysis of bridge deck plates at the intersection zone between main girders and transverse beams is complicated by biaxial membrane stresses, which may be in compression or tension in either direction depending on the bridge configuration and the specific location. This paper...

  10. Development of steady thermal-hydraulic analysis code for China advanced research reactor

    International Nuclear Information System (INIS)

    Tian Wenxi; Qiu Suizheng; Guo Yun; Su Guanghui; Jia Dounan; Liu Tiancai; Zhang Jianwei

    2006-01-01

    A multi-channel model steady-state thermal-hydraulic analysis code was developed for China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed flow distribution in the core was obtained. The result shows that the structure size plays the most important role in flow distribution and the influence of core power could be neglected under single-phase flow. The temperature field of fuel element under unsymmetrical cooling condition was also obtained, which is necessary for the further study such as stress analysis etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of hot channel was carried out and it is proved that all thermal-hydraulic parameters accord with the Safety Regulation of CARR. (authors)

  11. Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

    2007-01-01

    A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

  12. Stress analysis of a rupture disk

    International Nuclear Information System (INIS)

    Werne, R.W.

    1975-04-01

    The results of an elastic stress analysis of the rupture disk for an internal pressure of 45.5 MPa (6600 psi) indicate that the maximum von Mises stresses occur in the membrane and are on the order of 483 to 690 MPa (70,000 psi). This far exceeds the yield of the membrane material of 207 MPa (30,000 psi). These high stresses are expected since the membrane is designed to burst at that design pressure. The von Mises stresses in the rest of the body are less than 138 MPa (20,000 psi). An elastic-plastic analysis of the membrane alone subjected to the 45.5 MPa (6600 psi) pressure indicates that it becomes plastically unstable, i.e., it continues to deform under constant load. A second load case with a constant 6.9 MPa (1000 psi) pressure throughout the entire body (i.e., after release of pressure by burst of the membrane) was analyzed. The results indicate that the elastic von Mises stresses are less than 26.7 MPa (3880 psi) throughout the body. (U.S.)

  13. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  14. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  15. LOFT reactor vessel 290/sup 0/ downcomer stalk instrument penetration flange stress analysis

    Energy Technology Data Exchange (ETDEWEB)

    Finicle, D.P.

    1978-06-06

    The LOFT Reactor Vessel 290/sup 0/ Downcomer Stalk Instrument Penetration Flange Stress Analysis has been completed using normal operational and blowdown loading. A linear elastic analysis was completed using simplified hand analysis techniques. The analysis was in accordance with the 1977 ASME Boiler and Pressure Vessel Code, Section III, for a Class 1 component. Loading included internal pressure, bolt preload, and thermal gradients due to normal operating and blowdown.

  16. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  17. 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

  18. Uncertainty and sensitivity analysis using probabilistic system assessment code. 1

    International Nuclear Information System (INIS)

    Honma, Toshimitsu; Sasahara, Takashi.

    1993-10-01

    This report presents the results obtained when applying the probabilistic system assessment code under development to the PSACOIN Level 0 intercomparison exercise organized by the Probabilistic System Assessment Code User Group in the Nuclear Energy Agency (NEA) of OECD. This exercise is one of a series designed to compare and verify probabilistic codes in the performance assessment of geological radioactive waste disposal facilities. The computations were performed using the Monte Carlo sampling code PREP and post-processor code USAMO. The submodels in the waste disposal system were described and coded with the specification of the exercise. Besides the results required for the exercise, further additional uncertainty and sensitivity analyses were performed and the details of these are also included. (author)

  19. Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

    International Nuclear Information System (INIS)

    Maruyama, Soh; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Murakami, Tomoyuki.

    1988-09-01

    This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T 1-M ) with simulated fuel rods and fuel blocks. (author)

  20. Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

    Science.gov (United States)

    Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

    1988-09-01

    This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

  1. Transient analysis of ABWR reactor using a best estimate code

    International Nuclear Information System (INIS)

    Mizokami, S.; Kitamura, H.; Mototani, A.; Ono, H.

    2004-01-01

    Since the recirculation pumps are mounted internally within the ABWR, core flow will decrease rapidly in the event of a loss of their driving force. A rapid reduction in core flow may cause the onset of boiling transition (BT). Therefore, in order to prevent the onset of BT, a motor-generator (MG) set is added to the power supply system of the reactor internal pump (RIP). Recent studies, however, have shown that dryout within a fuel assembly over a short time period will result in only a small rise in fuel cladding temperature and thus does not pose a threat to fuel integrity. In response to this finding, the standards committee of the Atomic Energy Society of Japan (AESJ) has proposed a post-BT standard which incorporates a cladding temperature criterion. If it is assumed that the MG-set is not added to the RIP power supply system, the result of the safety analysis shows the onset of BT with a subsequent rise in fuel cladding temperature. Although BT occurs under the conservative assumptions of this safety analysis, a possibility exists that BT will not occur under actual operating conditions. The best estimate code TRACG was used to show that BT does not occur and that fuel integrity can be sufficiently maintained under actual conditions. (author)

  2. Analysis of fuel-coolant interaction with VAPEX code

    International Nuclear Information System (INIS)

    Melikhov, O.I.; Melikhov, V.I.; Sokolin, A.V.; Yakush, S.E.

    2004-01-01

    The analysis of the FARO L-33 test has been carried out with the VAPEX code in which a submodel for hydrogen release and transport was implemented. The FARO test was aimed at studying the premixing and quenching processes for large (about 100 kg) masses of corium. The specific features of the FARO L-33 test are: high subcooling (124 K), low pressure (4.1 bar), presence of non-condensable gas (argon) and triggered vapor explosion when melt reached the bottom of the vessel. A numerical simulation of FARO L-33 test was carried out using 2-D nodalization. The fragmentation model is based on the Saito correlation. The model for hydrogen release assumes direct proportionality between the total hydrogen mass release rate and the total fragmentation rate of the melt jet. The proportionality constant was taken from the experimental estimates for test conditions. Calculation of the premixing stage gave some delay in the pressure growth, which is most probably connected with inadequacy of the fragmentation model at the initial stage of melt jet-water interaction. The calculated pressurization rate, however, agrees reasonably with the measured one. Modeling of vapor explosion, which occurred in the test, yielded reasonable correlation with the test data when hydrogen formation was taken into account. Thus, VAPEX analysis of the FARO L-33 test has shown reasonable agreement between the experimental and calculated data. (author)

  3. Analysis of Memory Codes and Cumulative Rehearsal in Observational Learning

    Science.gov (United States)

    Bandura, Albert; And Others

    1974-01-01

    The present study examined the influence of memory codes varying in meaningfulness and retrievability and cumulative rehearsal on retention of observationally learned responses over increasing temporal intervals. (Editor)

  4. TRANSURANUS: A fuel rod analysis code ready for use

    Energy Technology Data Exchange (ETDEWEB)

    Lassmann, K; O` Carroll, C; Van de Laar, J [Commission of the European Communities, Karlsruhe (Germany). European Inst. for Transuranium Elements; Ott, C [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-12-31

    The basic concepts of fuel rod performance codes are discussed. The TRANSURANUS code developed at the Institute for Transuranium Elements, Karlsruhe (GE) is presented. It is a quasi two-dimensional (1{sub 1/2}-D) code designed for treatment of a whole fuel rod for any type of reactor and any situation. The fuel rods found in the majority of test- or power reactors can be analyzed for very different situations (normal, off-normal and accidental). The time scale of the problems to be treated may range from milliseconds to years. The TRANSURANUS code consists of a clearly defined mechanical/mathematical framework into which physical models can easily be incorporated. This framework has been extensively tested and the programming very clearly reflects this structure. The code is well structured and easy to understand. It has a comprehensive material data bank for different fuels, claddings, coolants and their properties. The code can be employed in a deterministic and a statistical version. It is written in standard FORTRAN 77. The code system includes: 2 preprocessor programs (MAKROH and AXORDER) for setting up new data cases; the post-processor URPLOT for plotting all important quantities as a function of the radius, the axial coordinate or the time; the post-processor URSTART evaluating statistical analyses. The TRANSURANUS code exhibits short running times. A new WINDOWS-based interactive interface is under development. The code is now in use in various European institutions and is available to all interested parties. 7 figs., 15 refs.

  5. Analysis of Simple Creep Stress Calculation Methods for Creep Life Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jun Min; Lee, Han Sang; Kim, Yun Jae [Korea Univ., Daejeon (Korea, Republic of)

    2017-08-15

    Creep analysis takes much more time than elastic or elastic-plastic analysis. In this study, we conducted elastic and elastic-plastic analysis and compared the results with creep analysis results. In the elastic analysis, we used primary stress, which can be classified by the Mα-tangent method and stress intensities recommended in the ASME code. In the elastic-plastic analysis, we calculated the parameters recommended in the R5 code. For the FE models, a bending load, uniaxial load, and biaxial load were applied to the cross shaped welded plate, and a bending load and internal pressure were applied to the elbow pipe. To investigate the element size sensitivity, we conducted FE analysis for various element sizes for the cases where bending load was applied to the cross shaped welded plate. There was no significant difference between the creep.

  6. TEMP-STRESS analysis of a reinforced concrete vessel under internal pressure

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Kennedy, J.M.; Pfeiffer, P.A.

    1987-01-01

    Prediction of the response of the Sandia National laboratory 1/6-scale reinforced concrete containment model test was obtained by Argonne National Laboratory (ANL) employing a computer program developed by ANL. The test model was internally pressurized to failure. The two-dimensional code TEMP-STRESS [1-5] has been developed at ANL for stress analysis of plane and axisymmetric 2-D reinforced structures under various thermal conditions. The program is applicable to a wide variety of nonlinear problems, and is utilized in the present study. The comparison of these pretest computations with test data on the containment model should be a good indication of the state of the code

  7. Mitigation method of thermal transient stress by thermalhydraulic-structure total analysis

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Jinbo, Masakazu; Hosogai, Hiromi

    2003-01-01

    This study proposes a rational evaluation and mitigation method of thermal transient loads in fast reactor components by utilizing relationships among plant system parameters and stresses induced by thermal transients of plants. A thermalhydraulic-structure total analysis procedure helps us to grasp relationship among system parameters and thermal stresses. Furthermore, it enables mitigation of thermal transient loads by adjusting system parameters. In order to overcome huge computations, a thermalhydraulic-structure total analysis code and the Design of Experiments methodology are utilized. The efficiency of the proposed mitigation method is validated through thermal stress evaluation of an intermediate heat exchanger in Japanese demonstration fast reactor. (author)

  8. Development of severe accident analysis code - Development of a finite element code for lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Lee, Choong Ho; Choi, Tae Hoon; Kim, Hyun Sup; Kim, Se Ho; Kang, Woo Jong; Seo, Chong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-08-01

    The study concerns the development of analysis models and computer codes for lower head failure analysis when a severe accident occurs in a nuclear reactor system. Although the lower head failure modes consists of several failure modes, the study this year was focused on the global rupture with the collapse pressure and mode by limit analysis and elastic deformation. The behavior of molten core causes elevation of temperature in the reactor vessel wall and deterioration of load-carrying capacity of a reactor vessel. The behavior of molten core and the heat transfer modes were, therefore, postulated in several types and the temperature distributions according to the assumed heat flux modes were calculated. The collapse pressure of a nuclear reactor lower head decreases rapidly with elevation of temperature as time passes. The calculation shows the safety of a nuclear reactor is enhanced with the lager collapse pressure when the hot spot is located far from the pole. 42 refs., 2 tabs., 31 figs. (author)

  9. The Stress and Stiffness Analysis of Diaphragm

    Directory of Open Access Journals (Sweden)

    Qu Dongyue

    2017-01-01

    Full Text Available Diaphragm coupling with its simple structure, small size, high reliability, which can compensate for its input and output displacement deviation by its elastic deformation, is widely used in aerospace, marine, and chemical etc. This paper uses the ANSYS software and its APDL language to analysis the stress distribution when the diaphragm under the load of torque, axial deviation, centrifugal force, angular deviation and multiple loads. We find that the value of maximum stress usually appears in the outer or inner transition region and the axial deviation has a greater influence to the distribution of the stress. Based on above, we got three kinds of stiffness for axial, angular and torque, which the stiffness of diaphragm is nearly invariable. The results can be regard as an important reference for design and optimization of diaphragm coupling.

  10. PDX toroidal field coils stress analysis

    International Nuclear Information System (INIS)

    Nikodem, Z.D.; Smith, R.A.

    1975-01-01

    A method used in the stress analysis of the PDX toroidal field coil is developed. A multilayer coil design of arbitrary dimensions in the shape of either a circle or an oval is considered. The analytical model of the coil and the supporting coil case with connections to the main support structure is analyzed using the finite element technique. The three dimensional magnetic fields and the non-uniform body forces which are a loading condition on a coil due to toroidal and poloidal fields are calculated. The method of analysis permits rapid and economic evaluations of design changes in coil geometry as well as in coil support structures. Some results pertinent to the design evolution and their comparison are discussed. The results of the detailed stress analysis of the final coil design due to toroidal field, poloidal field and temperature loads are presented

  11. Photoelastic stress analysis in mitred bend under internal pressure

    International Nuclear Information System (INIS)

    Sawa, Yoshiaki

    1987-01-01

    The stress analysis and stress relaxation in mitred bend subjected to internal pressure have been studied by means of the photoelastic stress freezing method. The experimental results show that stress concentration occurs in the wedge tip of the intersectional plane and it is considerably influenced by the bent angle. Then, the stress relaxation was obtained by planing the wedge tip. (author)

  12. Choreographer Pre-Testing Code Analysis and Operational Testing.

    Energy Technology Data Exchange (ETDEWEB)

    Fritz, David J. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States); Harrison, Christopher B. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States); Perr, C. W. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States); Hurd, Steven A [Sandia National Laboratories (SNL-CA), Livermore, CA (United States)

    2014-07-01

    Choreographer is a "moving target defense system", designed to protect against attacks aimed at IP addresses without corresponding domain name system (DNS) lookups. It coordinates actions between a DNS server and a Network Address Translation (NAT) device to regularly change which publicly available IP addresses' traffic will be routed to the protected device versus routed to a honeypot. More details about how Choreographer operates can be found in Section 2: Introducing Choreographer. Operational considerations for the successful deployment of Choreographer can be found in Section 3. The Testing & Evaluation (T&E) for Choreographer involved 3 phases: Pre-testing, Code Analysis, and Operational Testing. Pre-testing, described in Section 4, involved installing and configuring an instance of Choreographer and verifying it would operate as expected for a simple use case. Our findings were that it was simple and straightforward to prepare a system for a Choreographer installation as well as configure Choreographer to work in a representative environment. Code Analysis, described in Section 5, consisted of running a static code analyzer (HP Fortify) and conducting dynamic analysis tests using the Valgrind instrumentation framework. Choreographer performed well, such that only a few errors that might possibly be problematic in a given operating situation were identified. Operational Testing, described in Section 6, involved operating Choreographer in a representative environment created through EmulyticsTM . Depending upon the amount of server resources dedicated to Choreographer vis-á-vis the amount of client traffic handled, Choreographer had varying degrees of operational success. In an environment with a poorly resourced Choreographer server and as few as 50-100 clients, Choreographer failed to properly route traffic over half the time. Yet, with a well-resourced server, Choreographer handled over 1000 clients without missrouting. Choreographer

  13. Code development for eigenvalue total sensitivity analysis and total uncertainty analysis

    International Nuclear Information System (INIS)

    Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Zu, Tiejun; Shen, Wei

    2015-01-01

    Highlights: • We develop a new code for total sensitivity and uncertainty analysis. • The implicit effects of cross sections can be considered. • The results of our code agree well with TSUNAMI-1D. • Detailed analysis for origins of implicit effects is performed. - Abstract: The uncertainties of multigroup cross sections notably impact eigenvalue of neutron-transport equation. We report on a total sensitivity analysis and total uncertainty analysis code named UNICORN that has been developed by applying the direct numerical perturbation method and statistical sampling method. In order to consider the contributions of various basic cross sections and the implicit effects which are indirect results of multigroup cross sections through resonance self-shielding calculation, an improved multigroup cross-section perturbation model is developed. The DRAGON 4.0 code, with application of WIMSD-4 format library, is used by UNICORN to carry out the resonance self-shielding and neutron-transport calculations. In addition, the bootstrap technique has been applied to the statistical sampling method in UNICORN to obtain much steadier and more reliable uncertainty results. The UNICORN code has been verified against TSUNAMI-1D by analyzing the case of TMI-1 pin-cell. The numerical results show that the total uncertainty of eigenvalue caused by cross sections can reach up to be about 0.72%. Therefore the contributions of the basic cross sections and their implicit effects are not negligible

  14. Complete stress tensor determination by microearthquake analysis

    Science.gov (United States)

    Slunga, R.

    2010-12-01

    the depth based on the assumptions of a fractured crust, widely vary ing stress field, and a general closeness to instability as found by stress measurements (Jamison and Cook 1976). Wheather this approach is working or not is best answered by applying it to real data. This was provided by the IMO network in Iceland. Along Southern Iceland Seismic Zone (SISZ) more than 200,000 microearthquakes and a few M 5 EQs and 2 M=6.6 EQs have been recorded. The results will be presented it is obvious that the use of the stresses determined from the microearthquake recordings may significa ntly improve earthquake warnings and will make it possible to use the absolute C FS method for more deterministic predictions. Note that the microearthquake meth od only shows the part of the stress field that has caused slip. Volumes with st able stress will not show up. However stress measurements (Brown and Hoek 1978, Slunga 1988) have shown that the crustal stresses in general are close to instabi lity and microearthquake source analysis has shown that a large number of differ ent fractures become unstable within longer time windows. This may explain the e xcellent results given by the Icelandic tests of the absolute stress tensor fiel d as given by the microearthquakes. However I prefer to call this stress apparen t.

  15. 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

  16. Description and validation of ANTEO, an optimised PC code the thermalhydraulic analysis of fuel bundles

    International Nuclear Information System (INIS)

    Cevolani, S.

    1995-01-01

    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 such a code was made possible by two facts: firstly, the increase, in the computing power of the desk machines; secondly, the fact that 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. (author)

  17. 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

  18. Performance Analysis of Optical Code Division Multiplex System

    Science.gov (United States)

    Kaur, Sandeep; Bhatia, Kamaljit Singh

    2013-12-01

    This paper presents the Pseudo-Orthogonal Code generator for Optical Code Division Multiple Access (OCDMA) system which helps to reduce the need of bandwidth expansion and improve spectral efficiency. In this paper we investigate the performance of multi-user OCDMA system to achieve data rate more than 1 Tbit/s.

  19. FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

    International Nuclear Information System (INIS)

    Valtonen, Keijo; Hamalainen, Anitta; Cunningham, Mitchel E.

    2002-01-01

    The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

  20. FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

    Energy Technology Data Exchange (ETDEWEB)

    Valtonen, Keijo (Radiation and Nuclear Safety Authority, Finland); Hamalainen, Anitta (VTT Energy, Finland); Cunningham, Mitchel E.(BATTELLE (PACIFIC NW LAB))

    2002-05-01

    The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

  1. Sample problem manual for benchmarking of cask analysis codes

    International Nuclear Information System (INIS)

    Glass, R.E.

    1988-02-01

    A series of problems have been defined to evaluate structural and thermal codes. These problems were designed to simulate the hypothetical accident conditions given in Title 10 of the Code of Federal Regulation, Part 71 (10CFR71) while retaining simple geometries. This produced a problem set that exercises the ability of the codes to model pertinent physical phenomena without requiring extensive use of computer resources. The solutions that are presented are consensus solutions based on computer analyses done by both national laboratories and industry in the United States, United Kingdom, France, Italy, Sweden, and Japan. The intent of this manual is to provide code users with a set of standard structural and thermal problems and solutions which can be used to evaluate individual codes. 19 refs., 19 figs., 14 tabs

  2. ANDREA: Advanced nodal diffusion code for reactor analysis

    International Nuclear Information System (INIS)

    Belac, J.; Josek, R.; Klecka, L.; Stary, V.; Vocka, R.

    2005-01-01

    A new macro code is being developed at NRI which will allow coupling of the advanced thermal-hydraulics model with neutronics calculations as well as efficient use in core loading pattern optimization process. This paper describes the current stage of the macro code development. The core simulator is based on the nodal expansion method, Helios lattice code is used for few group libraries preparation. Standard features such as pin wise power reconstruction and feedback iterations on critical control rod position, boron concentration and reactor power are implemented. A special attention is paid to the system and code modularity in order to enable flexible and easy implementation of new features in future. Precision of the methods used in the macro code has been verified on available benchmarks. Testing against Temelin PWR operational data is under way (Authors)

  3. 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

  4. Safety analysis and code development for nuclear fuel cycle facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    We are estimating that the debris containing fuel are piled in the containment and the pressure vessel bottoms of Fukushima-Daiichi NPPs. A radioactive Xe concentration discharged in recriticality is being monitored by utilizing the gas management system set up in NPPs unit 1-3. For this reason, we can confirm the recriticality might not be broken out. However, the debris conditions distributed in the containment vessel and the pressure vessel bottoms are not clear. The internal and external surrounding changes will make recriticality become possible. According to TEPCO's roadmap, TEPCO will launch extracting task within 10 years. Even in the case that the fuel condition changes due to debris relocation and mixture, subcriticality must be secured. Criticality safety analysis with non-uniform effect should therefore be essential for the molten debris. For above reasons, we studies the optimum distributions with some parameters that have a large reactivity change were assessed with OPT-DANT code. Finally, the boron concentration was estimated in order to keep subcriticality. (author)

  5. Flow analysis of tubular fuel assembly using CFD code

    International Nuclear Information System (INIS)

    Park, J. H.; Park, C.; Chae, H. T.

    2004-01-01

    Based on the experiences of HANARO, a new research reactor is under conceptual design preparing for future needs of research reactor. Considering various aspects such as nuclear physics, thermal-hydraulics, mechanical structure and the applicability of HANARO technology, a tubular type fuel has been considered as that of a new research reactor. Tubular type fuel has several circular fuel layers, and each layer consists of 3 curved fuel plates arranged with constant small gap to build up cooling channels. In the thermal-hydraulic point, it is very important to maintain each channel flow velocity be equal as much as possible, because the small gaps between curved thin fuel plates independently forms separate coolant channels, which may cause a thermal-hydraulic problem in certain conditions. In this study, commercial CFD(Computational Fluid Dynamics) code, Fluent, has been used to investigate flow characteristics of tubular type fuel assembly. According to the computation results for the preliminary conceptual design, there is a serious lack of uniformity of average velocity on the each coolant channel. Some changes for initial conceptual design were done to improve the balance of velocity distribution, and analysis was done again, too. The results for the revised design showed that the uniformity of each channel velocity was improved significantly. The influence of outermost channel gap width on the velocity distribution was also examined

  6. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  7. Stress analysis and collapse time prediction of nuclear fuel cladding tube with wear scar

    International Nuclear Information System (INIS)

    Lee, J. S.; Kim, O. H.; Kim, H. K.; Hu, Y. H.; Kim, J. I.; Kim, K. T.

    2004-01-01

    In this analysis, the stress and collapse time analysis models for nuclear fuel rod with the fretting wear scar were developed in order to evaluate the effects of the wear depth on the integrity of nuclear fuel rod. The stress analysis result shows that the nuclear fuel rod with approximately 60% deep wear scar of the clad wall thickness, meets the allowable stress criteria and the collapse time analysis indicates that the fuel rod with less than roughly 56% deep wear scar of the clad wall thickness has longer collapse time than the expected fuel life-time. The both stress and collapse time results are evaluated to be very reasonable on considering the comparison with the outputs of existing design code for the simple model. However, the developed analysis models and the results will be confirmed by the tests

  8. Performance Analysis of Faulty Gallager-B Decoding of QC-LDPC Codes with Applications

    Directory of Open Access Journals (Sweden)

    O. Al Rasheed

    2014-06-01

    Full Text Available In this paper we evaluate the performance of Gallager-B algorithm, used for decoding low-density parity-check (LDPC codes, under unreliable message computation. Our analysis is restricted to LDPC codes constructed from circular matrices (QC-LDPC codes. Using Monte Carlo simulation we investigate the effects of different code parameters on coding system performance, under a binary symmetric communication channel and independent transient faults model. One possible application of the presented analysis in designing memory architecture with unreliable components is considered.

  9. 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)

  10. Performance Analysis of DPSK Signals with Selection Combining and Convolutional Coding in Fading Channel

    National Research Council Canada - National Science Library

    Ong, Choon

    1998-01-01

    The performance analysis of a differential phase shift keyed (DPSK) communications system, operating in a Rayleigh fading environment, employing convolutional coding and diversity processing is presented...

  11. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    International Nuclear Information System (INIS)

    Wan, C.; Cao, L.; Wu, H.; Zu, T.; Shen, W.

    2015-01-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  12. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Wan, C.; Cao, L.; Wu, H.; Zu, T., E-mail: chenghuiwan@stu.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: tiejun@mail.xjtu.edu.cn [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Shen, W., E-mail: Wei.Shen@cnsc-ccsn.gc.ca [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-07-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  13. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    Majalee, Aaditya V.; Chaturvedi, S.

    2015-01-01

    CAD has been developing a computer code 'FRAVIZ' for calculation of steady-state thermomechanical behaviour of nuclear reactor fuel rods. It contains four major modules viz., Thermal module, Fission Gas Release module, Material Properties module and Mechanical module. All these four modules are coupled to each other and feedback from each module is fed back to others to get a self-consistent evolution in time. The computer code has been checked against two FUMEX benchmarks. Modelling fuel performance in Advance Heavy Water Reactor would require additional inputs related to the fuel and some modification in the code.(author)

  14. Performance analysis of LDPC codes on OOK terahertz wireless channels

    International Nuclear Information System (INIS)

    Liu Chun; Wang Chang; Cao Jun-Cheng

    2016-01-01

    Atmospheric absorption, scattering, and scintillation are the major causes to deteriorate the transmission quality of terahertz (THz) wireless communications. An error control coding scheme based on low density parity check (LDPC) codes with soft decision decoding algorithm is proposed to improve the bit-error-rate (BER) performance of an on-off keying (OOK) modulated THz signal through atmospheric channel. The THz wave propagation characteristics and channel model in atmosphere is set up. Numerical simulations validate the great performance of LDPC codes against the atmospheric fading and demonstrate the huge potential in future ultra-high speed beyond Gbps THz communications. (paper)

  15. 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.

  16. Development of safety analysis codes for light water reactor

    International Nuclear Information System (INIS)

    Akimoto, Masayuki

    1985-01-01

    An overview is presented of currently used major codes for the prediction of thermohydraulic transients in nuclear power plants. The overview centers on the two-phase fluid dynamics of the coolant system and the assessment of the codes. Some of two-phase phenomena such as phase separation are not still predicted with engineering accuracy. MINCS-PIPE are briefly introduced. The MINCS-PIPE code is to assess constitutive relations and to aid development of various experimental correlations for 1V1T model to 2V2T model. (author)

  17. Development of dynamic explicit crystallographic homogenization finite element analysis code to assess sheet metal formability

    International Nuclear Information System (INIS)

    Nakamura, Yasunori; Tam, Nguyen Ngoc; Ohata, Tomiso; Morita, Kiminori; Nakamachi, Eiji

    2004-01-01

    The crystallographic texture evolution induced by plastic deformation in the sheet metal forming process has a great influence on its formability. In the present study, a dynamic explicit finite element (FE) analysis code is newly developed by introducing a crystallographic homogenization method to estimate the polycrystalline sheet metal formability, such as the extreme thinning and 'earing'. This code can predict the plastic deformation induced texture evolution at the micro scale and the plastic anisotropy at the macro scale, simultaneously. This multi-scale analysis can couple the microscopic crystal plasticity inhomogeneous deformation with the macroscopic continuum deformation. In this homogenization process, the stress at the macro scale is defined by the volume average of those of the corresponding microscopic crystal aggregations in satisfying the equation of motion and compatibility condition in the micro scale 'unit cell', where the periodicity of deformation is satisfied. This homogenization algorithm is implemented in the conventional dynamic explicit finite element code by employing the updated Lagrangian formulation and the rate type elastic/viscoplastic constitutive equation.At first, it has been confirmed through a texture evolution analyses in cases of typical deformation modes that Taylor's 'constant strain homogenization algorithm' yields extreme concentration toward the preferred crystal orientations compared with our homogenization one. Second, we study the plastic anisotropy effects on 'earing' in the hemispherical cup deep drawing process of pure ferrite phase sheet metal. By the comparison of analytical results with those of Taylor's assumption, conclusions are drawn that the present newly developed dynamic explicit crystallographic homogenization FEM shows more reasonable prediction of plastic deformation induced texture evolution and plastic anisotropy at the macro scale

  18. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  19. A comparison of LOCA analysis using SMOKIN and CERBERUS codes

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

  20. Vehicle Codes and Standards: Overview and Gap Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Blake, C.; Buttner, W.; Rivkin, C.

    2010-02-01

    This report identifies gaps in vehicle codes and standards and recommends ways to fill the gaps, focusing on six alternative fuels: biodiesel, natural gas, electricity, ethanol, hydrogen, and propane.

  1. Exposure calculation code module for reactor core analysis: BURNER

    International Nuclear Information System (INIS)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules

  2. Evaluation of conservatisms and environmental effects in ASME Code, Section III, Class 1 fatigue analysis

    International Nuclear Information System (INIS)

    Deardorff, A.F.; Smith, J.K.

    1994-08-01

    This report documents the results of a study regarding the conservatisms in ASME Code Section 3, Class 1 component fatigue evaluations and the effects of Light Water Reactor (LWR) water environments on fatigue margins. After review of numerous Class 1 stress reports, it is apparent that there is a substantial amount of conservatism present in many existing component fatigue evaluations. With little effort, existing evaluations could be modified to reduce the overall predicted fatigue usage. Areas of conservatism include design transients considerably more severe than those experienced during service, conservative grouping of transients, conservatisms that have been removed in later editions of Section 3, bounding heat transfer and stress analysis, and use of the ''elastic-plastic penalty factor'' (K 3 ). Environmental effects were evaluated for two typical components that experience severe transient thermal cycling during service, based on both design transients and actual plant data. For all reasonable values of actual operating parameters, environmental effects reduced predicted margins, but fatigue usage was still bounded by the ASME Section 3 fatigue design curves. It was concluded that the potential increase in predicted fatigue usage due to environmental effects should be more than offset by decreases in predicted fatigue usage if re-analysis were conducted to reduce the conservatisms that are present in existing component fatigue evaluations

  3. Evolutionary analysis reveals regulatory and functional landscape of coding and non-coding RNA editing.

    Science.gov (United States)

    Zhang, Rui; Deng, Patricia; Jacobson, Dionna; Li, Jin Billy

    2017-02-01

    Adenosine-to-inosine RNA editing diversifies the transcriptome and promotes functional diversity, particularly in the brain. A plethora of editing sites has been recently identified; however, how they are selected and regulated and which are functionally important are largely unknown. Here we show the cis-regulation and stepwise selection of RNA editing during Drosophila evolution and pinpoint a large number of functional editing sites. We found that the establishment of editing and variation in editing levels across Drosophila species are largely explained and predicted by cis-regulatory elements. Furthermore, editing events that arose early in the species tree tend to be more highly edited in clusters and enriched in slowly-evolved neuronal genes, thus suggesting that the main role of RNA editing is for fine-tuning neurological functions. While nonsynonymous editing events have been long recognized as playing a functional role, in addition to nonsynonymous editing sites, a large fraction of 3'UTR editing sites is evolutionarily constrained, highly edited, and thus likely functional. We find that these 3'UTR editing events can alter mRNA stability and affect miRNA binding and thus highlight the functional roles of noncoding RNA editing. Our work, through evolutionary analyses of RNA editing in Drosophila, uncovers novel insights of RNA editing regulation as well as its functions in both coding and non-coding regions.

  4. Piping stress analysis with personal computers

    International Nuclear Information System (INIS)

    Revesz, Z.

    1987-01-01

    The growing market of the personal computers is providing an increasing number of professionals with unprecedented and surprisingly inexpensive computing capacity, which if using with powerful software, can enhance immensely the engineers capabilities. This paper focuses on the possibilities which opened in piping stress analysis by the widespread distribution of personal computers, on the necessary changes in the software and on the limitations of using personal computers for engineering design and analysis. Reliability and quality assurance aspects of using personal computers for nuclear applications are also mentioned. The paper resumes with personal views of the author and experiences gained during interactive graphic piping software development for personal computers. (orig./GL)

  5. Development of in-vessel source term analysis code, tracer

    International Nuclear Information System (INIS)

    Miyagi, K.; Miyahara, S.

    1996-01-01

    Analyses of radionuclide transport in fuel failure accidents (generally referred to source terms) are considered to be important especially in the severe accident evaluation. The TRACER code has been developed to realistically predict the time dependent behavior of FPs and aerosols within the primary cooling system for wide range of fuel failure events. This paper presents the model description, results of validation study, the recent model advancement status of the code, and results of check out calculations under reactor conditions. (author)

  6. Demonstration study on shielding safety analysis code (8)

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  7. Performance Analysis of a Decoding Algorithm for Algebraic Geometry Codes

    DEFF Research Database (Denmark)

    Jensen, Helge Elbrønd; Nielsen, Rasmus Refslund; Høholdt, Tom

    1998-01-01

    We analyse the known decoding algorithms for algebraic geometry codes in the case where the number of errors is greater than or equal to [(dFR-1)/2]+1, where dFR is the Feng-Rao distance......We analyse the known decoding algorithms for algebraic geometry codes in the case where the number of errors is greater than or equal to [(dFR-1)/2]+1, where dFR is the Feng-Rao distance...

  8. 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

  9. Code Blue Emergencies: A Team Task Analysis and Educational Initiative.

    Science.gov (United States)

    Price, James W; Applegarth, Oliver; Vu, Mark; Price, John R

    2012-01-01

    The objective of this study was to identify factors that have a positive or negative influence on resuscitation team performance during emergencies in the operating room (OR) and post-operative recovery unit (PAR) at a major Canadian teaching hospital. This information was then used to implement a team training program for code blue emergencies. In 2009/10, all OR and PAR nurses and 19 anesthesiologists at Vancouver General Hospital (VGH) were invited to complete an anonymous, 10 minute written questionnaire regarding their code blue experience. Survey questions were devised by 10 recovery room and operation room nurses as well as 5 anesthesiologists representing 4 different hospitals in British Columbia. Three iterations of the survey were reviewed by a pilot group of nurses and anesthesiologists and their feedback was integrated into the final version of the survey. Both nursing staff (n = 49) and anesthesiologists (n = 19) supported code blue training and believed that team training would improve patient outcome. Nurses noted that it was often difficult to identify the leader of the resuscitation team. Both nursing staff and anesthesiologists strongly agreed that too many people attending the code blue with no assigned role hindered team performance. Identifiable leadership and clear communication of roles were identified as keys to resuscitation team functioning. Decreasing the number of people attending code blue emergencies with no specific role, increased access to mock code blue training, and debriefing after crises were all identified as areas requiring improvement. Initial team training exercises have been well received by staff.

  10. Business Ethics: International Analysis of Codes of Ethics and Conduct

    Directory of Open Access Journals (Sweden)

    Josmar Andrade

    2017-03-01

    Full Text Available Codes of ethics and code of conduct formalize an ideal of expected behavior patterns to managers and employees of organizations, providing standards and orientation that states companies interactions with the community, through products /services, sales force, marketing communications, investments, and relationships with other stakeholders, influencing company reputation and overall Marketing performance. The objective of this study is to analyze the differences in codes of ethics of the largest companies based in Brazil and in Portugal, given their cultural and linguistic similarities. Findings show that the use of codes of ethics are more common in Brazil than in Portugal and that codes of ethics are substantially more extensive and cover a larger number of categories in Brazilian companies, reflecting the organizations’ mission and perception of stakeholders concerns and priorities. We conclude that ethical issues severely impact company reputation and, in a comprehensive sense, overall Marketing performance. Marketing professionals should be systematically aware of how company core values are transmitted to different audiences, including the use of code of ethics to communicate both with internal and external publics. 0 0 1 171 966 CASA DOS ANDRADES 23 14 1123 14.0 96 800x600 Normal 0 false false false EN-US JA X-NONE  

  11. Promoter Analysis Reveals Globally Differential Regulation of Human Long Non-Coding RNA and Protein-Coding Genes

    KAUST Repository

    Alam, Tanvir

    2014-10-02

    Transcriptional regulation of protein-coding genes is increasingly well-understood on a global scale, yet no comparable information exists for long non-coding RNA (lncRNA) genes, which were recently recognized to be as numerous as protein-coding genes in mammalian genomes. We performed a genome-wide comparative analysis of the promoters of human lncRNA and protein-coding genes, finding global differences in specific genetic and epigenetic features relevant to transcriptional regulation. These two groups of genes are hence subject to separate transcriptional regulatory programs, including distinct transcription factor (TF) proteins that significantly favor lncRNA, rather than coding-gene, promoters. We report a specific signature of promoter-proximal transcriptional regulation of lncRNA genes, including several distinct transcription factor binding sites (TFBS). Experimental DNase I hypersensitive site profiles are consistent with active configurations of these lncRNA TFBS sets in diverse human cell types. TFBS ChIP-seq datasets confirm the binding events that we predicted using computational approaches for a subset of factors. For several TFs known to be directly regulated by lncRNAs, we find that their putative TFBSs are enriched at lncRNA promoters, suggesting that the TFs and the lncRNAs may participate in a bidirectional feedback loop regulatory network. Accordingly, cells may be able to modulate lncRNA expression levels independently of mRNA levels via distinct regulatory pathways. Our results also raise the possibility that, given the historical reliance on protein-coding gene catalogs to define the chromatin states of active promoters, a revision of these chromatin signature profiles to incorporate expressed lncRNA genes is warranted in the future.

  12. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-15

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and err000.

  13. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-01

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and error

  14. A Nucleus-localized Long Non-Coding RNA Enhances Drought and Salt Stress Tolerance

    KAUST Repository

    Qin, Tao; Zhao, Huayan; Cui, Peng; Albesher, Nour H.; Xiong, Liming

    2017-01-01

    stress. DRIR was expressed at a low level under non-stress conditions but can be significantly activated by drought and salt stress as well as by abscisic acid (ABA) treatment. We identified a T-DNA insertion mutant, drirD, which had higher expression

  15. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

    2014-10-15

    Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

  16. Stress analysis of cylinder to cylinder intersections

    International Nuclear Information System (INIS)

    Revesz, Z.

    1983-01-01

    Cylinder to cylinder intersections have numerous applications in the power industry from different piping junctions to pressure vessel nozzles. A specific purpose computer program has been installed at the author's establishment for finite element analysis of such geometries. Some of the experiences are presented giving a short overview of the analysis of unreinforced man-holes, demonstrating how a more economical design has been verified by analysis. The program installed has linear-elastic and elasto-plastic capabilities. Further, it is prepared for heat transfer analysis with subsequent thermal stress computation. An efficient pre- and post-processor has also been installed and enhanced by the author. The software used is at its present stage capable for problem definition with input data such as outside/ inside diameters, length and number of subdivisions. Similarly simple is the load definition and the graphic representation of the full output. (author)

  17. Analysis of the KUCA MEU experiments using the ANL code system

    Energy Technology Data Exchange (ETDEWEB)

    Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.

    1982-01-01

    This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis.

  18. Analysis of the KUCA MEU experiments using the ANL code system

    International Nuclear Information System (INIS)

    Shiroya, S.; Hayashi, M.; Kanda, K.; Shibata, T.; Woodruff, W.L.; Matos, J.E.

    1982-01-01

    This paper provides some preliminary results on the analysis of the KUCA critical experiments using the ANL code system. Since this system was employed in the earlier neutronics calculations for the KUHFR, it is important to assess its capabilities for the KUHFR. The KUHFR has a unique core configuration which is difficult to model precisely with current diffusion theory codes. This paper also provides some results from a finite-element diffusion code (2D-FEM-KUR), which was developed in a cooperative research program between KURRI and JAERI. This code provides the capability for mockup of a complex core configuration as the KUHFR. Using the same group constants generated by the EPRI-CELL code, the results of the 2D-FEM-KUR code are compared with the finite difference diffusion code (DIF3D(2D) which is mainly employed in this analysis

  19. Finite element formulation for thermal stress analysis of thin reactor structures

    International Nuclear Information System (INIS)

    Kulak, R.F.; Kennedy, J.M.; Belytschko, T.B.

    1978-01-01

    This paper describes the formulation of a finite-element procedure for the thermal stress analysis of thin wall reactor components. A general temperature-dependent constituent relationship is derived from a Gibbs potential function and a temperature-dependent yield surface. This form of constitutive relationship is applicable to problems of small strain. A similar form of a hypoelastic-plastic type is also developed for large strains. The variation of the yield surface with temperature is based upon a temperature-dependent, work-hardening model. The model uses a temperature-equivalent stress-plastic strain diagram which is generated from isothermal unaxial stress-strain data. The above constitutive relationships are incorporated into two computer codes and a previously developed numerical algorithm is used with minor modifications. A set of problems is presented validating the thermal analysis capability of the computer codes to a variety of problem types. (Auth.)

  20. Convolutional Sparse Coding for Static and Dynamic Images Analysis

    Directory of Open Access Journals (Sweden)

    B. A. Knyazev

    2014-01-01

    Full Text Available The objective of this work is to improve performance of static and dynamic objects recognition. For this purpose a new image representation model and a transformation algorithm are proposed. It is examined and illustrated that limitations of previous methods make it difficult to achieve this objective. Static images, specifically handwritten digits of the widely used MNIST dataset, are the primary focus of this work. Nevertheless, preliminary qualitative results of image sequences analysis based on the suggested model are presented.A general analytical form of the Gabor function, often employed to generate filters, is described and discussed. In this research, this description is required for computing parameters of responses returned by our algorithm. The recursive convolution operator is introduced, which allows extracting free shape features of visual objects. The developed parametric representation model is compared with sparse coding based on energy function minimization.In the experimental part of this work, errors of estimating the parameters of responses are determined. Also, parameters statistics and their correlation coefficients for more than 106 responses extracted from the MNIST dataset are calculated. It is demonstrated that these data correspond well with previous research studies on Gabor filters as well as with works on visual cortex primary cells of mammals, in which similar responses were observed. A comparative test of the developed model with three other approaches is conducted; speed and accuracy scores of handwritten digits classification are presented. A support vector machine with a linear or radial basic function is used for classification of images and their representations while principal component analysis is used in some cases to prepare data beforehand. High accuracy is not attained due to the specific difficulties of combining our model with a support vector machine (a 3.99% error rate. However, another method is

  1. Analysis of results of AZTRAN and AZKIND codes for a BWR

    International Nuclear Information System (INIS)

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

    2016-09-01

    This paper presents an analysis of results obtained from simulations performed with the neutron transport code AZTRAN and the kinetic code of neutron diffusion AZKIND, based on comparisons with models corresponding to a typical BWR, in order to verify the behavior and reliability of the values obtained with said code for its current development. For this, simulations of different geometries were made using validated nuclear codes, such as CASMO, MCNP5 and Serpent. The results obtained are considered adequate since they are comparable with those obtained and reported with other codes, based mainly on the neutron multiplication factor and the power distribution of the same. (Author)

  2. Non-Binary Protograph-Based LDPC Codes: Analysis,Enumerators and Designs

    OpenAIRE

    Sun, Yizeng

    2013-01-01

    Non-binary LDPC codes can outperform binary LDPC codes using sum-product algorithm with higher computation complexity. Non-binary LDPC codes based on protographs have the advantage of simple hardware architecture. In the first part of this thesis, we will use EXIT chart analysis to compute the thresholds of different protographs over GF(q). Based on threshold computation, some non-binary protograph-based LDPC codes are designed and their frame error rates are compared with binary LDPC codes. ...

  3. Development of statistical analysis code for meteorological data (W-View)

    International Nuclear Information System (INIS)

    Tachibana, Haruo; Sekita, Tsutomu; Yamaguchi, Takenori

    2003-03-01

    A computer code (W-View: Weather View) was developed to analyze the meteorological data statistically based on 'the guideline of meteorological statistics for the safety analysis of nuclear power reactor' (Nuclear Safety Commission on January 28, 1982; revised on March 29, 2001). The code gives statistical meteorological data to assess the public dose in case of normal operation and severe accident to get the license of nuclear reactor operation. This code was revised from the original code used in a large office computer code to enable a personal computer user to analyze the meteorological data simply and conveniently and to make the statistical data tables and figures of meteorology. (author)

  4. Analysis of the sodium concrete interactions with the NABE code

    International Nuclear Information System (INIS)

    Soule, N.

    1989-01-01

    Experimental studies have been performed in France to investigate sodium-concrete interactions: thermal decomposition of concrete, specific chemical reactions, experimentation in liquid and vapour phase, sodium-concrete interaction without liner protection. Simultaneously computer codes have been developed in order to study the response of the containment building of a liquid metal fast breeder reactor to a sodium pool fire worsened by a sodium-concrete interaction: the NABE code. This code takes into account: a) sodium combustion; b) thermal decomposition of concrete with associated chemical reactions: (liquid sodium-vapour water reaction, liquid sodium-carbon dioxide reaction, liquid sodium-solid compounds of concrete, hydrogen combustion); c) chemical reactions in vapour phase; d) decay heat; e) gas aerosol inlets/outlets; f) aerosol behaviour (sedimentation, diffusion, leak); g) thermal exchanges. An example of a situation, typical of assessment of beyond design basis situations in LMFBR, is given. (author)

  5. Fusion PIC code performance analysis on the Cori KNL system

    Energy Technology Data Exchange (ETDEWEB)

    Koskela, Tuomas S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Deslippe, Jack [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Friesen, Brian [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). National Energy Research Scientific Computing Center (NERSC); Raman, Karthic [INTEL Corp. (United States)

    2017-05-25

    We study the attainable performance of Particle-In-Cell codes on the Cori KNL system by analyzing a miniature particle push application based on the fusion PIC code XGC1. We start from the most basic building blocks of a PIC code and build up the complexity to identify the kernels that cost the most in performance and focus optimization efforts there. Particle push kernels operate at high AI and are not likely to be memory bandwidth or even cache bandwidth bound on KNL. Therefore, we see only minor benefits from the high bandwidth memory available on KNL, and achieving good vectorization is shown to be the most beneficial optimization path with theoretical yield of up to 8x speedup on KNL. In practice we are able to obtain up to a 4x gain from vectorization due to limitations set by the data layout and memory latency.

  6. Design and Analysis of LT Codes with Decreasing Ripple Size

    DEFF Research Database (Denmark)

    Sørensen, Jesper Hemming; Popovski, Petar; Østergaard, Jan

    2012-01-01

    In this paper we propose a new design of LT codes, which decreases the amount of necessary overhead in comparison to existing designs. The design focuses on a parameter of the LT decoding process called the ripple size. This parameter was also a key element in the design proposed in the original...... work by Luby. Specifically, Luby argued that an LT code should provide a constant ripple size during decoding. In this work we show that the ripple size should decrease during decoding, in order to reduce the necessary overhead. Initially we motivate this claim by analytical results related...... to the redundancy within an LT code. We then propose a new design procedure, which can provide any desired achievable decreasing ripple size. The new design procedure is evaluated and compared to the current state of the art through simulations. This reveals a significant increase in performance with respect...

  7. Code Blue Emergencies: A Team Task Analysis and Educational Initiative

    Directory of Open Access Journals (Sweden)

    James W. Price

    2012-04-01

    Full Text Available Introduction: The objective of this study was to identify factors that have a positive or negative influence on resuscitation team performance during emergencies in the operating room (OR and post-operative recovery unit (PAR at a major Canadian teaching hospital. This information was then used to implement a team training program for code blue emergencies. Methods: In 2009/10, all OR and PAR nurses and 19 anesthesiologists at Vancouver General Hospital (VGH were invited to complete an anonymous, 10 minute written questionnaire regarding their code blue experience. Survey questions were devised by 10 recovery room and operation room nurses as well as 5 anesthesiologists representing 4 different hospitals in British Columbia. Three iterations of the survey were reviewed by a pilot group of nurses and anesthesiologists and their feedback was integrated into the final version of the survey. Results: Both nursing staff (n = 49 and anesthesiologists (n = 19 supported code blue training and believed that team training would improve patient outcome. Nurses noted that it was often difficult to identify the leader of the resuscitation team. Both nursing staff and anesthesiologists strongly agreed that too many people attending the code blue with no assigned role hindered team performance. Conclusion: Identifiable leadership and clear communication of roles were identified as keys to resuscitation team functioning. Decreasing the number of people attending code blue emergencies with no specific role, increased access to mock code blue training, and debriefing after crises were all identified as areas requiring improvement. Initial team training exercises have been well received by staff.

  8. Review of provisions on corrosion fatigue and stress corrosion in WWER and Western LWR Codes and Standards

    International Nuclear Information System (INIS)

    Buckthorpe, D.; Filatov, V.; Tashkinov, A.; Evropin, S.V.; Matocha, K.; Guinovart, J.

    2003-01-01

    Results are presented from a collaborative project performed on behalf of the European Commission, Working Group Codes and Standards. The work covered the contents of current codes and standards, plant experience and R and D results. Current fatigue design rules use S-N curves based on tests in air. Although WWER and LWR design curves are often similar they are derived, presented and used in different ways and it is neither convenient nor appropriate to harmonise them. Similarly the fatigue crack growth laws used in the various design and in-service inspection rules differ significantly with respect to both growth rates in air and the effects of water reactor environments. Harmonised approaches to the effects of WWER and LWR environments are possible based on results from R and D programmes carried out over the last decade. For carbon and low alloy steels a consistent approach to both crack initiation and growth can be formulated based on the superposition of environmentally assisted cracking effects on the fatigue crack development. The approach indicates that effects of the water environment are minimal given appropriate control of the oxygen content of the water and/or the sulphur content of the steel. For austenitic stainless steels a different mechanisms may apply and a harmonised approach is possible at present only for S-N curves. Although substantial progress has been made with respect to corrosion fatigue, more data and a clearer understanding are required in order to write code provisions particularly in the area of high cycle fatigue. Reactor operation experience shows stress corrosion cracking of austenitic steels is the most common cause of failure. These failures are associated with high residual stresses combined with high levels of dissolved oxygen or the presence of contaminants. For primary circuit internals there is a potential threat to integrity from irradiated assisted stress corrosion cracking. Design and in-service inspection rules do not at

  9. OSSMETER D3.4 – Language-Specific Source Code Quality Analysis

    NARCIS (Netherlands)

    J.J. Vinju (Jurgen); A. Shahi (Ashim); H.J.S. Basten (Bas)

    2014-01-01

    htmlabstractThis deliverable is part of WP3: Source Code Quality and Activity Analysis. It provides descriptions and prototypes of the tools that are needed for source code quality analysis in open source software projects. It builds upon the results of: • Deliverable 3.1 where infra-structure and

  10. Improvements to the COBRA-TF (EPRI) computer code for steam generator analysis. Final report

    International Nuclear Information System (INIS)

    Stewart, C.W.; Barnhart, J.S.; Koontz, A.S.

    1980-09-01

    The COBRA-TF (EPRI) code has been improved and extended for pressurized water reactor steam generator analysis. New features and models have been added in the areas of subcooled boiling and heat transfer, turbulence, numerics, and global steam generator modeling. The code's new capabilities are qualified against selected experimental data and demonstrated for typical global and microscale steam generator analysis

  11. Modular Modeling System (MMS) code: a versatile power plant analysis package

    International Nuclear Information System (INIS)

    Divakaruni, S.M.; Wong, F.K.L.

    1987-01-01

    The basic version of the Modular Modeling System (MMS-01), a power plant systems analysis computer code jointly developed by the Nuclear Power and the Coal Combustion Systems Divisions of the Electric Power Research Institute (EPRI), has been released to the utility power industry in April 1983 at a code release workshop held in Charlotte, North Carolina. Since then, additional modules have been developed to analyze the Pressurized Water Reactors (PWRs) and the Boiling Water Reactors (BWRs) when the safety systems are activated. Also, a selected number of modules in the MMS-01 library have been modified to allow the code users more flexibility in constructing plant specific systems for analysis. These new PWR and BWR modules constitute the new MMS library, and it includes the modifications to the MMS-01 library. A year and half long extensive code qualification program of this new version of the MMS code at EPRI and the contractor sites, back by further code testing in an user group environment is culminating in the MMS-02 code release announcement seminar. At this seminar, the results of user group efforts and the code qualification program will be presented in a series of technical sessions. A total of forty-nine papers will be presented to describe the new code features and the code qualification efforts. For the sake of completion, an overview of the code is presented to include the history of the code development, description of the MMS code and its structure, utility engineers involvement in MMS-01 and MMS-02 validations, the enhancements made in the last 18 months to the code, and finally the perspective on the code future in the fossil and nuclear industry

  12. General data analysis code for TDCR liquid scintillation counting

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, D. [Laboratorio de Metrologia de Radioisotopos, Comision Nacional de Energia Atomica, Buenos Aires (Argentina)], E-mail: drodrigu@cae.cnea.gov.ar; Arenillas, P.; Capoulat, M.E.; Balpardo, C. [Laboratorio de Metrologia de Radioisotopos, Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2008-06-15

    A non-radionuclide-specific computer code to analyze data, calculate detection efficiency and activity in a TDCR system is presented. The program was developed prioritizing flexibility in measuring conditions, parameters and calculation models. It is also intended to be well structured in order to easily replace subroutines which could eventually be improved by the user. It is written in standard FORTRAN language but a graphical interface is also available. Several tests were performed to check the ability of the code to deal with different decay schemes such as H-3, C-14, Fe-55, Mn-54 and Co-60.

  13. General data analysis code for TDCR liquid scintillation counting

    International Nuclear Information System (INIS)

    Rodrigues, D.; Arenillas, P.; Capoulat, M.E.; Balpardo, C.

    2008-01-01

    A non-radionuclide-specific computer code to analyze data, calculate detection efficiency and activity in a TDCR system is presented. The program was developed prioritizing flexibility in measuring conditions, parameters and calculation models. It is also intended to be well structured in order to easily replace subroutines which could eventually be improved by the user. It is written in standard FORTRAN language but a graphical interface is also available. Several tests were performed to check the ability of the code to deal with different decay schemes such as H-3, C-14, Fe-55, Mn-54 and Co-60

  14. ANALYSIS OF FOREIGN EXPERIENCE OF INTRODUCTION OF THE SCHOOL DRESS CODE

    Directory of Open Access Journals (Sweden)

    O, I, Denisova

    2016-01-01

    Full Text Available The aim of this article is to analyze the international discourse on policy issues of school dress code; to identify problems relevant to the national system of education, and proposed ways of their solution.Methods. The methods involve summarizing of the content of the leading internationalsources regarding issues regarding school education and health; a systematic analysis of sociological researchResults. The polemic aspects to the requirements for the appearance of the students are considered. These requirements affect many important aspects and the principles of functioning of an educational system and all society in general: the observance of the constitutional rights of a child, including freedom of expression on grounds of gender; tolerance policy of intra school and state policy in the light of migration processes; problems of leveling of social and economic stratification of children, etc. The advantages and disadvantages of existing approaches to the development requirements in the countries of Europe, Asia, Africa, and North America are analysed.Scientific novelty. The influence of policy of a school dress-code on children’s behavior and health, relationships in multinational school groups, identification and safety of school students is proved. The ambiguity of information expressiveness and  ignificance of garments of clothing and other elements of appearance of modern school students is disclosed: symbols of belonging to a religious and/or national culture, a well-known fashion brand, a modern informal/political trend, as well as the severity of gender differences in the design of school uniforms. The correlation between the wording of the requirements of the dress code and the effectiveness of its implementation from the point of view of minimization of internal stresses in the school team is defined.Practical significance. Based on the systemized proposals for quality improvement of the socio-cultural environment of

  15. Stress analysis of the HFIR HB-2 and HB-3 beam tube nozzles

    International Nuclear Information System (INIS)

    Williams, P.T.

    1998-08-01

    The results of three-dimensional linear elastic stress analyses of the HFIR HB-2 and HB-3 nozzles are presented in this report. Finite element models were developed using the PATRAN pre-processing code and translated into ABAQUS input file format. A scoping analysis using simple geometries with internal pressure loading was carried out to assess the capabilities of the ABAQUS/Standard code to calculate maximum principal stress distributions within cylinders with and without holes. These scoping calculations were also used to provide estimates for the variation in tangential stress around the rim of a nozzle using the superposition of published closed-form solutions for the stress around a hole in an infinite flat plate under uniaxial tension. From the results of the detailed finite element models, peak stress concentration factors (based on the maximum principal stresses in tension) were calculated to be 3.0 for the HB-2 nozzle and 2.8 for the HB-3 nozzle. Submodels for each nozzle were built to calculate the maximum principal stress distribution in the weldment region around the nozzle, where displacement boundary conditions for the submodels were automatically calculated by ABAQUS using the results of the global nozzle models. Maximum principal stresses are plotted and tabulated for eight positions around each nozzle and nozzle weldment

  16. Code Pulse: Software Assurance (SWA) Visual Analytics for Dynamic Analysis of Code

    Science.gov (United States)

    2014-09-01

    31 4.5.1 Market Analysis...competitive market analysis to assess the tool potential. The final transition targets were selected and expressed along with our research on the topic...public release milestones. Details of our testing methodology is in our Software Test Plan deliv- erable, CP- STP -0001. A summary of this approach is

  17. ESE a 2D compressible multiphase flow code developed for MFCI analysis - code validation

    International Nuclear Information System (INIS)

    Leskovar, M.; Mavko, B.

    1998-01-01

    ESE (Evaluation of Steam Explosions) is a general second order accurate two-dimensional compressible multiphase flow computer code. It has been developed to model the interaction of molten core debris with water during the first premixing stage of a steam explosion. A steam explosion is a physical event, which may occur during a severe reactor accident following core meltdown when the molten fuel comes into contact with the coolant water. Since the exchanges of mass, momentum and energy are regime dependent, different exchange laws have been incorporated in ESE for the major flow regimes. With ESE a number of premixing experiments performed at the Oxford University and at the QUEOS facility at Forschungszentrum Karlsruhe has been simulated. In these premixing experiments different jets of spheres were injected in a water poll. The ESE validation plan was carefully chosen, starting from very simple, well-defined problems, and gradually working up to more complicated ones. The results of ESE simulations, which were compared to experimental data and also to first order accurate calculations, are presented in form graphs. Most of the ESE results agree qualitatively as quantitatively reasonably well with experimental data and in general better than the results obtained with the first order accurate calculation.(author)

  18. An Analysis of Countries which have Integrated Coding into their Curricula and the Content Analysis of Academic Studies on Coding Training in Turkey

    Directory of Open Access Journals (Sweden)

    Hüseyin Uzunboylu

    2017-11-01

    Full Text Available The first aim is to conduct a general analysis of countries which have integrated coding training into their curricula, and the second aim is to conduct a content analysis of studies on coding training in Turkey. It was identified that there are only a few academic studies on coding training in Turkey, and that the majority of them were published in 2016, the intended population was mainly “undergraduate students” and that the majority of these students were Computer Education and Instructional Technology undergraduates. It was determined that the studies mainly focused on the subjects of “programming” and “Scratch”, the terms programming and coding were used as synonyms, most of the studies were carried out using quantitative methods and data was obtained mostly by literature review and scale/survey interval techniques.

  19. Evaluation of the analysis models in the ASTRA nuclear design code system

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-11-15

    In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.

  20. Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

    1997-01-01

    A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)

  1. From text to codings: intercoder reliability assessment in qualitative content analysis.

    Science.gov (United States)

    Burla, Laila; Knierim, Birte; Barth, Jurgen; Liewald, Katharina; Duetz, Margreet; Abel, Thomas

    2008-01-01

    High intercoder reliability (ICR) is required in qualitative content analysis for assuring quality when more than one coder is involved in data analysis. The literature is short of standardized procedures for ICR procedures in qualitative content analysis. To illustrate how ICR assessment can be used to improve codings in qualitative content analysis. Key steps of the procedure are presented, drawing on data from a qualitative study on patients' perspectives on low back pain. First, a coding scheme was developed using a comprehensive inductive and deductive approach. Second, 10 transcripts were coded independently by two researchers, and ICR was calculated. A resulting kappa value of .67 can be regarded as satisfactory to solid. Moreover, varying agreement rates helped to identify problems in the coding scheme. Low agreement rates, for instance, indicated that respective codes were defined too broadly and would need clarification. In a third step, the results of the analysis were used to improve the coding scheme, leading to consistent and high-quality results. The quantitative approach of ICR assessment is a viable instrument for quality assurance in qualitative content analysis. Kappa values and close inspection of agreement rates help to estimate and increase quality of codings. This approach facilitates good practice in coding and enhances credibility of analysis, especially when large samples are interviewed, different coders are involved, and quantitative results are presented.

  2. LDPC Codes--Structural Analysis and Decoding Techniques

    Science.gov (United States)

    Zhang, Xiaojie

    2012-01-01

    Low-density parity-check (LDPC) codes have been the focus of much research over the past decade thanks to their near Shannon limit performance and to their efficient message-passing (MP) decoding algorithms. However, the error floor phenomenon observed in MP decoding, which manifests itself as an abrupt change in the slope of the error-rate curve,…

  3. 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

  4. Recent developments in seismic analysis in the code Aster

    International Nuclear Information System (INIS)

    Guihot, P.; Devesa, G.; Dumond, A.; Panet, M.; Waeckel, F.

    1996-01-01

    Progress in the field of seismic qualification and design methods made these last few years allows physical phenomena actually in play to be better considered, while cutting down the conservatism associated with some simplified design methods. So following the change in methods and developing the most advantageous ones among them contributes to the process of the seismic margins assessment and the preparation of new design tools for future series. In this paper, the main developments and improvements in methods which have been made these last two years in the Code Aster, in order to improve seismic calculation methods and seismic margin assessment are presented. The first development relates to making the MISS3D soil structure interaction code available, thanks to an interface made with the Code Aster. The second relates to the possibility of making modal basis time calculations on multi-supported structures by considering local non linearities like impact, friction or squeeze fluid forces. Recent developments in random dynamics and postprocessing devoted to earthquake designs are then mentioned. Three applications of these developments are then ut forward. The first application relates to a test case for soil structure interaction design using MISS3D-Aster coupling. The second is a test case for a multi-supported structure. The last application, more for manufacturing, refers to seismic qualification of Main Live Steam stop valves. First results of the independent validation of the Code Aster seismic design functionalities, which provide and improve the quality of software, are also recalled. (authors)

  5. 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

  6. Analysis of an XADS Target with the System Code TRACE

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Sanchez Espinoza, Victor H.; Feng, Bo

    2008-01-01

    Accelerator-driven systems (ADS) present an option to reduce the radioactive waste of the nuclear industry. The experimental Accelerator-Driven System (XADS) has been designed to investigate the feasibility of using ADS on an industrial scale to burn minor actinides. The target section lies in the middle of the subcritical core and is bombarded by a proton beam to produce spallation neutrons. The thermal energy produced from this reaction requires a heat removal system for the target section. The target is cooled by liquid lead-bismuth-eutectics (LBE) in the primary system which in turn transfers the heat via a heat exchanger (HX) to the secondary coolant, Diphyl THT (DTHT), a synthetic diathermic fluid. Since this design is still in development, a detailed investigation of the system is necessary to evaluate the behavior during normal and transient operations. Due to the lack of experimental facilities and data for ADS, the analyses are mostly done using thermal hydraulic codes. In addition to evaluating the thermal hydraulics of the XADS, this paper also benchmarks a new code developed by the NRC, TRACE, against other established codes. The events used in this study are beam power switch-on/off transients and a loss of heat sink accident. The obtained results from TRACE were in good agreement with the results of various other codes. (authors)

  7. Further Analysis of Motorcycle Helmet Effectiveness Using CODES Linked Data

    Science.gov (United States)

    1998-01-01

    Linked data from the Crash Outcome Data Evaluation System (CODES) in seven : states was used by the National Highway Traffic Safety Administration as the : basis of a 1996 Report to Congress on the Benefits of Safety Belts and : Motorcycle Helmets (D...

  8. Analysis of the OPERA-15 two-dimensional voiding experiment using the SAS4A code

    International Nuclear Information System (INIS)

    Briggs, L.L.

    1984-01-01

    Overall, SAS4A appears to do a good job for simulating the OPERA-15 experiment. For most of the experiment parameters, the code calculations compare quite well with the experimental data. The lack of a multi-dimensional voiding model has the effect of extending the flow coastdown time until voiding starts; otherwise, the code simulates the accident progression satisfactorily. These results indicate a need for further work in this area in the form of a tandem analysis by a two-dimensional flow code and a one-dimensional version of that code to confirm the observations derived from the SAS4A analysis

  9. Development and application of sub-channel analysis code based on SCWR core

    International Nuclear Information System (INIS)

    Fu Shengwei; Xu Zhihong; Yang Yanhua

    2011-01-01

    The sub-channel analysis code SABER was developed for thermal-hydraulic analysis of supercritical water-cooled reactor (SCWR) fuel assembly. The extended computational cell structure, a new boundary conditions, 3 dimensional heat conduction model and water properties package were implemented in SABER code, which could be used to simulate the thermal fuel assembly of SCWR. To evaluate the applicability of the code, a steady state calculation of the fuel assembly was performed. The results indicate good applicability of the SABER code to simulate the counter-current flow and the heat exchange between coolant and moderator channels. (authors)

  10. Analysis of ATLAS FLB-EC6 Experiment using SPACE Code

    International Nuclear Information System (INIS)

    Lee, Donghyuk; Kim, Yohan; Kim, Seyun

    2013-01-01

    The new code is named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). As a part of code validation effort, simulation of ATLAS FLB(Feedwater Line Break) experiment using SPACE code has been performed. The FLB-EC6 experiment is economizer break of a main feedwater line. The calculated results using the SPACE code are compared with those from the experiment. The ATLAS FLB-EC6 experiment, which is economizer feedwater line break, was simulated using the SPACE code. The calculated results were compared with those from the experiment. The comparisons of break flow rate and steam generator water level show good agreement with the experiment. The SPACE code is capable of predicting physical phenomena occurring during ATLAS FLB-EC6 experiment

  11. Construction and performance analysis of variable-weight optical orthogonal codes for asynchronous OCDMA systems

    Science.gov (United States)

    Li, Chuan-qi; Yang, Meng-jie; Zhang, Xiu-rong; Chen, Mei-juan; He, Dong-dong; Fan, Qing-bin

    2014-07-01

    A construction scheme of variable-weight optical orthogonal codes (VW-OOCs) for asynchronous optical code division multiple access (OCDMA) system is proposed. According to the actual situation, the code family can be obtained by programming in Matlab with the given code weight and corresponding capacity. The formula of bit error rate (BER) is derived by taking account of the effects of shot noise, avalanche photodiode (APD) bulk, thermal noise and surface leakage currents. The OCDMA system with the VW-OOCs is designed and improved. The study shows that the VW-OOCs have excellent performance of BER. Despite of coming from the same code family or not, the codes with larger weight have lower BER compared with the other codes in the same conditions. By taking simulation, the conclusion is consistent with the analysis of BER in theory. And the ideal eye diagrams are obtained by the optical hard limiter.

  12. Users' manual for fault tree analysis code: CUT-TD

    International Nuclear Information System (INIS)

    Watanabe, Norio; Kiyota, Mikio.

    1992-06-01

    The CUT-TD code has been developed to find minimal cut sets for a given fault tree and to calculate the occurrence probability of its top event. This code uses an improved top-down algorithm which can enhance the efficiency in deriving minimal cut sets. The features in processing techniques incorporated into CUT-TD are as follows: (1) Consecutive OR gates or consecutive AND gates can be coalesced into a single gate. As a result, this processing directly produces cut sets for the redefined single gate with each gate not being developed. (2) The independent subtrees are automatically identified and their respective cut sets are separately found to enhance the efficiency in processing. (3) The minimal cut sets can be obtained for the top event of a fault tree by combining their respective minimal cut sets for several gates of the fault tree. (4) The user can reduce the computing time for finding minimal cut sets and control the size and significance of cut sets by inputting a minimum probability cut off and/or a maximum order cut off. (5) The user can select events that need not to be further developed in the process of obtaining minimal cut sets. This option can reduce the number of minimal cut sets, save the computing time and assists the user in reviewing the result. (6) Computing time is monitored by the CUT-TD code so that it can prevent the running job from abnormally ending due to excessive CPU time and produce an intermediate result. The CUT-TD code has the ability to restart the calculation with use of the intermediate result. This report provides a users' manual for the CUT-TD code. (author)

  13. Energy Savings Analysis of the Proposed NYStretch-Energy Code 2018

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Bing [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhang, Jian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chen, Yan [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edelson, Jim [New Buildings Inst. (NBI), Portland, OR (United States); Lyles, Mark [New Buildings Inst. (NBI), Portland, OR (United States)

    2018-01-20

    This study was conducted by the Pacific Northwest National Laboratory (PNNL) in support of the stretch energy code development led by the New York State Energy Research and Development Authority (NYSERDA). In 2017 NYSERDA developed its 2016 Stretch Code Supplement to the 2016 New York State Energy Conservation Construction Code (hereinafter referred to as “NYStretch-Energy”). NYStretch-Energy is intended as a model energy code for statewide voluntary adoption that anticipates other code advancements culminating in the goal of a statewide Net Zero Energy Code by 2028. Since then, NYSERDA continues to develop the NYStretch-Energy Code 2018 edition. To support the effort, PNNL conducted energy simulation analysis to quantify the energy savings of proposed commercial provisions of the NYStretch-Energy Code (2018) in New York. The focus of this project is the 20% improvement over existing commercial model energy codes. A key requirement of the proposed stretch code is that it be ‘adoptable’ as an energy code, meaning that it must align with current code scope and limitations, and primarily impact building components that are currently regulated by local building departments. It is largely limited to prescriptive measures, which are what most building departments and design projects are most familiar with. This report describes a set of energy-efficiency measures (EEMs) that demonstrate 20% energy savings over ANSI/ASHRAE/IES Standard 90.1-2013 (ASHRAE 2013) across a broad range of commercial building types and all three climate zones in New York. In collaboration with New Building Institute, the EEMs were developed from national model codes and standards, high-performance building codes and standards, regional energy codes, and measures being proposed as part of the on-going code development process. PNNL analyzed these measures using whole building energy models for selected prototype commercial buildings and multifamily buildings representing buildings in New

  14. Electronic manual of the nuclear characteristics analysis code-set for FBR

    International Nuclear Information System (INIS)

    Makino, Tohru

    2001-03-01

    Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows; (1) The electronic manual includes explanations of following codes: JOINT : Code Interface Program. SLAROM, CASUP : Effective Cross Section Calculation Code. CITATION-FBR : Diffusion Analysis Code. PERKY : Perturbative Diffusion Analysis Code. SNPERT, SNPERT-3D : Perturbative Transport Analysis Code. SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code. NSHEX : Transport Analysis Code using Nodal Method. ABLE : Cross Section Adjustment Calculation Code. ACCEPT : Predicting Accuracy Evaluation Code. (2) The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3) Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4) User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5) Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other advantages of use

  15. Experiments and analysis of thermal stresses around the nozzle of the reactor vessel

    International Nuclear Information System (INIS)

    Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.

    1981-01-01

    This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively

  16. Stress Analysis for Mobile Hot Cell Design

    International Nuclear Information System (INIS)

    Muhammad Hannan Bahrin; Anwar Abdul Rahman; Mohd Arif Hamzah

    2015-01-01

    Prototype and Plant Development Centre (PDC) is developing a Mobile Hot Cell (MHC) to handle and manage Spent High Activity Radioactive Sources (SHARS), such as teletherapy heads and dry irradiators. At present, there are two units of MHC in the world, one in South Africa and the other one in China. Malaysian Mobile MHC is developed by Malaysian Nuclear Agency with the assistance of IAEA expert, based on the design of South Africa and China, but with improved features. Stress analysis has been performed on the design to fulfill the safety requirement in MHC operation. This paper discusses the loading effect analysis from the radiation shielding materials to the MHC wall structure, roof supporting column and window structure. (author)

  17. Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Jiang, Yang; Yang, Jue; Zhang, Bo

    2013-01-01

    Highlights: ► A new safety analysis code named SCTRAN is developed for SCWRs. ► Capability of SCTRAN is verified by comparing with code APROS and RELAP5-3D. ► A new passive safety system is proposed for CGNPC SCWR and analyzed with SCTRAN. ► CGNPC SCWR is able to cope with two critical accidents for SCWRs, LOFA and LOCA. - Abstract: Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety

  18. Development of an advanced code system for fast-reactor transient analysis

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

    FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

  19. A 3D moisture-stress FEM analysis for time dependent problems in timber structures

    Science.gov (United States)

    Fortino, Stefania; Mirianon, Florian; Toratti, Tomi

    2009-11-01

    This paper presents a 3D moisture-stress numerical analysis for timber structures under variable humidity and load conditions. An orthotropic viscoelastic-mechanosorptive material model is specialized on the basis of previous models. Both the constitutive model and the equations needed to describe the moisture flow across the structure are implemented into user subroutines of the Abaqus finite element code and a coupled moisture-stress analysis is performed for several types of mechanical loads and moisture changes. The presented computational approach is validated by analyzing some wood tests described in the literature and comparing the computational results with the reported experimental data.

  20. Simplified diagnostic coding sheet for computerized data storage and analysis in ophthalmology.

    Science.gov (United States)

    Tauber, J; Lahav, M

    1987-11-01

    A review of currently-available diagnostic coding systems revealed that most are either too abbreviated or too detailed. We have compiled a simplified diagnostic coding sheet based on the International Coding and Diagnosis (ICD-9), which is both complete and easy to use in a general practice. The information is transferred to a computer, which uses the relevant (ICD-9) diagnoses as database and can be retrieved later for display of patients' problems or analysis of clinical data.

  1. 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)

  2. Statistical mechanics analysis of LDPC coding in MIMO Gaussian channels

    Energy Technology Data Exchange (ETDEWEB)

    Alamino, Roberto C; Saad, David [Neural Computing Research Group, Aston University, Birmingham B4 7ET (United Kingdom)

    2007-10-12

    Using analytical methods of statistical mechanics, we analyse the typical behaviour of a multiple-input multiple-output (MIMO) Gaussian channel with binary inputs under low-density parity-check (LDPC) network coding and joint decoding. The saddle point equations for the replica symmetric solution are found in particular realizations of this channel, including a small and large number of transmitters and receivers. In particular, we examine the cases of a single transmitter, a single receiver and symmetric and asymmetric interference. Both dynamical and thermodynamical transitions from the ferromagnetic solution of perfect decoding to a non-ferromagnetic solution are identified for the cases considered, marking the practical and theoretical limits of the system under the current coding scheme. Numerical results are provided, showing the typical level of improvement/deterioration achieved with respect to the single transmitter/receiver result, for the various cases.

  3. Statistical mechanics analysis of LDPC coding in MIMO Gaussian channels

    International Nuclear Information System (INIS)

    Alamino, Roberto C; Saad, David

    2007-01-01

    Using analytical methods of statistical mechanics, we analyse the typical behaviour of a multiple-input multiple-output (MIMO) Gaussian channel with binary inputs under low-density parity-check (LDPC) network coding and joint decoding. The saddle point equations for the replica symmetric solution are found in particular realizations of this channel, including a small and large number of transmitters and receivers. In particular, we examine the cases of a single transmitter, a single receiver and symmetric and asymmetric interference. Both dynamical and thermodynamical transitions from the ferromagnetic solution of perfect decoding to a non-ferromagnetic solution are identified for the cases considered, marking the practical and theoretical limits of the system under the current coding scheme. Numerical results are provided, showing the typical level of improvement/deterioration achieved with respect to the single transmitter/receiver result, for the various cases

  4. Academic freedom, analysis, and the Code of Professional Conduct.

    Science.gov (United States)

    Snelling, Paul C; Lipscomb, Martin

    2004-11-01

    Despite nursing's move into higher education, academic freedom has received little attention within the literature. After discussing the concept of academic freedom, this paper argues that there is a potential tension between academic freedom and the requirement to educate student nurses who are fit for practice. One way in which this tension might be revealed is in the marking of student assignments. We ask the question--how should nurse educators mark an essay which is sufficiently analytical but reaches moral conclusions that lie outside the Code of Professional Conduct? We argue that despite an understandable temptation to penalise such an essay, invoking the Code of Professional Conduct to do so, no penalty should be applied, and academic freedom for students within higher education should be encouraged. This is because first, academic freedom is a good in itself especially as it allows unconventional and unpalatable conclusions to be discussed and rebutted, and second, applying a penalty on these grounds is necessarily inconsistent.

  5. 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)

  6. Numerical analysis of alpha spectra using two different codes

    International Nuclear Information System (INIS)

    Hurtado, S.; Jimenez-Ramos, M.C.; Villa, M.; Vioque, I.; Manjon, G.; Garcia-Tenorio, R.

    2008-01-01

    This work presents an intercomparison between commercial software for alpha-particle spectrometry, Genie 2000, and the new free available software, Winalpha, developed by International Atomic Energy Agency (IAEA). In order to compare both codes, different environmental spectra containing plutonium, uranium, thorium and polonium have been analyzed, together with IAEA test alpha spectra. A statistical study was performed in order to evaluate the precision and accuracy in the analyses, and to enhance the confidence in using the software on alpha spectrometric studies

  7. Approximation generation for correlations in thermal-hydraulic analysis codes

    International Nuclear Information System (INIS)

    Pereira, Luiz C.M.; Carmo, Eduardo G.D. do

    1997-01-01

    A fast and precise evaluation of fluid thermodynamic and transport properties is needed for the efficient mass, energy and momentum transport phenomena simulation related to nuclear plant power generation. A fully automatic code capable to generate suitable approximation for correlations with one or two independent variables is presented. Comparison in terms of access speed and precision with original correlations currently used shows the adequacy of the approximation obtained. (author). 4 refs., 8 figs., 1 tab

  8. Meanline Analysis of Turbines with Choked Flow in the Object-Oriented Turbomachinery Analysis Code

    Science.gov (United States)

    Hendricks, Eric S.

    2016-01-01

    The prediction of turbomachinery performance characteristics is an important part of the conceptual aircraft engine design process. During this phase, the designer must examine the effects of a large number of turbomachinery design parameters to determine their impact on overall engine performance and weight. The lack of detailed design information available in this phase necessitates the use of simpler meanline and streamline methods to determine the turbomachinery geometry characteristics and provide performance estimates prior to more detailed CFD (Computational Fluid Dynamics) analyses. While a number of analysis codes have been developed for this purpose, most are written in outdated software languages and may be difficult or impossible to apply to new, unconventional designs. The Object-Oriented Turbomachinery Analysis Code (OTAC) is currently being developed at NASA Glenn Research Center to provide a flexible meanline and streamline analysis capability in a modern object-oriented language. During the development and validation of OTAC, a limitation was identified in the code's ability to analyze and converge turbines as the flow approached choking. This paper describes a series of changes which can be made to typical OTAC turbine meanline models to enable the assessment of choked flow up to limit load conditions. Results produced with this revised model setup are provided in the form of turbine performance maps and are compared to published maps.

  9. Performance analysis of LDPC codes on OOK terahertz wireless channels

    Science.gov (United States)

    Chun, Liu; Chang, Wang; Jun-Cheng, Cao

    2016-02-01

    Atmospheric absorption, scattering, and scintillation are the major causes to deteriorate the transmission quality of terahertz (THz) wireless communications. An error control coding scheme based on low density parity check (LDPC) codes with soft decision decoding algorithm is proposed to improve the bit-error-rate (BER) performance of an on-off keying (OOK) modulated THz signal through atmospheric channel. The THz wave propagation characteristics and channel model in atmosphere is set up. Numerical simulations validate the great performance of LDPC codes against the atmospheric fading and demonstrate the huge potential in future ultra-high speed beyond Gbps THz communications. Project supported by the National Key Basic Research Program of China (Grant No. 2014CB339803), the National High Technology Research and Development Program of China (Grant No. 2011AA010205), the National Natural Science Foundation of China (Grant Nos. 61131006, 61321492, and 61204135), the Major National Development Project of Scientific Instrument and Equipment (Grant No. 2011YQ150021), the National Science and Technology Major Project (Grant No. 2011ZX02707), the International Collaboration and Innovation Program on High Mobility Materials Engineering of the Chinese Academy of Sciences, and the Shanghai Municipal Commission of Science and Technology (Grant No. 14530711300).

  10. Demonstration study on shielding safety analysis code. 7

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  11. Study of experimental validation for combustion analysis of GOTHIC code

    International Nuclear Information System (INIS)

    Lee, J. Y.; Yang, S. Y.; Park, K. C.; Jeong, S. H.

    2001-01-01

    In this study, present lumped and subdivided GOTHIC6 code analyses of the premixed hydrogen combustion experiment at the Seoul National University and comparison with the experiment results. The experimental facility has 16367 cc free volume and rectangular shape. And the test was performed with unit equivalence ratio of the hydrogen and air, and with various location of igniter position. Using the lumped and mechanistic combustion model in GOTHIC6 code, the experiments were simulated with the same conditions. In the comparison between experiment and calculated results, the GOTHIC6 prediction of the combustion response does not compare well with the experiment results. In the point of combustion time, the lumped combustion model of GOTHIC6 code does not simulate the physical phenomena of combustion appropriately. In the case of mechanistic combustion model, the combustion time is predicted well, but the induction time of calculation data is longer than the experiment data remarkably. Also, the laminar combustion model of GOTHIC6 has deficiency to simulate combustion phenomena unless control the user defined value appropriately. And the pressure is not a proper variable that characterize the three dimensional effect of combustion

  12. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  13. Improvement on reaction model for sodium-water reaction jet code and application analysis

    International Nuclear Information System (INIS)

    Itooka, Satoshi; Saito, Yoshinori; Okabe, Ayao; Fujimata, Kazuhiro; Murata, Shuuichi

    2000-03-01

    In selecting the reasonable DBL on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.30) and application analysis to the water injection tests for confirmation of code propriety were performed. On the improvement of the code, a gas-liquid interface area density model was introduced to develop a chemical reaction model with a little dependence on calculation mesh size. The test calculation using the improved code (LEAP-JET ver.1.40) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results and the influence to analysis result of a model are reasonable. For the application analysis to the water injection tests, water injection behavior and SWR jet behavior analyses on the new SWAT-1 (SWAT-1R) and SWAT-3 (SWAT-3R) tests were performed using the LEAP-BLOW code and the LEAP-JET code. In the application analysis of the LEAP-BLOW code, parameter survey study was performed. As the results, the condition of the injection nozzle diameter needed to simulate the water leak rate was confirmed. In the application analysis of the LEAP-JET code, temperature behavior of the SWR jet was investigated. (author)

  14. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.

    2006-01-01

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

  15. Structural evaluation method for class 1 vessels by using elastic-plastic finite element analysis in code case of JSME rules on design and construction

    International Nuclear Information System (INIS)

    Asada, Seiji; Hirano, Takashi; Nagata, Tetsuya; Kasahara, Naoto

    2008-01-01

    A structural evaluation method by using elastic-plastic finite element analysis has been developed and published as a code case of Rules on Design and Construction for Nuclear Power Plants (The First Part: Light Water Reactor Structural Design Standard) in the JSME Codes for Nuclear Power Generation Facilities. Its title is 'Alternative Structural Evaluation Criteria for Class 1 Vessels Based on Elastic-Plastic Finite Element Analysis' (NC-CC-005). This code case applies elastic-plastic analysis to evaluation of such failure modes as plastic collapse, thermal ratchet, fatigue and so on. Advantage of this evaluation method is free from stress classification, consistently use of Mises stress and applicability to complex 3-dimensional structures which are hard to be treated by the conventional stress classification method. The evaluation method for plastic collapse has such variation as the Lower Bound Approach Method, Twice-Elastic-Slope Method and Elastic Compensation Method. Cyclic Yield Area (CYA) based on elastic analysis is applied to screening evaluation of thermal ratchet instead of secondary stress evaluation, and elastic-plastic analysis is performed when the CYA screening criteria is not satisfied. Strain concentration factors can be directly calculated based on elastic-plastic analysis. (author)

  16. Feasibility analysis of the modified ATHLET code for supercritical water cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Chong, E-mail: ch.zhou@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany); Yang Yanhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Cheng Xu [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Modification of system code ATHLET for supercritical water application. Black-Right-Pointing-Pointer Development and assessment of a heat transfer package for supercritical water. Black-Right-Pointing-Pointer Validation of the modified code at supercritical pressures with the theoretical point-hydraulics model and the SASC code. Black-Right-Pointing-Pointer Application of the modified code to LOCA analysis of a supercritical water cooled in-pile fuel qualification test loop. - Abstract: Since the existing thermal-hydraulic computer codes for light water reactors are not applicable to supercritical water cooled reactors (SCWRs) owing to the limitation of physical models and numerical treatments, the development of a reliable thermal-hydraulic computer code is very important to design analysis and safety assessment of SCWRs. Based on earlier modification of ATHLET for SCWR, a general interface is implemented to the code, which serves as the platform for information exchange between ATHLET and the external independent physical modules. A heat transfer package containing five correlations for supercritical water is connected to the ATHLET code through the interface. The correlations are assessed with experimental data. To verify the modified ATHLET code, the Edwards-O'Brian blow-down test is simulated. As first validation at supercritical pressures, a simplified supercritical water cooled loop is modeled and its stability behavior is analyzed. Results are compared with that of the theoretical model and SASC code in the reference and show good agreement. To evaluate its feasibility, the modified ATHLET code is applied to a supercritical water cooled in-pile fuel qualification test loop. Loss of coolant accidents (LOCAs) due to break of coolant supply lines are calculated for the loop. Sensitivity analysis of some safety system parameters is performed to get further knowledge about their influence on the function of the

  17. Pipe stress analysis on HCCR-TBS ancillary systems in conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Eo Hwak [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • Pipe stress is performed on Korean HCCR-TBS for the load combinations including seismic events. • The resultant stress meets the requirement of the design code & standard except one position where modification is needed. • The results gives useful information for the design evolution in the next desgin phase. - Abstract: Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be tested in ITER to demonstrate feasibility of the breeding blanket concept. The HCCR-TBS comprises Test Blanket Module (TBM) with associated shield, and ancillary systems located in various positions of ITER building. Currently, conceptual design for the HCCR-TBS is in progress. This paper presents pipe stress analysis results for the HCCR-TBS ancillary systems. The pipe stress analysis was performed in accordance with ASME B31.3 for major pipes of the Helium Cooling System (HCS) and the Coolant Purification System (CPS), which are operated in high pressure and temperature. The pipe stress for various load cases and load combinations were calculated. Operational pressure and temperature during plasma operation are applied as pressure load and thermal load, respectively. In addition seismic events were combined to investigate the code compliance for sustained load case and occasional load case. It was confirmed that the resultant stress meets the requirements of ASME B31.3 except one position in which it needs modification. These results give useful information for the next design phase, for example, nozzle loads for the component selection, the support design parameters, etc.

  18. Pipe stress analysis on HCCR-TBS ancillary systems in conceptual design

    International Nuclear Information System (INIS)

    Ahn, Mu-Young; Cho, Seungyon; Lee, Eo Hwak; Park, Yi-Hyun; Lee, Youngmin

    2016-01-01

    Highlights: • Pipe stress is performed on Korean HCCR-TBS for the load combinations including seismic events. • The resultant stress meets the requirement of the design code & standard except one position where modification is needed. • The results gives useful information for the design evolution in the next desgin phase. - Abstract: Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be tested in ITER to demonstrate feasibility of the breeding blanket concept. The HCCR-TBS comprises Test Blanket Module (TBM) with associated shield, and ancillary systems located in various positions of ITER building. Currently, conceptual design for the HCCR-TBS is in progress. This paper presents pipe stress analysis results for the HCCR-TBS ancillary systems. The pipe stress analysis was performed in accordance with ASME B31.3 for major pipes of the Helium Cooling System (HCS) and the Coolant Purification System (CPS), which are operated in high pressure and temperature. The pipe stress for various load cases and load combinations were calculated. Operational pressure and temperature during plasma operation are applied as pressure load and thermal load, respectively. In addition seismic events were combined to investigate the code compliance for sustained load case and occasional load case. It was confirmed that the resultant stress meets the requirements of ASME B31.3 except one position in which it needs modification. These results give useful information for the next design phase, for example, nozzle loads for the component selection, the support design parameters, etc.

  19. A review of residual stress analysis using thermoelastic techniques

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, A F; Dulieu-Barton, J M; Quinn, S [University of Southampton, School of Engineering Sciences, Highfield, Southampton, SO17 1BJ (United Kingdom); Burguete, R L [Airbus UK Ltd., New Filton House, Filton, Bristol, BS99 7AR (United Kingdom)

    2009-08-01

    Thermoelastic Stress Analysis (TSA) is a full-field technique for experimental stress analysis that is based on infra-red thermography. The technique has proved to be extremely effective for studying elastic stress fields and is now well established. It is based on the measurement of the temperature change that occurs as a result of a stress change. As residual stress is essentially a mean stress it is accepted that the linear form of the TSA relationship cannot be used to evaluate residual stresses. However, there are situations where this linear relationship is not valid or departures in material properties due to manufacturing procedures have enabled evaluations of residual stresses. The purpose of this paper is to review the current status of using a TSA based approach for the evaluation of residual stresses and to provide some examples of where promising results have been obtained.

  20. A review of residual stress analysis using thermoelastic techniques

    International Nuclear Information System (INIS)

    Robinson, A F; Dulieu-Barton, J M; Quinn, S; Burguete, R L

    2009-01-01

    Thermoelastic Stress Analysis (TSA) is a full-field technique for experimental stress analysis that is based on infra-red thermography. The technique has proved to be extremely effective for studying elastic stress fields and is now well established. It is based on the measurement of the temperature change that occurs as a result of a stress change. As residual stress is essentially a mean stress it is accepted that the linear form of the TSA relationship cannot be used to evaluate residual stresses. However, there are situations where this linear relationship is not valid or departures in material properties due to manufacturing procedures have enabled evaluations of residual stresses. The purpose of this paper is to review the current status of using a TSA based approach for the evaluation of residual stresses and to provide some examples of where promising results have been obtained.

  1. X-ray stress analysis of residual stress gradients in surface layers of steel

    International Nuclear Information System (INIS)

    Ganev, N.; Kraus, I.; Gosmanova, G.; Pfeiffer, L.; Tietz, H.-D.

    2001-01-01

    The aim of the contribution is to present the theoretical possibilities of X-ray non-destructive identification of stress gradients within the penetration depth of used radiation and its utilization for experimental stress analysis. Practical usefullness of outlined speculations is illustrated with results of stress measurements on cut and shot-penned steel samples. (author)

  2. The Role of Non-­Coding RNA in Plant Stress

    KAUST Repository

    MacPherson, Cameron R.

    2012-01-01

    Post-transcriptional gene silencing (PTGS) is a powerful mechanism that can be adapted to genetically modify crop plants. PTGS operates in many plant signaling pathways including those mediating stress responses. Given the small number of mi

  3. Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

    International Nuclear Information System (INIS)

    1997-08-01

    In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

  4. 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)

  5. MKENO-DAR: a direct angular representation Monte Carlo code for criticality safety analysis

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Komuro, Yuichi; Tsunoo, Yukiyasu; Nakayama, Mitsuo.

    1984-03-01

    Improving the Monte Carlo code MULTI-KENO, the MKENO-DAR (Direct Angular Representation) code has been developed for criticality safety analysis in detail. A function was added to MULTI-KENO for representing anisotropic scattering strictly. With this function, the scattering angle of neutron is determined not by the average scattering angle μ-bar of the Pl Legendre polynomial but by the random work operation using probability distribution function produced with the higher order Legendre polynomials. This code is avilable for the FACOM-M380 computer. This report is a computer code manual for MKENO-DAR. (author)

  6. 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.)

  7. 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.)

  8. 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.)

  9. 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

  10. Perspectives on the development of next generation reactor systems safety analysis codes

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  11. Perspectives on the development of next generation reactor systems safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  12. 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)

  13. Development of an analysis code for pressure wave propagation, (1)

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Sakano, Kosuke; Shindo, Yoshihisa

    1974-11-01

    We analyzed the propagation of the pressure-wave in the piping system of SWAT-1B rig by using SWAC-5 Code. We carried out analyses on the following parts. 1) A straight pipe 2) Branches 3) A piping system The results obtained in these analyses are as follows. 1) The present our model simulates well the straight pipe and the branch with the same diameters. 2) The present our model simulates approximately the branch with the different diameters and the piping system. (auth.)

  14. Performance analysis of simultaneous dense coding protocol under decoherence

    Science.gov (United States)

    Huang, Zhiming; Zhang, Cai; Situ, Haozhen

    2017-09-01

    The simultaneous dense coding (SDC) protocol is useful in designing quantum protocols. We analyze the performance of the SDC protocol under the influence of noisy quantum channels. Six kinds of paradigmatic Markovian noise along with one kind of non-Markovian noise are considered. The joint success probability of both receivers and the success probabilities of one receiver are calculated for three different locking operators. Some interesting properties have been found, such as invariance and symmetry. Among the three locking operators we consider, the SWAP gate is most resistant to noise and results in the same success probabilities for both receivers.

  15. Using MCNP code for neutron and photon skyshine analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Netecha, M.E. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-03-01

    The MCNP Monte-Carlo code was used for the investigation of the sensitivity of neutron and neutron-induced secondary photon dose rate, total and thermal neutron fluxes and space-energy distributions to energy and angular distribution of radiation source, to thickness and composition of the ground, air density (including it changing with height), humidities of air and ground, thermalization effects, detector's dimension and its disposal above the ground level. The calculations were performed with the assumption that the source or released radiation into the atmosphere can be treated as a point source and the source containment structure has a negligible perturbation on the skyshine radiation field. (author)

  16. 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)

  17. Criticality qualification of a new Monte Carlo code for reactor core analysis

    International Nuclear Information System (INIS)

    Catsaros, N.; Gaveau, B.; Jaekel, M.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.; Zisis, Th.

    2009-01-01

    In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal-hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the 'Accelerator part' of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the 'Reactor part' of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.

  18. Thermal-hydraulic analysis of the Three Mile Island Unit 2 reactor accident with THALES code

    International Nuclear Information System (INIS)

    Hashimoto, Kazuichiro; Soda, Kunihisa

    1991-10-01

    The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident computer codes and to assess the capability of the codes. The TMI-2 Analysis Exercise was performed at the Japan Atomic Energy Research Institute (JAERI) using the THALES (Thermal-Hydraulic Analysis of Loss-of-Coolant, Emergency Core Cooling and Severe Core Damage) - PM1/TMI code. The purpose of the analysis is to verify the capability of THALES-PM1/TMI code to describe accident progression in the actual plant. The present paper describes the final result of the TMI-2 Analysis Exercise performed at JAERI. (author)

  19. PHEBUS FP release analysis using a microstructure-based code

    International Nuclear Information System (INIS)

    Carlucci, L.N.

    1992-03-01

    The results of pre-test fission-product (FP) release analyses of the first two PHEBUS FP experiments, FPT0 and FPT1, indicate that the FREEDOM microstructure-based code predicts significant differences in both the timing and percent of gaseous FP releases for the two tests. To provide an indication of its predictive capability, FREEDOM was also used to model the high-burnup fuel tested in the Oak Ridge National Laboratory experiments VI-2 and VI-3. For these, the code was found to overpredict releases during the early stages of the tests and to underpredict releases during the later stages. The release kinetics in both tests were reasonably predicted, however. In view of the above, it is likely that the FREEDOM predictions of the final cumulative releases for the first two PHEBUS FP tests are lower-bound estimates. However, the significant difference in the predicted timing of initial releases for the two tests is felt to be indicative of what will occur. Therefore, this difference should be considered in the planning and conduct of the two tests, particularly aspects related to on-line measurements

  20. EDPUFF- a Gaussian dispersion code for consequence analysis

    International Nuclear Information System (INIS)

    Oza, R.B.; Bapat, V.N.; Nair, R.N.; Hukkoo, R.K.; Krishnamoorthy, T.M.

    1995-01-01

    EDPUFF- Equi Distance Puff is a Gaussian dispersion code in FORTRAN language to model atmospheric dispersion of instantaneous or continuous point source releases. It is designed to incorporate the effect of changing meteorological conditions and source release rates on the spatial distribution profiles and its consequences. Effects of variation of parameters like puff spacing, puff packing, averaging schemes are discussed and the choice of the best values for minimum errors and minimum computer CPU time are identified. The code calculates the doses to individual receptors as well as average doses for population zones from internal and external routes over the area of interest. Internal dose computations are made for inhalation and ingestion pathways while the doses from external route consists of cloud doses and doses from surface deposited activity. It computes inhalation and ingestion dose (milk route only) for critical group (1 yr old child). In case of population zones it finds out maximum possible doses in a given area along with the average doses discussed above. Report gives the doses from various pathways for unit release of fixed duration. (author). 7 refs., figs., 7 appendixes

  1. The error performance analysis over cyclic redundancy check codes

    Science.gov (United States)

    Yoon, Hee B.

    1991-06-01

    The burst error is generated in digital communication networks by various unpredictable conditions, which occur at high error rates, for short durations, and can impact services. To completely describe a burst error one has to know the bit pattern. This is impossible in practice on working systems. Therefore, under the memoryless binary symmetric channel (MBSC) assumptions, the performance evaluation or estimation schemes for digital signal 1 (DS1) transmission systems carrying live traffic is an interesting and important problem. This study will present some analytical methods, leading to efficient detecting algorithms of burst error using cyclic redundancy check (CRC) code. The definition of burst error is introduced using three different models. Among the three burst error models, the mathematical model is used in this study. The probability density function, function(b) of burst error of length b is proposed. The performance of CRC-n codes is evaluated and analyzed using function(b) through the use of a computer simulation model within CRC block burst error. The simulation result shows that the mean block burst error tends to approach the pattern of the burst error which random bit errors generate.

  2. Argentine nuclear fuels MOX irradiated in the Petten reactor: Analysis of experience with the BACO code

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A C; Perez, E; Adelfang, P [Argentine Atomic Energy Commission, Buenos Aires (Argentina)

    1997-08-01

    The irradiation of our first prototypes of MOX nuclear fuels fabricated in Argentina began in 1986. These experiences had been made in the HFR-Petten reactor, Holland. The six rods were fabricated in the {alpha} Facility (GAID-CNEA-Argentina). The first rod has been used for destructive pre-irradiation analysis in the KFK (Kernforschungszentrum Karlsruhe), Germany. The second one was a pathfinder for calibrating systems in the HFR. Another two rods included doped pellets based on iodine. One of them included CsI and auxiliary components. The second one included elemental iodine. The concentration of iodine was intended to simulate 15 MWd/ton(M) of burnup. We defined the power histories with the BACO code. We assumed a cycle of 15 days that included interaction treatments of cladding and pellet due to the power cycling. The last ramp is let run until stress corrosion cracking (SCC) is induced. The experience named BU15 was done with the last two rods. The final burnup was 15 MWd/ton(M), and a final ramp test was arranged for one of them. This burnup is the same as the previous two rods. The power level during irradiation was low and without major solicitations, only the normal shutdowns of the HFR. The ramp was similar to that used for the iodine test. We attempt to see the correct correspondence between the BU15 and the doping test. The pathfinder had an excellent behavior in the HFR reactor. The presence of microcracks inside the cladding was observed in the iodine test as we predicted with the BACO code. The post-irradiation tests of the BU15 experience has just ended. The development of the ramp was interrupted due to an increase of activity in the system. We presumed the presence of a failure in the rod. The visual inspection of the rod shows an atypical failure for this kind of fuel, i.e. they found a small circular hole. We use the BACO code for the behavior analysis of the fuel rods. 23 refs, 29 figs, 5 tabs.

  3. Argentine nuclear fuels MOX irradiated in the Petten reactor: Analysis of experience with the BACO code

    International Nuclear Information System (INIS)

    Marino, A.C.; Perez, E.; Adelfang, P.

    1997-01-01

    The irradiation of our first prototypes of MOX nuclear fuels fabricated in Argentina began in 1986. These experiences had been made in the HFR-Petten reactor, Holland. The six rods were fabricated in the α Facility (GAID-CNEA-Argentina). The first rod has been used for destructive pre-irradiation analysis in the KFK (Kernforschungszentrum Karlsruhe), Germany. The second one was a pathfinder for calibrating systems in the HFR. Another two rods included doped pellets based on iodine. One of them included CsI and auxiliary components. The second one included elemental iodine. The concentration of iodine was intended to simulate 15 MWd/ton(M) of burnup. We defined the power histories with the BACO code. We assumed a cycle of 15 days that included interaction treatments of cladding and pellet due to the power cycling. The last ramp is let run until stress corrosion cracking (SCC) is induced. The experience named BU15 was done with the last two rods. The final burnup was 15 MWd/ton(M), and a final ramp test was arranged for one of them. This burnup is the same as the previous two rods. The power level during irradiation was low and without major solicitations, only the normal shutdowns of the HFR. The ramp was similar to that used for the iodine test. We attempt to see the correct correspondence between the BU15 and the doping test. The pathfinder had an excellent behavior in the HFR reactor. The presence of microcracks inside the cladding was observed in the iodine test as we predicted with the BACO code. The post-irradiation tests of the BU15 experience has just ended. The development of the ramp was interrupted due to an increase of activity in the system. We presumed the presence of a failure in the rod. The visual inspection of the rod shows an atypical failure for this kind of fuel, i.e. they found a small circular hole. We use the BACO code for the behavior analysis of the fuel rods. 23 refs, 29 figs, 5 tabs

  4. High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

    International Nuclear Information System (INIS)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

    2013-01-01

    The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

  5. User's manual for seismic analysis code 'SONATINA-2V'

    Energy Technology Data Exchange (ETDEWEB)

    Hanawa, Satoshi; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-08-01

    The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

  6. Privacy and Psychosomatic Stress: An Empirical Analysis.

    Science.gov (United States)

    Webb, Stephen D.

    1978-01-01

    Examines the supposition that insufficient privacy is stressful to the individual. Data were obtained from urban centers in New Zealand. Findings support the hypothesis that a percieved lack of privacy is associated with psychosomatic stress. The relationship is specified by measures of stress and sex of respondents. (Author)

  7. Yield stress fluids slowly yield to analysis

    NARCIS (Netherlands)

    Bonn, D.; Denn, M.M.

    2009-01-01

    We are surrounded in everyday life by yield stress fluids: materials that behave as solids under small stresses but flow like liquids beyond a critical stress. For example, paint must flow under the brush, but remain fixed in a vertical film despite the force of gravity. Food products (such as

  8. Development and Validation of A Nuclear Fuel Cycle Analysis Tool: A FUTURE Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yoon Hee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy) code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

  9. DEVELOPMENT AND VALIDATION OF A NUCLEAR FUEL CYCLE ANALYSIS TOOL: A FUTURE CODE

    Directory of Open Access Journals (Sweden)

    S.K. KIM

    2013-10-01

    Full Text Available This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C# and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

  10. Development of seismic analysis model for HTGR core on commercial FEM code

    International Nuclear Information System (INIS)

    Tsuji, Nobumasa; Ohashi, Kazutaka

    2015-01-01

    The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test. (author)

  11. Jointly Decoded Raptor Codes: Analysis and Design for the BIAWGN Channel

    Directory of Open Access Journals (Sweden)

    Venkiah Auguste

    2009-01-01

    Full Text Available Abstract We are interested in the analysis and optimization of Raptor codes under a joint decoding framework, that is, when the precode and the fountain code exchange soft information iteratively. We develop an analytical asymptotic convergence analysis of the joint decoder, derive an optimization method for the design of efficient output degree distributions, and show that the new optimized distributions outperform the existing ones, both at long and moderate lengths. We also show that jointly decoded Raptor codes are robust to channel variation: they perform reasonably well over a wide range of channel capacities. This robustness property was already known for the erasure channel but not for the Gaussian channel. Finally, we discuss some finite length code design issues. Contrary to what is commonly believed, we show by simulations that using a relatively low rate for the precode , we can improve greatly the error floor performance of the Raptor code.

  12. Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

    International Nuclear Information System (INIS)

    Araya, F.

    1995-01-01

    After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

  13. Development of Deterministic and Probabilistic Fracture Mechanics Analysis Code PROFAS-RV for Reactor Pressure Vessel - Progress of the Work

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Min; Lee, Bong Sang [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, a deterministic/probabilistic fracture mechanics analysis program for reactor pressure vessel, PROFAS-RV, is developed. This program can evaluate failure probability of RPV using recent radiation embrittlement model of 10CFR50.61a and stress intensity factor calculation method of RCC-MRx code as well as the required basic functions of PFM program. Applications of some new radiation embrittlement model, material database, calculation method of stress intensity factors, and others which can improve fracture mechanics assessment of RPV are introduced. The purpose of this study is to develop a probabilistic fracture mechanics (PFM) analysis program for RPV considering above modification and application of newly developed models and calculation methods. In this paper, it deals with the development progress of the PFM analysis program for RPV, PROFAS-RV. The PROFAS-RV is being tested with other codes, and it is expected to revise and upgrade by reflecting the latest model and calculation method continuously. These efforts can minimize the uncertainty of the integrity evaluation for the reactor pressure vessel.

  14. Proteomic analysis of cold stress responses in tobacco seedlings ...

    African Journals Online (AJOL)

    Cold stress is one of the major abiotic stresses limiting the productivity and the geographical distribution of many important crops. To gain a better understanding of cold stress responses in tobacco (Nicotiana tabacum), we carried out a comparative proteomic analysis. Five-week-old tobacco seedlings were treated at 4°C ...

  15. Finite element stress analysis of brick-mortar masonry under ...

    African Journals Online (AJOL)

    Stress analysis of a brick-mortar couplet as a substitute for brick wall structure has been performed by finite element method, and algorithm for determining the element stiffness matrix for a plane stress problem using the displacement approach was developed. The nodal displacements were derived for the stress in each ...

  16. Evaluation of Long Stress-Induced Non-coding Transcripts 5 Polymorphism in Iranian Patients with Bladder Cancer

    Directory of Open Access Journals (Sweden)

    Mahla Nazari

    2016-07-01

    Full Text Available Background: Bladder cancer (BC is the most commonly diagnosed genitourinary cancer in Iran, presented in both men and women. BC is a multifactorial trait resulting from the complex interaction between several genes and environmental factors. Long stress-induced non-coding transcript 5 (LSINCT5, a member of the long non-coding RNAs, is abundantly expressed in high proliferative cells, as well as the cells vulnerable to cellular stress in response to chemical carcinogens. This case-control study aimed to determine any association between LSINCT5 rs2962586 polymorphism and bladder cancer. Materials and Methods: A group of 150 patients with BC were compared with 143 subjects as a control group. Genotyping of the rs2962586 polymorphism was done using tetra- primer amplification refractory mutation system-polymerase chain reaction (T-ARMS PCR method. Results: Genotype and allele distribution were not significantly different between the case and control groups. Smoking was found to be the confounding risk factor for bladder cancer. Conclusion: Considering the result of our analyses, it seems that LSINCT5 could not affect individual susceptibility to BC among Iranian patients, however, it can be considered as a disease predictor among smokers.

  17. Ice condenser containment analysis with the GOTHIC code

    International Nuclear Information System (INIS)

    Yadon, T.P.

    1996-01-01

    Analytical methodologies have recently been developed by Duke Power Company (Duke) to calculate the thermodynamic response of the ice condenser containment buildings at the McGuire and Catawba Nuclear Stations to high-energy line breaks. The GOTHIC computer code (Version 4.0) was utilized for these analyses. In the ice condenser containment design, a large mass of ice stored within the containment building is used to absorb the energy released from high-energy line breaks, thereby limiting the peak pressure and temperature in the containment building to within design limits. The McGuire and Catawba Nuclear Stations (both two-unit, 3411 MWth four-loop Westinghouse plants) are of the ice condenser containment design

  18. An Adaptation of the HELIOS/MASTER Code System to the Analysis of VHTR Cores

    International Nuclear Information System (INIS)

    Noh, Jae Man; Lee, Hyun Chul; Kim, Kang Seog; Kim, Yong Hee

    2006-01-01

    KAERI is developing a new computer code system for an analysis of VHTR cores based on the existing HELIOS/MASTER code system which was originally developed for a LWR core analysis. In the VHTR reactor physics, there are several unique neutronic characteristics that cannot be handled easily by the conventional computer code system applied for the LWR core analysis. Typical examples of such characteristics are a double heterogeneity problem due to the particulate fuels, the effects of a spectrum shift and a thermal up-scattering due to the graphite moderator, and a strong fuel/reflector interaction, etc. In order to facilitate an easy treatment of such characteristics, we developed some methodologies for the HELIOS/MASTER code system and tested their applicability to the VHTR core analysis

  19. Investigation and proposal of the system to affect nuclear fuel type authorization and analysis code certification

    International Nuclear Information System (INIS)

    2006-03-01

    In order to develop the system to affect more advanced and rational regulations of nuclear fuels and earlier introduction of new technologies in nuclear power plants, domestic and overseas safety regulation systems and state of their implementation for water cooled reactor fuel and safety analysis code had been investigated and new regulation system to affect nuclear fuel type authorization and analysis code certification was proposed. Topical report system for common parts related with nuclear fuel type authorization and analysis code certification was firstly proposed for knowledge base. Maintaining consistent safety examination supported by experts, introduction of domestic efficient system for lead irradiation test fuel, and analysis code certification and quality assurance were also proposed. (T. Tanaka)

  20. Comparative analysis of SLB for OPR1000 by using MEDUSA and CESEC-III codes

    International Nuclear Information System (INIS)

    Park, Jong Cheol; Park, Chan Eok; Kim, Shin Whan

    2005-01-01

    The MEDUSA is a system thermal hydraulics code developed by Korea Power Engineering Company (KOPEC) for Non-LOCA and LOCA analysis, using two fluid, three-field governing equations for two phase flow. The detailed descriptions for the MEDUSA code are given in Reference. A lot of effort is now being made to investigate the applicability of the MEDUSA code especially to Non-LOCA analysis, by comparing the analysis results with those from the current licensing code, CESEC-III: The comparative simulations of Pressurizer Level Control System(PLCS) Malfunction and Feedwater Line Break(FLB), which have been accomplished by C.E.Park and M.T.Oh, respectively, already showed that the MEDUSA code is applicable to the analysis of Non-LOCA events. In this paper, detailed thermal hydraulic analyses for Steam Line Break(SLB) without loss of off-site power were performed using the MEDUSA code. The calculation results were also compared with the CESEC-III, 1000(OPR1000), for the purpose of the code verification

  1. Empirical Analysis of Using Erasure Coding in Outsourcing Data Storage With Provable Security

    Science.gov (United States)

    2016-06-01

    computing and communication technologies become powerful and advanced , people are exchanging a huge amount of data, and they are de- manding more storage...NAVAL POSTGRADUATE SCHOOL MONTEREY, CALIFORNIA THESIS EMPIRICAL ANALYSIS OF USING ERASURE CODING IN OUTSOURCING DATA STORAGEWITH PROVABLE SECURITY by...2015 to 06-17-2016 4. TITLE AND SUBTITLE EMPIRICAL ANALYSIS OF USING ERASURE CODING IN OUTSOURCING DATA STORAGE WITH PROVABLE SECURITY 5. FUNDING

  2. A probabilistic analysis of PWR and BWR fuel rod performance using the code CASINO-SLEUTH

    International Nuclear Information System (INIS)

    Bull, A.J.

    1987-01-01

    This paper presents a brief description of the Monte Carlo and response surface techniques used in the code, and a probabilistic analysis of fuel rod performance in PWR and BWR applications. The analysis shows that fission gas release predictions are very sensitive to changes in certain of the code's inputs, identifies the most dominant input parameters and compares their effects in the two cases. (orig./HP)

  3. Study plan for the sensitivity analysis of the Terrain-Responsive Atmospheric Code (TRAC)

    International Nuclear Information System (INIS)

    Restrepo, L.F.; Deitesfeld, C.A.

    1987-01-01

    Rocky Flats Plant, Golden, Colorado is presently developing a computer code to model the dispersion of potential or actual releases of radioactive or toxic materials to the environment, along with the public consequences from these releases. The model, the Terrain-Responsive Atmospheric Code (TRAC), considers several complex features which could affect the overall dispersion and consequences. To help validate TRAC, a sensitivity analysis is being planned to determine how sensitive the model's solutions are to input variables. This report contains a brief description of the code, along with a list of tasks and resources needed to complete the sensitivity analysis

  4. Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Park, Jee Won; Chung, Bub Dong; Kim, Soo Hyung; Kim, See Dal

    2007-07-01

    The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE

  5. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    Neufert, A.; Van de Velde, A.

    1999-01-01

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  6. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  7. Satellite III non-coding RNAs show distinct and stress-specific patterns of induction

    International Nuclear Information System (INIS)

    Sengupta, Sonali; Parihar, Rashmi; Ganesh, Subramaniam

    2009-01-01

    The heat shock response in human cells is associated with the transcription of satellite III repeats (SatIII) located in the 9q12 locus. Upon induction, the SatIII transcripts remain associated with the locus and recruit several transcription and splicing factors to form the nuclear stress bodies (nSBs). The nSBs are thought to modulate epigenetic changes during the heat shock response. We demonstrate here that the nSBs are induced by a variety of stressors and show stress-specific patterns of induction. While the transcription factor HSF1 is required for the induction of SatIII locus by the stressors tested, its specific role in the transcriptional process appears to be stress dependent. Our results suggest the existence of multiple transcriptional loci for the SatIII transcripts and that their activation might depend upon the type of stressors. Thus, induction of SatIII transcripts appears to be a generic response to a variety of stress conditions.

  8. Round robin analysis on stress intensity factor of inner surface cracks in welded stainless steel pipes

    Energy Technology Data Exchange (ETDEWEB)

    Han, Chang Gi; Chang, Yoon Suk [Dept. of Nuclear Engineering, College of Engineering, Kyung Hee University, Yongin (Korea, Republic of); Kim, Jong Sung [Dept. of Mechanical Engineering, Sunchon National University, Sunchon (Korea, Republic of); Kim, Maan Won [Central Research Institute, Korea Hydro and Nuclear Power Company, Daejeon (Korea, Republic of)

    2016-12-15

    Austenitic stainless steels (ASSs) are widely used for nuclear pipes as they exhibit a good combination of mechanical properties and corrosion resistance. However, high tensile residual stresses may occur in ASS welds because postweld heat treatment is not generally conducted in order to avoid sensitization, which causes a stress corrosion crack. In this study, round robin analyses on stress intensity factors (SIFs) were carried out to examine the appropriateness of structural integrity assessment methods for ASS pipe welds with two types of circumferential cracks. Typical stress profiles were generated from finite element analyses by considering residual stresses and normal operating conditions. Then, SIFs of cracked ASS pipes were determined by analytical equations represented in fitness-for-service assessment codes as well as reference finite element analyses. The discrepancies of estimated SIFs among round robin participants were confirmed due to different assessment procedures and relevant considerations, as well as the mistakes of participants. The effects of uncertainty factors on SIFs were deducted from sensitivity analyses and, based on the similarity and conservatism compared with detailed finite element analysis results, the R6 code, taking into account the applied internal pressure and combination of stress components, was recommended as the optimum procedure for SIF estimation.

  9. 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…

  10. Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

    1998-09-01

    The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs.

  11. Assessment on the characteristics of the analysis code for KALIMER PSDRS

    Energy Technology Data Exchange (ETDEWEB)

    Eoh, Jae Hyuk; Sim, Yoon Sub; Kim, Seong O.; Kim, Yeon Sik; Kim, Eui Kwang; Wi, Myung Hwan [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The PARS2 code was developed to analyze the RHR(Residual Heat Removal) system, especially PSDRS(Passive Safety Decay Heat Removal System), of KALIMER. In this report, preliminary verification and sensitivity analyses for PARS2 code were performed. From the results of the analyses, the PARS2 code has a good agreement with the experimental data of CRIEPI in the range of turbulent airside flow, and also the radiation heat transfer mode was well predicted. In this verification work, it was founded that the code calculation stopped in a very low air flowrate, and the numerical scheme related to the convergence of PARS2 code was adjusted to solve this problem. Through the sensitivity analysis on the PARS2 calculation results from the change of the input parameters, the pool-mixing coefficient related to the heat capacity of the structure in the system was improved such that the physical phenomenon can be well predicted. Also the initial conditions for the code calculation such as the hot and cold pool temperatures at the PSDRS commencing time were set up by using the transient analysis of the COMMIX code, and the surface emissivity of PSDRS was investigated and its permitted variation rage was set up. From this study, overall sensitivity characteristics of the PARS2 code were investigated and the results of the sensitivity analyses can be used in the design of the RHR system of KALIMER. 14 refs., 28 figs., 2 tabs. (Author)

  12. Development and application of best-estimate LWR safety analysis codes

    International Nuclear Information System (INIS)

    Reocreux, M.

    1997-01-01

    This paper is a review of the status and the future orientations of the development and application of best estimate LWR safety analysis codes. The present status of these codes exhibits a large success and almost a complete fulfillment of the objectives which were assigned in the 70s. The applications of Best Estimate codes are numerous and cover a large variety of safety questions. However these applications raised a number of problems. The first ones concern the need to have a better control of the quality of the results. This means requirements on code assessment and on uncertainties evaluation. The second ones concern needs for code development and specifically regarding physical models, numerics, coupling with other codes and programming. The analysis of the orientations for code developments and applications in the next years, shows that some developments should be made without delay in order to solve today questions whereas some others are more long term and should be tested for example in some pilot programmes before being eventually applied in main code development. Each of these development programmes are analyzed in the paper by detailing their main content and their possible interest. (author)

  13. Improvement of multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok

    1998-09-01

    The MARS (Multi-dimensional Analysis of Reactor Safety) code is a multi-dimensional, best-estimate thermal-hydraulic system analysis code. This report describes the new features that have been improved in the MARS 1.3 code since the release of MARS 1.3 in July 1998. The new features include: - implementation of point kinetics model into the 3D module - unification of the heat structure model - extension of the control function to the 3D module variables - improvement of the 3D module input check function. Each of the items has been implemented in the developmental version of the MARS 1.3.1 code and, then, independently verified and assessed. The effectiveness of the new features is well verified and it is shown that these improvements greatly extend the code capability and enhance the user friendliness. Relevant input data changes are also described. In addition to the improvements, this report briefly summarizes the future code developmental activities that are being carried out or planned, such as coupling of MARS 1.3 with the containment code CONTEMPT and the three-dimensional reactor kinetics code MASTER 2.0. (author). 8 refs

  14. Progress on RMC: a Monte Carlo neutron transport code for reactor analysis

    International Nuclear Information System (INIS)

    Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin

    2011-01-01

    This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)

  15. Development of three dimensional transient analysis code STTA for SCWR core

    International Nuclear Information System (INIS)

    Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping

    2015-01-01

    Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation

  16. Development and analysis of hydraulic-material transfer analysis code taking density current into account

    International Nuclear Information System (INIS)

    Saito, Hiroaki; Iriya, Yoshikazu

    1999-03-01

    It is an important issue to select site for the underground disposal of high level radioactive waste in a stable environment. Modelling of hydraulics in the freshwater/seawater boundaries is required. In this study, the analyzer code has been modified, in order to enable the analysis under more various conditions. Input/output functions were modified, after the functions of each module and major parameters were reconsidered. The modification included the change of input mode, from dialogue mode to file mode. Specifications of input/output and parameters are described. (A. Yamamoto)

  17. ATALANTA: a multicomponent pulsed neutron diffraction analysis code

    International Nuclear Information System (INIS)

    Benham, M.J.; Ross, D.K.

    1986-01-01

    The analysis of powder diffraction patterns from metal hydrogen systems present certain problems which have been addressed in a restructured profile analysis program. The heart of this program, ATALANTA, is a routine which locates and processes small sections of the data field to which a minimal number of Bragg peaks contribute intensity. The analysis of a three component test data set is presented in order to demonstrate the method. (author)

  18. Altered expression of long non-coding RNAs during genotoxic stress-induced cell death in human glioma cells.

    Science.gov (United States)

    Liu, Qian; Sun, Shanquan; Yu, Wei; Jiang, Jin; Zhuo, Fei; Qiu, Guoping; Xu, Shiye; Jiang, Xuli

    2015-04-01

    Long non-coding RNAs (lncRNAs), a recently discovered class of non-coding genes, are transcribed throughout the genome. Emerging evidence suggests that lncRNAs may be involved in modulating various aspects of tumor biology, including regulating gene activity in response to external stimuli or DNA damage. No data are available regarding the expression of lncRNAs during genotoxic stress-induced apoptosis and/or necrosis in human glioma cells. In this study, we detected a change in the expression of specific candidate lncRNAs (neat1, GAS5, TUG1, BC200, Malat1, MEG3, MIR155HG, PAR5, and ST7OT1) during DNA damage-induced apoptosis in human glioma cell lines (U251 and U87) using doxorubicin (DOX) and resveratrol (RES). We also detected the expression pattern of these lncRNAs in human glioma cell lines under necrosis induced using an increased dose of DOX. Our results reveal that the lncRNA expression patterns are distinct between genotoxic stress-induced apoptosis and necrosis in human glioma cells. The sets of lncRNA expressed during genotoxic stress-induced apoptosis were DNA-damaging agent-specific. Generally, MEG3 and ST7OT1 are up-regulated in both cell lines under apoptosis induced using both agents. The induction of GAS5 is only clearly detected during DOX-induced apoptosis, whereas the up-regulation of neat1 and MIR155HG is only found during RES-induced apoptosis in both cell lines. However, TUG1, BC200 and MIR155HG are down regulated when necrosis is induced using a high dose of DOX in both cell lines. In conclusion, our findings suggest that the distinct regulation of lncRNAs may possibly involve in the process of cellular defense against genotoxic agents.

  19. STAC -- a new Swedish code for statistical analysis of cracks in SG-tubes

    International Nuclear Information System (INIS)

    Poern, K.

    1997-01-01

    Steam generator (SG) tubes in pressurized water reactor plants are exposed to various types of degradation processes, among which stress corrosion cracking in particular has been observed. To be able to evaluate the safety importance of such cracking of SG-tubes one has to have a good and empirically founded knowledge about the scope and the size of the cracks as well as the rate of their continuous growth. The basis of experience is to a large extent constituted of the annually performed SG-inspections and crack sizing procedures. On the basis of this experience one can estimate the distribution of existing crack lengths, and modify this distribution with regard to maintenance (plugging) and the predicted rate of crack propagation. Finally, one can calculate the rupture probability of SG-tubes as a function of a given critical crack length. On account of the Swedish Nuclear Power Inspectorate an introductory study has been performed in order to get a survey of what has been done elsewhere in this field. The study resulted in a proposal of a computerizable model to be able to estimate the distribution of true cracks, to modify this distribution due to the crack growth and to compute the probability of tube rupture. The model has now been implemented in a compute code, called STAC (STatistical Analysis of Cracks). This paper is aimed to give a brief outline of the model to facilitate the understanding of the possibilities and limitations associated with the model

  20. Uncertainty and sensitivity analysis applied to coupled code calculations for a VVER plant transient

    International Nuclear Information System (INIS)

    Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K. D.

    2004-01-01

    The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics, is an important step to perform best-estimate plant transient calculations. It is generally agreed that the application of best-estimate methods should be supplemented by an uncertainty and sensitivity analysis to quantify the uncertainty of the results. The paper presents results from the application of the GRS uncertainty and sensitivity method for a VVER-440 plant transient, which was already studied earlier for the validation of coupled codes. For this application, the main steps of the uncertainty method are described. Typical results of the method applied to the analysis of the plant transient by several working groups using different coupled codes are presented and discussed The results demonstrate the capability of an uncertainty and sensitivity analysis. (authors)