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.

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.

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

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

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...

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

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.)

Validations and applications of the FEAST code

Energy Technology Data Exchange (ETDEWEB)

Xu, Z.; Tayal, M.; Lau, J.H.; Evinou, D. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Jun, J.S. [Korea Atomic Energy Research Inst. (Korea, Republic of)

1999-07-01

The FEAST (Finite Element Analysis for STresses) code is part of a suite of computer codes that are used to assess the structural integrity of CANDu fuel elements and bundles. A detailed validation of the FEAST code was recently performed. The FEAST calculations are in good agreement with a variety of analytical solutions (18 cases) for stresses, strains and displacements. This consistency shows that the FEAST code correctly incorporates the fundamentals of stress analysis. Further, the calculations of the FEAST code match the variations in axial and hoop strain profiles, measured by strain gauges near the sheath-endcap weld during an out-reactor compression test. The code calculations are also consistent with photoelastic measurements in simulated endcaps. (author)

Validations and applications of the FEAST code

International Nuclear Information System (INIS)

Xu, Z.; Tayal, M.; Lau, J.H.; Evinou, D.; Jun, J.S.

1999-01-01

The FEAST (Finite Element Analysis for STresses) code is part of a suite of computer codes that are used to assess the structural integrity of CANDu fuel elements and bundles. A detailed validation of the FEAST code was recently performed. The FEAST calculations are in good agreement with a variety of analytical solutions (18 cases) for stresses, strains and displacements. This consistency shows that the FEAST code correctly incorporates the fundamentals of stress analysis. Further, the calculations of the FEAST code match the variations in axial and hoop strain profiles, measured by strain gauges near the sheath-endcap weld during an out-reactor compression test. The code calculations are also consistent with photoelastic measurements in simulated endcaps. (author)

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

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)

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

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.

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)

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)

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.)

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

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.)

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)

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

Strength evaluation code STEP for brittle materials

International Nuclear Information System (INIS)

Ishihara, Masahiro; Futakawa, Masatoshi.

1997-12-01

In a structural design using brittle materials such as graphite and/or ceramics it is necessary to evaluate the strength of component under complex stress condition. The strength of ceramic materials is said to be influenced by the stress distribution. However, in the structural design criteria simplified stress limits had been adopted without taking account of the strength change with the stress distribution. It is, therefore, important to evaluate the strength of component on the basis of the fracture model for brittle material. Consequently, the strength evaluation program, STEP, on a brittle fracture of ceramic materials based on the competing risk theory had been developed. Two different brittle fracture modes, a surface layer fracture mode dominated by surface flaws and an internal fracture mode by internal flaws, are treated in the STEP code in order to evaluate the strength of brittle fracture. The STEP code uses stress calculation results including complex shape of structures analyzed by the generalized FEM stress analysis code, ABAQUS, so as to be possible to evaluate the strength of brittle fracture for the structures having complicate shapes. This code is, therefore, useful to evaluate the structural integrity of arbitrary shapes of components such as core graphite components in the HTTR, heat exchanger components made of ceramics materials etc. This paper describes the basic equations applying to the STEP code, code system with a combination of the STEP and the ABAQUS codes and the result of the verification analysis. (author)

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.)

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

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

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

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

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%

Elastic creep-fatigue evaluation for ASME code

International Nuclear Information System (INIS)

Severud, L.K.; Winkel, B.V.

1987-01-01

Experience with applying the ASME Code Case N-47 rules for evaluation of creep-fatigue with elastic analysis results has been problematic. The new elastic evaluation methods are intended to bound the stress level and strain range values needed for use in employing the code inelastic analysis creep-fatigue damage counting procedures. To account for elastic followup effects, ad hoc rules for stress classification, shakedown, and ratcheting are employed. Because elastic followup, inelastic strain concentration, and stress-time effects are accounted for, the design fatigue curves in Case N-47 for inelastic analysis are used instead of the more conservative elastic analysis curves. Creep damage assessments are made using an envelope stress-time history that treats multiple load events and repeated cycles during elevated temperature service life. (orig./GL)

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

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)

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.)

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

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.

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

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.)

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)

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)

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

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

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

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)....

Finite-element formulations for the thermal stress analysis of two- and three-dimensional thin ractor structures

International Nuclear Information System (INIS)

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

1977-01-01

This paper describes finite-element formulations for the thermal stress analysis of LMFBR structures. The first formulation is applicable to large displacement rotation problems in which the strains are small. For this formulation, a general temperature-dependent constituent relationship is derived from a Gibbs potential function and a temperature dependent yield surface. The temperature dependency of the yield surface is based upon a temperature-dependent, material-hardening model. The model uses a temperature-equivalent stress-plastic strain diagram which is generated from isothermal uniaxial stress-strain data. A second formulation is presented for problems characterized by both large displacement-rotations and large strains. Here a set of large strain hypoelastic-plastic relationships are developed to linearly relate the rate of stress to the rate of deformation. The temperature field is described through time-dependent values at mesh node points; the temperature fields in each element are then obtained by interpolation formulas. Hence, problems with both spatial and temporal dependent temperature fields can easily be treated. The above developments were incorporated into two ANL developed finite-element computer codes: the implicit version of STRAW and the 3D Implicit Structural Analysis Code. STRAW is a two-dimensional code with a plane stress/plane strain beam element. The 3D Implicit code has a triangular flat plate element which is capable of sustaining both membrane and bending loads. To insure numerical stability both codes are based on an iterative-incremental solution procedure with equilibrium checks based on an error in energy

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

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.

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.

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)

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

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

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)

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.

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.)

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.

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.

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.)

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.)

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

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.

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

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)

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)

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

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

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.

Documentation and verification of the SHAFT code

International Nuclear Information System (INIS)

St John, C.M.

1991-12-01

The SHAFT code incorporates equations to compute stresses in a shaft liner when the rock through which a shaft passes is subject to known three-dimensional states of stress or strain. The deformation modes considered are hoop deformation, axial deformation, and shear on a plane normal to the shaft axis. Interaction between the liner and the soil and rock is considered, and it is assumed that the liner is in place before loading is applied. This code is intended to be used interactively but creates a permanent record complete with necessary quality assurance information. The code has been carefully verified for the case of generalized plane strain, in which an arbitrary axial strain can be defined. It may also be used for plane stress analysis. Output is given in the form of stresses at selected sample points in the linear and the rock and a simple graphical representation of the distribution of stress through the liner. 12 figs., 13 tabs

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

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)

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...

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.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

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

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

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...

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.

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)

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)

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...

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)

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

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)

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

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)

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

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

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)

STIGMA STIG STEGT STAGT STABA, Stress Analysis of Dragon HTR Graphite Structure

International Nuclear Information System (INIS)

Kinkead, A.N.

2002-01-01

1 - Nature of the physical problem solved: Stress analysis of graphite structures for the DRAGON high temperature reactor is performed by this family of computer codes. Two-dimensional plane strain irradiation dose dependent core problems have been solved. 2 - Method of solution: STAGT, which is the oldest in this series of programmes, can handle multiply connected regions but is confined to plane strain in x-y geometry. Variations in temperature loading during irradiation is accounted for (Wigner strain component.) STIG, is a version of STAGT where an anisotropic elasticity matrix has been introduced to handle transversely isotropic materials. An additional feature of 'STIG' is the introduction of a boundary restraint condition of practical importance to prismatic gas cooled reactor core construction. This is defined as rotational plane strain in which free distortion of the prism arising from overall gradient of temperature and/or fast neutron damage flux coincident with any single direction may be assumed to occur if variation of thermal expansion coefficient with irradiation is included. 'STIGMA' is intended for evaluation of stress and displacement in composite axisymmetrical bodies subject to variable loadings in the axial and radial directions. The code has been prepared to take account of transverse isotropy in material characteristics for up to four separate bonded interface zones within a single composite material problem. Although specifically designed for the analysis of graphite structural components in the fast neutron irradiation environment of a reactor core, it is equally applicable to initial state design of prestressed concrete pressure vessels and other problems involving rotational symmetry. 'STABA'-stress,temperature and bowing analysis. The aim of this quasi 3-D computer code is to apply the principle of rotational plane strain over the full length of a prismatic core component, taking into account spatial variations in fast neutron and

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

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.

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

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.

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/

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

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)

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.

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.

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

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.

Analysis of stress intensity factors for surface cracks in pre/post penetration

International Nuclear Information System (INIS)

Miyoshi, Toshiro; Yoshida, Yuichiro

1988-01-01

It is important to evaluate the penetration of surface cracks in a Leak-Before-Break analysis. Because the stress intensity factors for surface cracks in pre/post penetration had not yet been analyzed, the authors carried three-dimensional boundary element analyses in order to obtain them. First, the authors developed the technique of nodal breakdown appropriate for cracks with short ligament length in a two-dimensional boundary element analysis. Next, analyses of stress intensity factor for surface cracks in pre/post penetration were carried out using the technique of nodal breakdown for cracks with short ligament length and the three-dimensional boundary element code BEM 3 D which was designed for a supercomputer. (author)

The Aster code

International Nuclear Information System (INIS)

Delbecq, J.M.

1999-01-01

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.)

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)

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

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.

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.

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.

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)

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

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)

Baking system for ports of experimental advanced super-conducting tokamak vacuum vessel and thermal stress analysis

International Nuclear Information System (INIS)

Cheng Yali; Bao Liman; Song Yuntao; Yao Damao

2006-01-01

The baking system of Experimental Advanced Super-Conducting Toakamk (EAST) vacuum vessel is necessary to obtain the baking temperature of 150 degree C. In order to define suitable alloy heaters and achieve their reasonable layouts, thermal analysis was carried out with ANSYS code. The analysis results indicate that the temperature distribution and thermal stress of most parts of EAST vacuum vessel ports are uniform, satisfied for the requirement, and are safe based on ASME criterion. Feasible idea on reducing the stress focus is also considered. (authors)

GAPCON-THERMAL-3 code description

International Nuclear Information System (INIS)

Lanning, D.D.; Mohr, C.L.; Panisko, F.E.; Stewart, K.B.

1978-01-01

GAPCON-3 is a computer program that predicts the thermal and mechanical behavior of an operating fuel rod during its normal lifetime. The code calculates temperatures, dimensions, stresses, and strains for the fuel and the cladding in both the radial and axial directions for each step of the user specified power history. The method of weighted residuals is for the steady state temperature calculation, and is combined with a finite difference approximation of the time derivative for transient conditions. The stress strain analysis employs an iterative axisymmetric finite element procedure that includes plasticity and creep for normal and pellet-clad mechanical interaction loads. GAPCON-3 can solve steady state and operational transient problems. Comparisons of GAPCON-3 predictions to both closed form analytical solutions and actual inpile instrumented fuel rod data have demonstrated the ability of the code to calculate fuel rod behavior. GAPCON-3 features a restart capability and an associated plot package unavailable in previous GAPCON series codes

GAPCON-THERMAL-3 code description

Energy Technology Data Exchange (ETDEWEB)

Lanning, D.D.; Mohr, C.L.; Panisko, F.E.; Stewart, K.B.

1978-01-01

GAPCON-3 is a computer program that predicts the thermal and mechanical behavior of an operating fuel rod during its normal lifetime. The code calculates temperatures, dimensions, stresses, and strains for the fuel and the cladding in both the radial and axial directions for each step of the user specified power history. The method of weighted residuals is for the steady state temperature calculation, and is combined with a finite difference approximation of the time derivative for transient conditions. The stress strain analysis employs an iterative axisymmetric finite element procedure that includes plasticity and creep for normal and pellet-clad mechanical interaction loads. GAPCON-3 can solve steady state and operational transient problems. Comparisons of GAPCON-3 predictions to both closed form analytical solutions and actual inpile instrumented fuel rod data have demonstrated the ability of the code to calculate fuel rod behavior. GAPCON-3 features a restart capability and an associated plot package unavailable in previous GAPCON series codes.

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

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)

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)

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)

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

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

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.

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)

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…

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)

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

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)

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

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

Evaluation on double-wall-tube residual stress distribution of sodium-heated steam generator by neutron diffraction and numerical analysis

International Nuclear Information System (INIS)

Kisohara, N.; Suzuki, H.; Akita, K.; Kasahara, N.

2012-01-01

A double-wall-tube is nominated for the steam generator heat transfer tube of future sodium fast reactors (SFRs) in Japan, to decrease the possibility of sodium/water reaction. The double-wall-tube consists of an inner tube and an outer tube, and they are mechanically contacted to keep the heat transfer of the interface between the inner and outer tubes by their residual stress. During long term SG operation, the contact stress at the interface gradually falls down due to stress relaxation. This phenomenon might increase the thermal resistance of the interface and degrade the tube heat transfer performance. The contact stress relaxation can be predicted by numerical analysis, and the analysis requires the data of the initial residual stress distributions in the tubes. However, unclear initial residual stress distributions prevent precious relaxation evaluation. In order to resolve this issue, a neutron diffraction method was employed to reveal the tri-axial (radius, hoop and longitudinal) initial residual stress distributions in the double-wall-tube. Strain gauges also were used to evaluate the contact stress. The measurement results were analyzed using a JAEA's structural computer code to determine the initial residual stress distributions. Based on the stress distributions, the structural computer code has predicted the transition of the relaxation and the decrease of the contact stress. The radial and longitudinal temperature distributions in the tubes were input to the structural analysis model. Since the radial thermal expansion difference between the inner (colder) and outer (hotter) tube reduces the contact stress and the tube inside steam pressure contributes to increasing it, the analytical model also took these effects into consideration. It has been conduced that the inner and outer tubes are contacted with sufficient stresses during the plant life time, and that effective heat transfer degradation dose not occur in the double-wall-tube SG. (authors)

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)

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.

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

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)

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

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

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.)

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)

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

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.)

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)

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

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

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.

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

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

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.

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)

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).

Documentation and verification of the SHAFT code; Yucca Mountain Site Characterization Project

Energy Technology Data Exchange (ETDEWEB)

St.John, C.M. [Agapito (J.F.T.) and Associates, Inc., Grand Junction, CO (United States)

1991-12-01

The SHAFT code incorporates equations to compute stresses in a shaft liner when the rock through which a shaft passes is subject to known three-dimensional states of stress or strain. The deformation modes considered are hoop deformation, axial deformation, and shear on a plane normal to the shaft axis. Interaction between the liner and the soil and rock is considered, and it is assumed that the liner is in place before loading is applied. This code is intended to be used interactively but creates a permanent record complete with necessary quality assurance information. The code has been carefully verified for the case of generalized plane strain, in which an arbitrary axial strain can be defined. It may also be used for plane stress analysis. Output is given in the form of stresses at selected sample points in the linear and the rock and a simple graphical representation of the distribution of stress through the liner. 12 figs., 13 tabs.

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)

Modeling of PHWR fuel elements using FUDA code

International Nuclear Information System (INIS)

Tripathi, Rahul Mani; Soni, Rakesh; Prasad, P.N.; Pandarinathan, P.R.

2008-01-01

The computer code FUDA (Fuel Design Analysis) is used for modeling PHWR fuel bundle operation history and carry out fuel element thermo-mechanical analysis. The radial temperature profile across fuel and sheath, fission gas release, internal gas pressure, sheath stress and strains during the life of fuel bundle are estimated

Estimation of stress intensity factors for circumferential cracked pipes under welding residual stress filed

International Nuclear Information System (INIS)

Oh, Chang Young; Kim, Yun Jae; Oh, Young Jin; Song, Tae Kwang; Kim, Yong Beum; Oh, Young Jin; Song, Tae Kwang; Kim, Yong Beum

2012-01-01

Recently, stress corrosion cracking(SCC) have been found in dissimilar metal welds of nozzles in some pressurized water reactors and on low carbon stainless steel piping systems of boiling water reactors. The important factor of SCC is the residual stress field caused by weld. For the evaluation of crack growth analysis due to SCC, stress intensity factor under a residual stress field should be estimated. Several solutions for stress intensity factor under residual stress field were recommended in flaw assessment codes such as the American Society of Mechanical Engineers (ASME) Section XI, R6, American Petroleum Institute (API579). Some relevant works have been studied. Dong et al. evaluated stress intensity factors in welded structures. Miyazaki et al. estimated stress intensity factors of surface crack in simple stress fields. This paper presents a simple method to estimate stress intensity factors in welding residual stress field. For general application, results of structure integrity assessment codes KI solutions were compared Finite element analyses of welding simulation and cracked pipes are described. Comparison results of KI solutions and proposed simplified solution are presented in the works

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)

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.

Development of a computer code 'CRACK' for elastic and elastoplastic fracture mechanics analysis of 2-D structures by finite element technique

International Nuclear Information System (INIS)

Dutta, B.K.; Kakodkar, A.; Maiti, S.K.

1986-01-01

The fracture mechanics analysis of nuclear components is required to ensure prevention of sudden failure due to dynamic loadings. The linear elastic analysis near to a crack tip shows presence of stress singularity at the crack tip. The simulation of this singularity in numerical methods enhance covergence capability. In finite element technique this can be achieved by placing mid nodes of 8 noded or 6 noded isoparametric elements, at one fourth ditance from crack tip. Present report details this characteristic of finite element, implementation of this element in a code 'CRACK', implementation of J-integral to compute stress intensity factor and solution of number of cases for elastic and elastoplastic fracture mechanics analysis. 6 refs., 6 figures. (author)

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)

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

Integrity evaluation of the pressure vessels of Angra-2 and Angra-3 reactors by stress analysis

International Nuclear Information System (INIS)

Gomes, E.

1978-01-01

The integrity of the reactor pressure vessel of the unit II/III of the Nuclear Power Station at 'Angras do Reis' is evaluated by stress analysis, through the dynamics relaxation method. For the solution of the problem an axisymmetric model is fixed. Initially, the data of the Oak Ridge Vessel V-7 is compared with those obtained by two computer programs used in this study. The methods used in the computer programs are FEM and DEM. A11 the results are compared with the ASME Code Section III 1974 edition. The range deviation is determined to 99% confidence limit, in order to minimize the error probabilities. Finally, the equivalent intensity stress obtained is calculated and compared with the acceptable values of the ASME Code Section III, 1974 edition [pt

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.

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

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 ...

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.

Strain-encoded cardiac MRI as an adjunct for dobutamine stress testing: incremental value to conventional wall motion analysis.

Science.gov (United States)

Korosoglou, Grigorios; Lossnitzer, Dirk; Schellberg, Dieter; Lewien, Antje; Wochele, Angela; Schaeufele, Tim; Neizel, Mirja; Steen, Henning; Giannitsis, Evangelos; Katus, Hugo A; Osman, Nael F

2009-03-01

High-dose dobutamine stress MRI is safe and feasible for the diagnosis of coronary artery disease (CAD) in humans. However, the assessment of cine scans relies on the visual interpretation of regional wall motion, which is subjective. Recently, strain-encoded MRI (SENC) has been proposed for the direct color-coded visualization of myocardial strain. The purpose of our study was to compare the diagnostic value of SENC with that provided by conventional wall motion analysis for the detection of inducible ischemia during dobutamine stress MRI. Stress-induced ischemia was assessed by wall motion analysis and by SENC in 101 patients with suspected or known CAD and in 17 healthy volunteers who underwent dobutamine stress MRI in a clinical 1.5-T scanner. Quantitative coronary angiography deemed as the standard reference for the presence or absence of significant CAD (> or =50% diameter stenosis). On a coronary vessel level, SENC detected inducible ischemia in 86 of 101 versus 71 of 101 diseased coronary vessels (P or =50% stenosis (area under the curve, 0.96; SE, 0.01; 95% CI, 0.94 to 0.98; P<0.001). The direct color-coded visualization of strain on MR images is a useful adjunct for dobutamine stress MRI, which provides incremental value for the detection of CAD compared with conventional wall motion readings on cine images.

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

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.

Preliminary evaluation of the stress analysis reports for Angra I reactor coolant loop - part 1

International Nuclear Information System (INIS)

Ribeiro, S.V.G.; Andrade, J.E.L. de

1980-03-01

A methodology that will allow CNEN to approve the stress analysis reports of the components of the Brazilian nuclear power plants, was developed. The reactor coolant loop (RCL)of Angra I was checkd. This is the first part of the complete report and consists of the approval of the design documents, the approval of the equipment support models and the aproval of the steam generator dynamic model. The second part of this work is under way now and should contain the approval of the RCL stress and fatigue analysis according to ASME code section III. As shown in section 7 it appears necessary additional information from Westinghouse about the design of the RCL. (Author) [pt

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...

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

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)

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.

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.)

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

Reducing teacher stress and burnout in high-risk secondary schools in South Africa using transactional analysis

Directory of Open Access Journals (Sweden)

Sharon Mary Johnson

2015-01-01

Full Text Available One of a number of articles arising from PhD research, this paper focuses on the results of applying transactional analysis as one of three approaches to reducing stress and burnout for teachers in high-risk secondary schools in the gangland areas of the Cape Flats, Western Cape, South Africa. The other approaches were Trauma Release Exercises (TRE and Transpersonal Psychology (TP, and related articles on these and on the quantitative statistical analysis elements of this research are being disseminated elsewhere. A total of 43 teachers in three different schools took part in one intervention held weekly over 10 weeks for one and a half hours (15 hours in total at their school as part of staff development, with a control group of 20 teachers at a fourth school. Qualitative TA intervention questionnaire coding analysis and focus group post-intervention thematic analysis of the mixed-methods study are presented. Coding analysis focused on the intra- and inter-individual tools that impacted teachers, and it was found that TA generated self-awareness, self-help tools and a strong group connection. Thematic analysis gave insights into the physical, emotional and cognitive responses to stress and burnout interventions on the individual, interpersonal and organisational levels and revealed new perspectives on classroom competency, with teachers taking more responsibility for discipline in the classroom. The study gave insights into the well-being and coping of educators who survive in these challenging contexts, and it is proposed that TRE, TP and TA approaches can be incorporated, and possibly combined, into integrative and eclectic ways of dealing with complex psychological challenges of stress and burnout reduction in traumatic environments.

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.

Finite-element formulations for the thermal stress analysis of two- and three-dimensional thin reactor structures

International Nuclear Information System (INIS)

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

1977-01-01

This paper describes finite-element formulations for the thermal stress analysis of LMFBR structures. The first formulation is applicable to large displacement rotation problems in which the strains are small. For this formulation, a general temperature-dependent constituent relationship is derived from a Gibbs potential and a temperature dependent surface. A second formulation is presented for problems characterized by both large displacement-rotations and large strains. Here a set of large strain hypoelastic-plastic relationships are developed to linearly relate the rate of stress to the rate of deformation. These developments were incorporated into two ANL developed finite-element computer codes: the implicit version of STRAW and the 3D Implicit Structural Analaysis code. A set of problems is presented to validate both the 3D and 2D programs and to illustrate their applicability to a variety of problems. (Auth.)

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.

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

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

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.

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

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

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)

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)

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)

Development of the containment transient analysis code for the passive reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-05-01

This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

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)

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.)

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...

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.

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

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.)

FEM Analysis and Measurement of Residual Stress by Neutron Diffraction on the Dissimilar Overlay Weld Pipe

International Nuclear Information System (INIS)

Kim, Kang Soo; Lee, Ho Jin; Woo, Wan Chuck; Seong, Baek Seok; Byeon, Jin Gwi; Park, Kwang Soo; Jung, In Chul

2010-01-01

Much research has been done to estimate the residual stress on a dissimilar metal weld. There are many methods to estimate the weld residual stress and FEM (Finite Element Method) is generally used due to the advantage of the parametric study. And the X-ray method and a Hole Drilling technique for an experimental method are also usually used. The aim of this paper is to develop the appropriate FEM model to estimate the residual stresses of the dissimilar overlay weld pipe. For this, firstly, the specimen of the dissimilar overlay weld pipe was manufactured. The SA 508 Gr3 nozzle, the SA 182 safe end and SA376 pipe were welded by the Alloy 182. And the overlay weld by the Alloy 52M was performed. The residual stress of this specimen was measured by using the Neutron Diffraction device in the HANARO (High-flux Advanced Neutron Application ReactOr) research reactor, KAERI (Korea Atomic Energy Research Institute). Secondly, FEM Model on the dissimilar overlay weld pipe was made and analyzed by the ABAQUS Code (ABAQUS, 2004). Thermal analysis and stress analysis were performed, and the residual stress was calculated. Thirdly, the results of the FEM analysis were compared with those of the experimental methods

Code conforming determination of cumulative usage factors for general elastic-plastic finite element analyses

International Nuclear Information System (INIS)

Rudolph, Juergen; Goetz, Andreas; Hilpert, Roland

2012-01-01

The procedures of fatigue analyses of several relevant nuclear and conventional design codes (ASME, KTA, EN, AD) for power plant components differentiate between an elastic, simplified elastic-plastic and elastic-plastic fatigue check. As a rule, operational load levels will exclude the purely elastic fatigue check. The application of the code procedure of the simplified elastic-plastic fatigue check is common practice. Nevertheless, resulting cumulative usage factors may be overly conservative mainly due to high code based plastification penalty factors Ke. As a consequence, the more complex and still code conforming general elastic-plastic fatigue analysis methodology based on non-linear finite element analysis (FEA) is applied for fatigue design as an alternative. The requirements of the FEA and the material law to be applied have to be clarified in a first step. Current design codes only give rough guidelines on these relevant items. While the procedure for the simplified elastic-plastic fatigue analysis and the associated code passages are based on stress related cycle counting and the determination of pseudo elastic equivalent stress ranges, an adaptation to elastic-plastic strains and strain ranges is required for the elastic-plastic fatigue check. The associated requirements are explained in detail in the paper. If the established and implemented evaluation mechanism (cycle counting according to the peak and valley respectively the rainflow method, calculation of stress ranges from arbitrary load-time histories and determination of cumulative usage factors based on all load events) is to be retained, a conversion of elastic-plastic strains and strain ranges into pseudo elastic stress ranges is required. The algorithm to be applied is described in the paper. It has to be implemented in the sense of an extended post processing operation of FEA e.g. by APDL scripts in ANSYS registered . Variations of principal stress (strain) directions during the loading

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.

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

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.

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.

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)

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.

Utilization of the MCNP-3A code for criticality safety analysis

International Nuclear Information System (INIS)

Maragni, M.G.; Moreira, J.M.L.

1996-01-01

In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

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)

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. ...

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)

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.

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

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

Evaluation of the MMCLIFE 3.0 code in predicting crack growth in titanium aluminide composites

International Nuclear Information System (INIS)

Harmon, D.; Larsen, J.M.

1999-01-01

Crack growth and fatigue life predictions made with the MMCLIFE 3.0 code are compared to test data for unidirectional, continuously reinforced SCS-6/Ti-14Al-21Nb (wt pct) composite laminates. The MMCLIFE 3.0 analysis package is a design tool capable of predicting strength and fatigue performance in metal matrix composite (MMC) laminates. The code uses a combination of micromechanic lamina and macromechanic laminate analyses to predict stresses and uses linear elastic fracture mechanics to predict crack growth. The crack growth analysis includes a fiber bridging model to predict the growth of matrix flaws in 0 degree laminates and is capable of predicting the effects of interfacial shear stress and thermal residual stresses. The code has also been modified to include edge-notch flaws in addition to center-notch flaws. The model was correlated with constant amplitude, isothermal data from crack growth tests conducted on 0- and 90 degree SCS-6/Ti-14-21 laminates. Spectrum fatigue tests were conducted, which included dwell times and frequency effects. Strengths and areas for improvement for the analysis are discussed

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

Validation of a Subchannel Analysis Code MATRA Version 1.0

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Seo, Kyung Won; Kwon, Hyouk

2008-10-15

A subchannel analysis code MATRA has been developed for the thermal hydraulic analysis of SMART core. The governing equations and important models were established, and validation calculations have been performed for subchannel flow and enthalpy distributions in rod bundles under steady-state conditions. The governing equations of the MATRA were on the basis of integral balance equation of the two-phase mixture. The effects of non-homogeneous and non-equilibrium states were considered by employing the subcooled boiling model and the phasic slip model. Solution scheme and main structure of the MATRA code, as well as the difference of MATRA and COBRA-IV-I codes, were summarized. Eight different test data sets were employed for the validation of the MATRA code. The collected data consisted of single-phase subchannel flow and temperature distribution data, single-phase inlet flow maldistribution data, single-phase partial flow blockage data, and two-phase subchannel flow and enthalpy distribution data. The prediction accuracy as well as the limitation of the MATRA code was evaluated from this analysis.

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)

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.)

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.

Coupled transient thermo-fluid/thermal-stress analysis approach in a VTBM setting

International Nuclear Information System (INIS)

Ying, A.; Narula, M.; Zhang, H.; Abdou, M.

2008-01-01

A virtual test blanket module (VTBM) has been envisioned as a utility to aid in streamlining and optimizing the US ITER TBM design effort by providing an integrated multi-code, multi-physics modeling environment. Within this effort, an integrated simulation approach is being developed for TBM design calculations and performance evaluation. Particularly, integrated thermo-fluid/thermal-stress analysis is important for enabling TBM design and performance calculations. In this paper, procedures involved in transient coupled thermo-fluid/thermal-stress analysis are investigated. The established procedure is applied to study the impact of pulsed operational phenomenon on the thermal-stress response of the TBM first wall. A two-way coupling between the thermal strain and temperature field is also studied, in the context of a change in thermal conductivity of the beryllium pebble bed in a solid breeder blanket TBM due to thermal strain. The temperature field determines the thermal strain in beryllium, which in turn changes the temperature field. Iterative thermo-fluid/thermal strain calculations have been applied to both steady-state and pulsed operation conditions. All calculations have been carried out in three dimensions with representative MCAD models, including all the TBM components in their entirety

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.

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

Stress-intensity factors for surface cracks in pipes: a computer code for evaluation by use of influence functions. Final report

International Nuclear Information System (INIS)

Dedhia, D.D.; Harris, D.O.

1982-06-01

A user-oriented computer program for the evaluation of stress intensity factors for cracks in pipes is presented. Stress intensity factors for semi-elliptical, complete circumferential and long longitudinal cracks can be obtained using this computer program. The code is based on the method of influence functions which makes it possible to treat arbitrary stresses on the plane of the crack. The stresses on the crack plane can be entered as a mathematical or tabulated function. A user's manual is included in this report. Background information is also included

Effect of residual stress on the integrity of a branch connection

International Nuclear Information System (INIS)

Law, M.; Kirstein, O.; Luzin, V.

2012-01-01

A new connection to an existing gas pipeline was made by hot-tapping, welding directly onto a pressurised pipeline. The welds were not post-weld heat treated, causing significant residual stresses. The critical weld had residual stresses determined by neutron diffraction using ANSTO's residual stress diffractometer, Kowari. The maximum measured residual stress (290 MPa) was 60% of the yield strength. The magnitudes of errors from a number of sources were estimated. An integrity assessment of the welded branch connection was performed with the measured residual stress values and with residual stress distributions from the BS 7910 and API 579 analysis codes. Analysis using estimates of residual stress from API 579 overestimated the critical crack size. Highlights: ► Residual stresses were measured by neutron diffraction in a thick section, non post-weld heat treated ferritic weld. ► There is little published data on these welds. ► The work compares the measured residual stresses with code-based residual stress distributions.

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.

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".

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

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

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.

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

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)

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.)

FEMAXI-III, a computer code for fuel rod performance analysis

International Nuclear Information System (INIS)

Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.

1983-01-01

This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)

Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

International Nuclear Information System (INIS)

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

2013-01-01

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

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.

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)

Contributions of the ORNL piping program to nuclear piping design codes and standards

International Nuclear Information System (INIS)

Moore, S.E.

1975-11-01

The ORNL Piping Program was conceived and established to develop basic information on the structural behavior of nuclear power plant piping components and to prepare this information in forms suitable for use in design analysis and codes and standards. One of the objectives was to develop and qualify stress indices and flexibility factors for direct use in Code-prescribed design analysis methods. Progress in this area is described

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

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)

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)

Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, J. Y.; Lee, K. H. (and others)

2007-07-15

Advanced Suites of deterministic codes for VHTR physics analysis has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. These code suites include the conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation, and a whole core transport code that can model local heterogeneities directly at the core level. Particular modeling issues in physics analysis of the gas-cooled VHTRs were resolved, which include a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment, temperature and burnup. And the geometry handling capability of the DeCART code were extended to deal with the hexagonal fuel elements of the VHTR core. The developed code suites were validated and verified by comparing the computational results with those of the Monte Carlo calculations for the benchmark problems.

Development of probabilistic fracture mechanics code PASCAL and user's manual

Energy Technology Data Exchange (ETDEWEB)

Shibata, Katsuyuki; Onizawa, Kunio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Li, Yinsheng; Kato, Daisuke [Fuji Research Institute Corporation, Tokyo (Japan)

2001-03-01

As a part of the aging and structural integrity research for LWR components, a new PFM (Probabilistic Fracture Mechanics) code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed since FY1996. This code evaluates the failure probability of an aged reactor pressure vessel subjected to transient loading such as PTS (Pressurized Thermal Shock). 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 methodologies and computer performance. The code has some new functions in optimized sampling and cell dividing procedure in stratified Monte Carlo simulation, elastic-plastic fracture criterion of R6 method, extension analysis models in semi-elliptical crack, evaluation of effect of thermal annealing and etc. In addition, an input data generator of temperature and stress distribution time histories was also prepared in the code. Functions and performance of the code have been confirmed based on the verification analyses and some case studies on the influence parameters. The present phase of the development will be completed in FY2000. Thus this report provides the user's manual and theoretical background of the code. (author)

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.

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.)

Analysis of irradiation-induced stresses in coating layers of coated fuel particles for the High Temperature Engineering Test Reactor (HTTR)

International Nuclear Information System (INIS)

Hayashi, Kimio; Kikuchi, Teruo; Fukuda, Kousaku; Sato, Sadao; Toyota, Junji; Shiozawa, Shusaku; Sawa, Kazuhiro; Kashimura, Satoru.

1991-07-01

Irradiation-induced stresses in coating layers of coated fuel particles were analyzed by the MICROS-2 code for the fuels of the High Temperature Engineering Test Reactor (HTTR) under its operating conditions. The analyses were made on the standard core fuel (A-type) and the test fuels comprising the advanced SiC-coated particle fuel (B-1 type) and the ZrC-coated particle fuel (B-2 type). For the B-1 type fuel, the stresses were relieved due to the thicker buffer and SiC layers than for the A type fuel. The slightly decreased thickness of the fourth layer for the B-1 type than for the A type fuel had no significant effect on the stresses. As for the B-2 type fuel, almost the same results as for the B-1 type were obtained under an assumption that the ZrC layer as well as the SiC layer undergoes negligible dimension change within the analysis conditions. The obtained results indicated that the B-1 and B-2 type fuels are better than the A type fuel in terms of integrity against the irradiation-induced stresses. Finally, research subjects for development of the analysis code on the fuel behavior are discussed. (author)

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)

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)

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.

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.

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)

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

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

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

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

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)

Purifying selection acts on coding and non-coding sequences of paralogous genes in Arabidopsis thaliana.

Science.gov (United States)

Hoffmann, Robert D; Palmgren, Michael

2016-06-13

Whole-genome duplications in the ancestors of many diverse species provided the genetic material for evolutionary novelty. Several models explain the retention of paralogous genes. However, how these models are reflected in the evolution of coding and non-coding sequences of paralogous genes is unknown. Here, we analyzed the coding and non-coding sequences of paralogous genes in Arabidopsis thaliana and compared these sequences with those of orthologous genes in Arabidopsis lyrata. Paralogs with lower expression than their duplicate had more nonsynonymous substitutions, were more likely to fractionate, and exhibited less similar expression patterns with their orthologs in the other species. Also, lower-expressed genes had greater tissue specificity. Orthologous conserved non-coding sequences in the promoters, introns, and 3' untranslated regions were less abundant at lower-expressed genes compared to their higher-expressed paralogs. A gene ontology (GO) term enrichment analysis showed that paralogs with similar expression levels were enriched in GO terms related to ribosomes, whereas paralogs with different expression levels were enriched in terms associated with stress responses. Loss of conserved non-coding sequences in one gene of a paralogous gene pair correlates with reduced expression levels that are more tissue specific. Together with increased mutation rates in the coding sequences, this suggests that similar forces of purifying selection act on coding and non-coding sequences. We propose that coding and non-coding sequences evolve concurrently following gene duplication.

Stress index development for piping with trunnion attachment under pressure and moment loading

International Nuclear Information System (INIS)

Lee, D. H.; Kim, J. M.; Park, S. H.

1997-01-01

A finite element analysis of a trunnion pipe anchor is presented. The structure is analyzed for the case of internal pressure and moment loadings. The stress results are categorized into the average (membrane) stress, the linearly varying (bending) stress and the peak stress through the thickness. The resulting stresses are interpreted per section III of the ASME boiler and pressure vessel code from which the Primary (B 1 ), Secondary (C 1 ) and Peak (K 1 ) stress indices for pressure, the Primary (B 2 ), Secondary (C 2 ) and Peak (K 2 ) stress indices for moment are developed. Based on the comparison between stress value by stress indices derived in this paper and stress value represented by the ASME Code Case N-391-1, the empirical equations for stress indices are effectively used in the piping stress analysis. Therefore, the use of empirical equations can simplify the procedure of evaluating the local stress in the piping design stage. (author)

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

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

Application of the French codes to the pressurized thermal shocks assessment

Energy Technology Data Exchange (ETDEWEB)

Chen, Mingya; Wang, Rong Shan; Yu, Weiwei; Lu, Feng; Zhang, Guo Dong; Xue, Fei; Chen, Zhilin [Suzhou Nuclear Power Research Institute, Life Management Center, Suzhou (China); Qian, Guian [Paul Scherrer Institute, Nuclear Energy and Safety Department, Villigen (Switzerland); Shi, Jinhua [Amec Foster Wheeler, Clean Energy Department, Gloucester (United Kingdom)

2016-12-15

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the 'screening criterion' for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no 'screening criterion'. In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

Application of the French Codes to the Pressurized Thermal Shocks Assessment

Directory of Open Access Journals (Sweden)

Mingya Chen

2016-12-01

Full Text Available The integrity of a reactor pressure vessel (RPV related to pressurized thermal shocks (PTSs has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the “screening criterion” for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no “screening criterion”. In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

Application of the French codes to the pressurized thermal shocks assessment

International Nuclear Information System (INIS)

Chen, Mingya; Wang, Rong Shan; Yu, Weiwei; Lu, Feng; Zhang, Guo Dong; Xue, Fei; Chen, Zhilin; Qian, Guian; Shi, Jinhua

2016-01-01

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the 'screening criterion' for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no 'screening criterion'. In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed

Surface displacements and pillar stresses associated with nuclear waste disposal in salt

International Nuclear Information System (INIS)

Hardy, M.P.; St John, C.M.

1977-01-01

A numerical model for regional analysis of stresses and displacement, resulting from heat generating waste placement in underground salt excavations, is presented. The model, which is an extension of that described by McClain and Starfield (1971), is based upon the displacement discontinuity method of stress analysis. It incorporates an empirical characterization of creep behavior of material on the excavation horizon and accounts for thermally induced stresses and displacements. The versatility of this approach is illustrated by the results of three relatively short simulations of test scale disposal facilities at shallow and greater depths. In addition, a three-dimensional code was used to evaluate the surface displacement history for a full-scale repository. This latter code, a thermoelastic analysis, gives an upper bound for the surface movements. It is concluded that the pillar stresses are the result of a complex non-linear interaction of many variables, and the maximum pillar stress can reach several multiples of the tributory-area pillar stress

Light water reactor fuel analysis code FEMAXI-V (Ver.1)

International Nuclear Information System (INIS)

Suzuki, Motoe

2000-09-01

A light water fuel analysis code FEMAXI-V is an advanced version which has been produced by integrating FEMAXI-IV(Ver.2), high burn-up fuel code EXBURN-I, and a number of functional improvements and extensions, to predict fuel rod behavior in normal and transient (not accident) conditions. The present report describes in detail the basic theories and structure, models and numerical solutions applied, improvements and extensions, and the material properties adopted in FEMAXI-V(Ver.1). FEMAXI-V deals with a single fuel rod. It predicts thermal and mechanical response of fuel rod to irradiation, including FP gas release. The thermal analysis predicts rod temperature distribution on the basis of pellet heat generation, changes in pellet thermal conductivity and gap thermal conductance, (transient) change in surface heat transfer to coolant, using radial one-dimensional geometry. The heat generation density profile of pellet can be determined by adopting the calculated results of burning analysis code. The mechanical analysis performs elastic/plastic, creep and PCMI calculations by FEM. The FP gas release model calculates diffusion of FP gas atoms and accumulation in bubbles, release and increase in internal pressure of rod. In every analysis, it is possible to allow some materials properties and empirical equations to depend on the local burnup or heat flux, which enables particularly analysis of high burnup fuel behavior and boiling transient of BWR rod. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

Verification of computer code FPRETAIN with respect to RIA data from SPERT and PBF experiments

International Nuclear Information System (INIS)

Heo, Young-Ho; Yanagisawa, Kazuaki.

1992-12-01

This report presents the comparisons between calculated and measured fuel rod behavior and the analysis of stress for preirradiated LWR type fuel rods during reactivity initiated accident (RIA) conditions. For the calculations, FPRETAIN computer code which can simulate the fuel behavior under RIA conditions at extended burnup stage was used. For the experimental, data obtained from the Special Power Excursion Reactor Test (SPERT) and the Power Burst Facility (PBF) tests were used. The results of the comparisons showed that the FPRETAIN code predicted well the tendency of the fuel rod behavior during RIA. From the results of the stress analysis, it was found that the maximum hoop stress of cladding was not proportional to the energy deposition of fuel rod. Calculated cladding maximum hoop stress of failed fuel at high burnup was not lower than that of intact fresh or low burnup fuel. (author)

Qinshan phase II extension nuclear power project thermal stratification and fatigue stress analysis for pressurizer surge line

International Nuclear Information System (INIS)

Yu Xiaofei; Zhang Yixiong; Ai Honglei

2010-01-01

Thermal stratification of pressurizer surge line induced by the inside fluid brings on 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, the loadings at connections of surge line to main pipe and RCP and the displacements of surge line at supports are obtained. (authors)

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)

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.

Development of a kinetics analysis code for fuel solution combined with thermal-hydraulics analysis code PHOENICS and analysis of natural-cooling characteristic test of TRACY. Contract research

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Shouichi; Yamane, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

Since exact information is not always acquired in the criticality accident of fuel solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation with can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a two-dimensional (R-Z) nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating with the thermal-hydraulics analysis code PHOENICS for all-purpose use the computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change. (author)

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

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

TRAC-CFD code integration and its application to containment analysis

International Nuclear Information System (INIS)

Tahara, M.; Arai, K.; Oikawa, H.

2004-01-01

Several safety systems utilizing natural driving force have been recently adopted for operating reactors, or applied to next-generation reactor design. Examples of these safety systems are the Passive Containment Cooling System (PCCS) and the Drywell Cooler (DWC) for removing decay heat, and the Passive Auto-catalytic Recombiner (PAR) for removing flammable gas in reactor containment during an accident. DWC is used in almost all Boiling Water Reactors (BWR) in service. PAR has been introduced for some reactors in Europe and will be introduced for Japanese reactors. PCCS is a safety device of next-generation BWR. The functional mechanism of these safety systems is closely related to the transient of the thermal-hydraulic condition of the containment atmosphere. The performance depends on the containment atmospheric condition, which is eventually affected by the mass and energy changes caused by the safety system. Therefore, the thermal fluid dynamics in the containment vessel should be appropriately considered in detail to properly estimate the performance of these systems. A computational fluid dynamics (CFD) code is useful for evaluating detailed thermal hydraulic behavior related to this equipment. However, it also requires a considerable amount of computational resources when it is applied to whole containment system transient analysis. The paper describes the method and structure of the integrated analysis tool, and discusses the results of its application to the start-up behavior analysis of a containment cooling system, a drywell local cooler. The integrated analysis code was developed and applied to estimate the DWC performance during a severe accident. The integrated analysis tool is composed of three codes, TRAC-PCV, CFD-DW and TRAC-CC, and analyzes the interaction of the natural convection and steam condensation of the DWC as well as analyzing the thermal hydraulic transient behavior of the containment vessel during a severe accident in detail. The

Sensitivity Analysis of FEAST-Metal Fuel Performance Code: Initial Results

International Nuclear Information System (INIS)

Edelmann, Paul Guy; Williams, Brian J.; Unal, Cetin; Yacout, Abdellatif

2012-01-01

This memo documents the completion of the LANL milestone, M3FT-12LA0202041, describing methodologies and initial results using FEAST-Metal. The FEAST-Metal code calculations for this work are being conducted at LANL in support of on-going activities related to sensitivity analysis of fuel performance codes. The objective is to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. This report summarizes our preliminary results for the sensitivity analysis using 6 calibration datasets for metallic fuel developed at ANL for EBR-II experiments. Sensitivity ranking methodology was deployed to narrow down the selected parameters for the current study. There are approximately 84 calibration parameters in the FEAST-Metal code, of which 32 were ultimately used in Phase II of this study. Preliminary results of this sensitivity analysis led to the following ranking of FEAST models for future calibration and improvements: fuel conductivity, fission gas transport/release, fuel creep, and precipitation kinetics. More validation data is needed to validate calibrated parameter distributions for future uncertainty quantification studies with FEAST-Metal. Results of this study also served to point out some code deficiencies and possible errors, and these are being investigated in order to determine root causes and to improve upon the existing code models.

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

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)

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)

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.

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)

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

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.

Subchannel analysis of a boiloff experiment by a system thermalhydraulic code

International Nuclear Information System (INIS)

Bousbia-Salah, A.; D'Auria, F.

2001-01-01

This paper presents the results of system thermalhydraulic code using the sub-channel analysis approach in predicting the Neptun boil off experiments. This approach will be suitable for further works in view of coupling the system code with a 3D neutron kinetic one. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the Neptun low pressure test No5002, which is a good repeat experiment, is considered. The calculations were carried out using the system transient analysis code Relap5/Mod3.2. A detailed nodalization of the Neptun test section was developed. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory, this demonstrates, as well, the reasonable success of the subchannel analysis approach adopted in the present context for a system thermalhydraulic code.(author)

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)

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)

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)

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

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....

Modelling of thermal stress in vapor generator supports

International Nuclear Information System (INIS)

Halpert, S.; Vazquez, L.

1997-01-01

To assure safety and availability of a nuclear power plant components or equipment stress analysis are done. When thermal loads are involved it's necessary to know the temperature field of the component or equipment. This paper describes the structural analysis of a steam generator lug with thermal load including the model used for computer simulation and presents the evolution of the temperature profile, the stress intensity and principal stress during start up and shut down of a nuclear power reactor. Temperature field obtained from code calculation show good agreement with the experimental data while stress analysis results are in agreement with a preview estimation. (author) [es

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.

The PARET code and the analysis of the SPERT I transients

Energy Technology Data Exchange (ETDEWEB)

Woodruff, William L [Argonne National Laboratory, Argonne (United States)

1983-09-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients.

The PARET code and the analysis of the SPERT I transients

International Nuclear Information System (INIS)

Woodruff, William L.

1983-01-01

The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients

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

Stress-Strain Analysis in TiN Nanocoating Deposited on Polymer with respect to Au Nanointerlayer

Directory of Open Access Journals (Sweden)

Magdalena Kopernik

2014-01-01

Full Text Available The multiscale analysis in the authors’ finite element code confirmed possibility of fracture, because of not sufficiently high level of compressive residual stress in the TiN deposited by physical deposition method and varied mechanical properties of the thin film and substrate. The residual stress cannot be identified by X-ray technique for amorphous polymer and layer with domains of crystalline TiN. It is assumed that the buffer biocompatible thin film of Au in the TiN/Bionate II material system will alter the evolution of residual stress and, therefore, will allow to determine the residual stress in profilometry studies, and helps to improve toughness of the connection between TiN and Bionate II. The introduction of Au nanocoating in the material system results in bending of the sample and a compressive residual stress in the TiN coating. Results of finite element simulation show improvement of connection between the polymer and TiN, and an increase of compressive residual stress in the coating by introduction of Au nanointerlayer results in reduction of stress and strain in the substrate (close to the boundary between substrate and coating.

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.

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.

Method for calculating internal radiation and ventilation with the ADINAT heat-flow code

International Nuclear Information System (INIS)

Butkovich, T.R.; Montan, D.N.

1980-01-01

One objective of the spent fuel test in Climax Stock granite (SFTC) is to correctly model the thermal transport, and the changes in the stress field and accompanying displacements from the application of the thermal loads. We have chosen the ADINA and ADINAT finite element codes to do these calculations. ADINAT is a heat transfer code compatible to the ADINA displacement and stress analysis code. The heat flow problem encountered at SFTC requires a code with conduction, radiation, and ventilation capabilities, which the present version of ADINAT does not have. We have devised a method for calculating internal radiation and ventilation with the ADINAT code. This method effectively reproduces the results from the TRUMP multi-dimensional finite difference code, which correctly models radiative heat transport between drift surfaces, conductive and convective thermal transport to and through air in the drifts, and mass flow of air in the drifts. The temperature histories for each node in the finite element mesh calculated with ADINAT using this method can be used directly in the ADINA thermal-mechanical calculation

Geometrically Nonlinear Shell Analysis of Wrinkled Thin-Film Membranes with Stress Concentrations

Science.gov (United States)

Tessler, Alexander; Sleight, David W.

2006-01-01

Geometrically nonlinear shell finite element analysis has recently been applied to solar-sail membrane problems in order to model the out-of-plane deformations due to structural wrinkling. Whereas certain problems lend themselves to achieving converged nonlinear solutions that compare favorably with experimental observations, solutions to tensioned membranes exhibiting high stress concentrations have been difficult to obtain even with the best nonlinear finite element codes and advanced shell element technology. In this paper, two numerical studies are presented that pave the way to improving the modeling of this class of nonlinear problems. The studies address the issues of mesh refinement and stress-concentration alleviation, and the effects of these modeling strategies on the ability to attain converged nonlinear deformations due to wrinkling. The numerical studies demonstrate that excessive mesh refinement in the regions of stress concentration may be disadvantageous to achieving wrinkled equilibrium states, causing the nonlinear solution to lock in the membrane response mode, while totally discarding the very low-energy bending response that is necessary to cause wrinkling deformation patterns.

Containment vessel stability analysis

International Nuclear Information System (INIS)

Harstead, G.A.; Morris, N.F.; Unsal, A.I.

1983-01-01

The stability analysis for a steel containment shell is presented herein. The containment is a freestanding shell consisting of a vertical cylinder with a hemispherical dome. It is stiffened by large ring stiffeners and relatively small longitudinal stiffeners. The containment vessel is subjected to both static and dynamic loads which can cause buckling. These loads must be combined prior to their use in a stability analysis. The buckling loads were computed with the aid of the ASME Code case N-284 used in conjunction with general purpose computer codes and in-house programs. The equations contained in the Code case were used to compute the knockdown factors due to shell imperfections. After these knockdown factors were applied to the critical stress states determined by freezing the maximum dynamic stresses and combining them with other static stresses, a linear bifurcation analysis was carried out with the aid of the BOSOR4 program. Since the containment shell contained large penetrations, the Code case had to be supplemented by a local buckling analysis of the shell area surrounding the largest penetration. This analysis was carried out with the aid of the NASTRAN program. Although the factor of safety against buckling obtained in this analysis was satisfactory, it is claimed that the use of the Code case knockdown factors are unduly conservative when applied to the analysis of buckling around penetrations. (orig.)

Analysis code for pressure in reactor containment vessel of ATR. CONPOL

International Nuclear Information System (INIS)

1997-08-01

For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

Stress Analysis of a TRISO Coated Particle Fuel by Using ABAQUS Finite Element Visco-Elastoplastic Solutions

International Nuclear Information System (INIS)

Cho, Moon Sung; Kim, Y. M.; Lee, Y. W.

2006-01-01

The fundamental design for a gas-cooled reactor relies on an understanding of the behavior of a coated particle fuel. KAERI, which has been carrying out the Korean VHTR (Very High Temperature modular gas cooled Reactor) Project since 2004, is developing a fuel performance analysis code for a VHTR named COPA (COated Particle fuel Analysis). A validation of COPA was attempted by comparing its benchmark results with the visco-elastic solutions obtained from the ABAQUS code calculations for the IAEA-CRP-6 TRISO coated particle benchmark problems involving a creep, swelling, and pressure. However, the ABAQUS finite element model used for the above-mentioned analysis did not consider the material nonlinearity of the SiC coating layer that showed stress levels higher than the assumed yield point of the material. In this study, a consideration of the material nonlinearity is included in the ABAQUS model to obtain the visco-elastoplastic solutions and the results are compared with the visco-elastic solutions obtained from the previous ABAQUS model

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

Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

Energy Technology Data Exchange (ETDEWEB)

Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of)

2014-05-15

The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will

Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

International Nuclear Information System (INIS)

Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong

2014-01-01

The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will

Thermal stress analysis of an Am/Cm stabilization bushing melter

International Nuclear Information System (INIS)

Gong, C.; Hardy, B.J.

1996-01-01

Decades of nuclear material production at the Savannah River Site (SRS) has resulted in the generation of large quantities of the isotopes Am 243 and Cm 244 . Currently, the Am and Cm isotopes are stored as a nitric acid solution in a tank. The Am and Cm isotopes have great commercial value but must be transferred to the Oak Ridge National Laboratory (ORNL) for processing. The nitric acid solution contains other isotopes and is intensely radioactive, which makes storage a problem and precludes shipment in the liquid form. In order to stabilize the material for onsite storage and to permit transport the material from SRS to ORNL, it has been proposed that the Am and Cm be separated from other isotopes in the solution and vitrified. The vitrification process in the Platinum-Rhodium alloy vessel generates a wide spectrum of temperature distributions. The melter is partially supported by a suspension system and confined by the flexible insulation. The combination of the fluctuation of temperature distribution and variable boundary conditions, induces stresses and strains in the melter. The thermal stress analysis is carried out with the finite element code ABAQUS. This analysis is closely associated with the design, manufacture and testing of the melter. The results were compared with the test data

Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE

International Nuclear Information System (INIS)

Kang, H-S; Jang, M-S; Kim, S-R; Park, J-M; Kim, K-N

2015-01-01

There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis

Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE

Energy Technology Data Exchange (ETDEWEB)

Kang, H-S; Jang, M-S; Kim, S-R [NESS, Daejeon (Korea, Republic of); Park, J-M; Kim, K-N [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis.

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)

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.

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

Energy Technology Data Exchange (ETDEWEB)

Tachibana, Haruo; Sekita, Tsutomu; Yamaguchi, Takenori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

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)

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.

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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)

[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.

User's manual and guide to SALT3 and SALT4: two-dimensional computer codes for analysis of test-scale underground excavations for the disposal of radioactive waste in bedded salt deposits

International Nuclear Information System (INIS)

Lindner, E.N.; St John, C.M.; Hart, R.D.

1984-02-01

SALT3 and SALT4 are two-dimensional analytical/displacement-discontinuity codes designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. These codes were developed by the University of Minnesota for the Office of Nuclear Waste Isolation in 1979. The present documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of these computer codes. The SALT3 and SALT4 codes can simulate: (a) viscoelastic behavior in pillars adjacent to excavations; (b) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (c) excavation sequence. Major advantages of these codes are: (a) computational efficiency; (b) the small amount of input data required; and (c) a creep law based on laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of the codes, i.e., the homogeneous elastic half-space and temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT3 and SALT4 codes can be used for parameter sensitivity analyses of two-dimensional, repository-scale, 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. SALT3 is a refinement of an earlier code, SALT, and includes a fully anelastic creep model and thermal stress routine. SALT4 is a later version, and incorporates a revised creep model which is strain-hardening

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.

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)

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.

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

Description and applicability of the BEFEM-CODE

Energy Technology Data Exchange (ETDEWEB)

Groth, T.

1980-05-15

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

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)

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.

Development of probabilistic fracture mechanics code PASCAL and user's manual

Energy Technology Data Exchange (ETDEWEB)

Shibata, Katsuyuki; Onizawa, Kunio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Li, Yinsheng; Kato, Daisuke [Fuji Research Institute Corporation, Tokyo (Japan)

2001-03-01

As a part of the aging and structural integrity research for LWR components, a new PFM (Probabilistic Fracture Mechanics) code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed since FY1996. This code evaluates the failure probability of an aged reactor pressure vessel subjected to transient loading such as PTS (Pressurized Thermal Shock). 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 methodologies and computer performance. The code has some new functions in optimized sampling and cell dividing procedure in stratified Monte Carlo simulation, elastic-plastic fracture criterion of R6 method, extension analysis models in semi-elliptical crack, evaluation of effect of thermal annealing and etc. In addition, an input data generator of temperature and stress distribution time histories was also prepared in the code. Functions and performance of the code have been confirmed based on the verification analyses and some case studies on the influence parameters. The present phase of the development will be completed in FY2000. Thus this report provides the user's manual and theoretical background of the code. (author)

JAERI thermal reactor standard code system for reactor design and analysis SRAC

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro

1985-01-01

SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

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)

Status of the CONTAIN computer code for LWR containment analysis

International Nuclear Information System (INIS)

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

1983-01-01

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

Status of the CONTAIN computer code for LWR containment analysis

International Nuclear Information System (INIS)

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

1982-01-01

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

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)

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.

Stress categorization in nozzle to pressure vessel connections finite elements models

International Nuclear Information System (INIS)

Albuquerque, Levi Barcelos de

1999-01-01

The ASME Boiler and Pressure Vessel Code, Section III , is the most important code for nuclear pressure vessels design. Its design criteria were developed to preclude the various pressure vessel failure modes throughout the so-called 'Design by Analysis', some of them by imposing stress limits. Thus, failure modes such as plastic collapse, excessive plastic deformation and incremental plastic deformation under cyclic loading (ratchetting) may be avoided by limiting the so-called primary and secondary stresses. At the time 'Design by Analysis' was developed (early 60's) the main tool for pressure vessel design was the shell discontinuity analysis, in which the results were given in membrane and bending stress distributions along shell sections. From that time, the Finite Element Method (FEM) has had a growing use in pressure vessels design. In this case, the stress results are neither normally separated in membrane and bending stress nor classified in primary and secondary stresses. This process of stress separation and classification in Finite Element (FE) results is what is called stress categorization. In order to perform the stress categorization to check results from FE models against the ASME Code stress limits, mainly from 3D solid FE models, several research works have been conducted. This work is included in this effort. First, a description of the ASME Code design criteria is presented. After that, a brief description of how the FEM can be used in pressure vessel design is showed. Several studies found in the literature on stress categorization for pressure vessel FE models are reviewed and commented. Then, the analyses done in this work are presented in which some typical nozzle to pressure vessel connections subjected to internal pressure and concentrated loads were modeled with solid finite elements. The results from linear elastic and limit load analyses are compared to each other and also with the results obtained by formulae for simple shell

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

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

Comparative transcriptome analysis of the Asteraceae halophyte Karelinia caspica under salt stress.

Science.gov (United States)

Zhang, Xia; Liao, Maoseng; Chang, Dan; Zhang, Fuchun

2014-12-17

Much attention has been given to the potential of halophytes as sources of tolerance traits for introduction into cereals. However, a great deal remains unknown about the diverse mechanisms employed by halophytes to cope with salinity. To characterize salt tolerance mechanisms underlying Karelinia caspica, an Asteraceae halophyte, we performed Large-scale transcriptomic analysis using a high-throughput Illumina sequencing platform. Comparative gene expression analysis was performed to correlate the effects of salt stress and ABA regulation at the molecular level. Total sequence reads generated by pyrosequencing were assembled into 287,185 non-redundant transcripts with an average length of 652 bp. Using the BLAST function in the Swiss-Prot, NCBI nr, GO, KEGG, and KOG databases, a total of 216,416 coding sequences associated with known proteins were annotated. Among these, 35,533 unigenes were classified into 69 gene ontology categories, and 18,378 unigenes were classified into 202 known pathways. Based on the fold changes observed when comparing the salt stress and control samples, 60,127 unigenes were differentially expressed, with 38,122 and 22,005 up- and down-regulated, respectively. Several of the differentially expressed genes are known to be involved in the signaling pathway of the plant hormone ABA, including ABA metabolism, transport, and sensing as well as the ABA signaling cascade. Transcriptome profiling of K. caspica contribute to a comprehensive understanding of K. caspica at the molecular level. Moreover, the global survey of differentially expressed genes in this species under salt stress and analyses of the effects of salt stress and ABA regulation will contribute to the identification and characterization of genes and molecular mechanisms underlying salt stress responses in Asteraceae plants.

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.

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.

An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

International Nuclear Information System (INIS)

Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

1991-05-01

The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)

Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

Energy Technology Data Exchange (ETDEWEB)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

2007-03-15

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

International Nuclear Information System (INIS)

Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong

2007-03-01

The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow

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

Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-01

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

Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes

International Nuclear Information System (INIS)

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

2006-01-01

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

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)

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

Sensitivity analysis of MIDAS tests using SPACE code. Effect of nodalization

International Nuclear Information System (INIS)

Eom, Shin; Oh, Seung-Jong; Diab, Aya

2018-01-01

The nodalization sensitivity analysis for the ECCS (Emergency Core Cooling System) bypass phe�nomena was performed using the SPACE (Safety and Performance Analysis CodE) thermal hydraulic analysis computer code. The results of MIDAS (Multi-�dimensional Investigation in Downcomer Annulus Simulation) test were used. The MIDAS test was conducted by the KAERI (Korea Atomic Energy Research Institute) for the performance evaluation of the ECC (Emergency Core Cooling) bypass phenomenon in the DVI (Direct Vessel Injection) system. The main aim of this study is to examine the sensitivity of the SPACE code results to the number of thermal hydraulic channels used to model the annulus region in the MIDAS experiment. The numerical model involves three nodalization cases (4, 6, and 12 channels) and the result show that the effect of nodalization on the bypass fraction for the high steam flow rate MIDAS tests is minimal. For computational efficiency, a 4 channel representation is recommended for the SPACE code nodalization. For the low steam flow rate tests, the SPACE code over-�predicts the bypass fraction irrespective of the nodalization finesse. The over-�prediction at low steam flow may be attributed to the difficulty to accurately represent the flow regime in the vicinity of the broken cold leg.

Sensitivity analysis of MIDAS tests using SPACE code. Effect of nodalization

Energy Technology Data Exchange (ETDEWEB)

Eom, Shin; Oh, Seung-Jong; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering

2018-02-15

The nodalization sensitivity analysis for the ECCS (Emergency Core Cooling System) bypass phe�nomena was performed using the SPACE (Safety and Performance Analysis CodE) thermal hydraulic analysis computer code. The results of MIDAS (Multi-�dimensional Investigation in Downcomer Annulus Simulation) test were used. The MIDAS test was conducted by the KAERI (Korea Atomic Energy Research Institute) for the performance evaluation of the ECC (Emergency Core Cooling) bypass phenomenon in the DVI (Direct Vessel Injection) system. The main aim of this study is to examine the sensitivity of the SPACE code results to the number of thermal hydraulic channels used to model the annulus region in the MIDAS experiment. The numerical model involves three nodalization cases (4, 6, and 12 channels) and the result show that the effect of nodalization on the bypass fraction for the high steam flow rate MIDAS tests is minimal. For computational efficiency, a 4 channel representation is recommended for the SPACE code nodalization. For the low steam flow rate tests, the SPACE code over-�predicts the bypass fraction irrespective of the nodalization finesse. The over-�prediction at low steam flow may be attributed to the difficulty to accurately represent the flow regime in the vicinity of the broken cold leg.

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

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.

A methodology for interpretation of overcoring stress measurements in anisotropic rock

International Nuclear Information System (INIS)

Hakala, M.; Sjoeberg, J.

2006-11-01

The in situ state of stress is an important parameter for the design of a repository for final disposal of spent nuclear fuel. This report presents work conducted to improve the quality of overcoring stress measurements, focused on the interpretation of overcoring rock stress measurements when accounting for possible anisotropic behavior of the rock. The work comprised: (i) development/upgrading of a computer code for calculating stresses from overcoring strains for anisotropic materials and for a general overcoring probe configuration (up to six strain rosettes with six gauges each), (ii) development of a computer code for determining elastic constants for transversely isotropic rocks from biaxial testing, and (iii) analysis of case studies of selected overcoring measurements in both isotropic and anisotropic rocks from the Posiva and SKB sites in Finland and Sweden, respectively. The work was principally limited to transversely isotropic materials, although the stress calculation code is applicable also to orthotropic materials. The developed computer codes have been geared to work primarily with the Borre and CSIRO HI three-dimensional overcoring measurement probes. Application of the codes to selected case studies, showed that the developed tools were practical and useful for interpreting overcoring stress measurements conducted in anisotropic rock. A quantitative assessment of the effects of anisotropy may thus be obtained, which provides increased reliability in the stress data. Potential gaps in existing data and/or understanding can also be identified. (orig.)

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.

Development of a detailed core flow analysis code for prismatic fuel reactors

International Nuclear Information System (INIS)

Bennett, R.G.

1990-01-01

The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR

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.

Stress analysis of perforated plates and plates with single and clustered nozzles

International Nuclear Information System (INIS)

Hulbert, L.E.; Hopper, A.T.; Rybicki, E.F.

1976-01-01

It is well known that stress distributions around a hole in a plate are drastically increased if there are other holes in the plate sufficiently near. Such stress interaction effects must be analyzed and accounted for in the design of a plate containing multiple perforations. Computer programs have been developed for the accurate analysis of the in-plane stresses of arbitrary multiholed plates. Stress interactions also occur between closely spaced nozzles attached to pressure vessels such as nuclear reactor vessels. Results of the investigations carried out during the second year of the research program are summarized. The first section of the report describes the structures analyzed, the loadings on these structures, and the mathematical statements of these problems. The second section explains the analysis techniques employed. The third section deals with the numerical results obtained for all of the problems solved. This section also gives a complete statement of each problem and the series of trial functions used to solve each nozzle-plate problem. Several appendices are also included. These appendices are intended to be users' manuals for the computer codes used in this research effort but which are not explained in the literature. Appendix A treats program TABLES while Appendix B explains the use of PEBBLES. Appendix C is the users' manual for program INTRTIE which is a large program actually consisting of six individual programs. Appendix D discusses the changes in the input data to program NONLIN to obtain certain input data for INTRTIE. Finally, Appendix E treats a small program called CONVERT which also generates input data for INTRTIE

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

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

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.

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.

Thermal-stress analysis and testing of DIII-D armor tiles

International Nuclear Information System (INIS)

Baxi, C.B.; Anderson, P.M.; Reis, E.E.; Smith, J.P.; Smith, P.D.; Croesmann, C.; Watkins, J.; Whitley, J.

1987-10-01

It is planned to install about 1500 new armor tiles in the DIII-D tokamak. The armor tiles currently installed in DIII-D are made by brazing Poco AXF-5Q graphite onto Inconel X-750 stock. A small percentage of these have failed by breakage of graphite. These failures were believed to be related to significant residual stress remaining in graphite after brazing. Hence, an effort was undertaken to improve the design with all-graphite tiles. Three criteria must be satisfied by the armor tiles and the hardware used to attach the tiles to the vessel walls: tiles should not structurally fail, peak tile temperature must be less than 2500 K, and peak vessel stresses must be below acceptable levels. A number of alternate design concepts were first analyzed with the two-dimensional finite element codes TOPAZ2D and NIKE2D. Promising designs were optimized for best parameters such as thicknesses, etc. The two best designs were further analyzed for thermal stresses with the three-dimensional codes P/THERMAL and P/STRESS. Prototype tiles of a number of materials were fabricated by GA and tested at the Plasma Materials Test Facility of the Sandia National Laboratory at Albuquerque. The tests simulated the heat flux and cooling conditions in DIII-D. This paper describes the 2-D and 3-D thermal stress analyses, the test results and logic which led to the selected design of the DIII-D armor tiles. 5 refs., 7 figs., 3 tabs

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)

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)

Parameter analysis on the temperature and thermal stress of the cylindrical structure with multiple holes

International Nuclear Information System (INIS)

Kang, Y. H.; Lee, Y. S.; Choi, Y. J.

2001-01-01

During fuel irradiation tests in the in-core of HANARO, all components of the capsule with multiple holes act as heat sources due to high gamma and fission heat. In the design stage, a series of design parameter study were performed to confirm the thermal integrity of the capsule with multiple holes. The main parameter reviewed in this study are as follows: the position of the specimen, the thickness of the support tube and gap size. From the analysis performed by using of the FEM code ANSYS. it is confirmed that gap size effect on the capsule is one of the most important parameter of the capsule integrity. And the final thermal stress and displacement of the support tube with a increase of gap size are also under the allowable limits of ASME code

Relationship between various pressure vessel and piping codes

International Nuclear Information System (INIS)

Canonico, D.A.

1976-01-01

Section VIII of the ASME Code provides stress allowable values for material specifications that are provided in Section II Parts A and B. Since the adoption of the ASME Code over 60 years ago the incidence of failure has been greatly reduced. The Codes are currently based on strength criteria and advancements in the technology of fracture toughness and fracture mechanics should permit an even greater degree of reliability and safety. This lecture discusses the various Sections of the Code. It describes the basis for the establishment of design stress allowables and promotes the idea of the use of fracture mechanics

Development of system analysis code for pyrochemical process using molten salt electrorefining

International Nuclear Information System (INIS)

Tozawa, K.; Matsumoto, T.; Kakehi, I.

2000-04-01

This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and development of an analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. The cathode processor calculation code with distillation process was developed. A code validation calculation has been conducted on the basic of the benchmark problem for natural convection in a square cavity. Results by using the present code agreed well for the velocity-temperature fields, the maximum velocity and its location with the benchmark solution published in a paper. The functions have been added to advance the reality in simulation and to increase the efficiency in utilization. The test run has been conducted using the code with the above modification for an axisymmetric enclosed vessel simulating a cathode processor, and the capability of the distillation process simulation with the code has been confirmed. An analytical model for cooling behavior of the pyrochemical process cell was developed. The analytical model was selected by comparing benchmark analysis with detailed analysis on engineering workstation. Flow and temperature distributions were confirmed by the result of steady state analysis. In the result of transient cooling analysis, an initial transient peak of temperature occurred at balanced heat condition in the steady-state analysis. Final gas temperature distribution was dependent on gas circulation flow in transient condition. Then there were different final gas temperature distributions on the basis of the result of steady-state analysis. This phenomenon has a potential for it's own metastable condition. Therefore it was necessary to design gas cooling flow pattern without cooling gas circulation

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

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)

Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

International Nuclear Information System (INIS)

Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min

2016-01-01

As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI

Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

Energy Technology Data Exchange (ETDEWEB)

Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI.