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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Validation of comprehensive space radiation transport code

International Nuclear Information System (INIS)

Shinn, J.L.; Simonsen, L.C.; Cucinotta, F.A.

1998-01-01

The HZETRN code has been developed over the past decade to evaluate the local radiation fields within sensitive materials on spacecraft in the space environment. Most of the more important nuclear and atomic processes are now modeled and evaluation within a complex spacecraft geometry with differing material components, including transition effects across boundaries of dissimilar materials, are included. The atomic/nuclear database and transport procedures have received limited validation in laboratory testing with high energy ion beams. The codes have been applied in design of the SAGE-III instrument resulting in material changes to control injurious neutron production, in the study of the Space Shuttle single event upsets, and in validation with space measurements (particle telescopes, tissue equivalent proportional counters, CR-39) on Shuttle and Mir. The present paper reviews the code development and presents recent results in laboratory and space flight validation

Validation uncertainty of MATRA code for subchannel void distributions

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae-Hyun; Kim, S. J.; Kwon, H.; Seo, K. W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

To extend code capability to the whole core subchannel analysis, pre-conditioned Krylov matrix solvers such as BiCGSTAB and GMRES are implemented in MATRA code as well as parallel computing algorithms using MPI and OPENMP. It is coded by fortran 90, and has some user friendly features such as graphic user interface. MATRA code was approved by Korean regulation body for design calculation of integral-type PWR named SMART. The major role subchannel code is to evaluate core thermal margin through the hot channel analysis and uncertainty evaluation for CHF predictions. In addition, it is potentially used for the best estimation of core thermal hydraulic field by incorporating into multiphysics and/or multi-scale code systems. In this study we examined a validation process for the subchannel code MATRA specifically in the prediction of subchannel void distributions. The primary objective of validation is to estimate a range within which the simulation modeling error lies. The experimental data for subchannel void distributions at steady state and transient conditions was provided on the framework of OECD/NEA UAM benchmark program. The validation uncertainty of MATRA code was evaluated for a specific experimental condition by comparing the simulation result and experimental data. A validation process should be preceded by code and solution verification. However, quantification of verification uncertainty was not addressed in this study. The validation uncertainty of the MATRA code for predicting subchannel void distribution was evaluated for a single data point of void fraction measurement at a 5x5 PWR test bundle on the framework of OECD UAM benchmark program. The validation standard uncertainties were evaluated as 4.2%, 3.9%, and 2.8% with the Monte-Carlo approach at the axial levels of 2216 mm, 2669 mm, and 3177 mm, respectively. The sensitivity coefficient approach revealed similar results of uncertainties but did not account for the nonlinear effects on the

Methodological considerations for observational coding of eating and feeding behaviors in children and their families.

Science.gov (United States)

Pesch, Megan H; Lumeng, Julie C

2017-12-15

Behavioral coding of videotaped eating and feeding interactions can provide researchers with rich observational data and unique insights into eating behaviors, food intake, food selection as well as interpersonal and mealtime dynamics of children and their families. Unlike self-report measures of eating and feeding practices, the coding of videotaped eating and feeding behaviors can allow for the quantitative and qualitative examinations of behaviors and practices that participants may not self-report. While this methodology is increasingly more common, behavioral coding protocols and methodology are not widely shared in the literature. This has important implications for validity and reliability of coding schemes across settings. Additional guidance on how to design, implement, code and analyze videotaped eating and feeding behaviors could contribute to advancing the science of behavioral nutrition. The objectives of this narrative review are to review methodology for the design, operationalization, and coding of videotaped behavioral eating and feeding data in children and their families, and to highlight best practices. When capturing eating and feeding behaviors through analysis of videotapes, it is important for the study and coding to be hypothesis driven. Study design considerations include how to best capture the target behaviors through selection of a controlled experimental laboratory environment versus home mealtime, duration of video recording, number of observations to achieve reliability across eating episodes, as well as technical issues in video recording and sound quality. Study design must also take into account plans for coding the target behaviors, which may include behavior frequency, duration, categorization or qualitative descriptors. Coding scheme creation and refinement occur through an iterative process. Reliability between coders can be challenging to achieve but is paramount to the scientific rigor of the methodology. Analysis approach

Results from the First Validation Phase of CAP code

International Nuclear Information System (INIS)

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

2010-01-01

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

Validity of vascular trauma codes at major trauma centres.

Science.gov (United States)

Altoijry, Abdulmajeed; Al-Omran, Mohammed; Lindsay, Thomas F; Johnston, K Wayne; Melo, Magda; Mamdani, Muhammad

2013-12-01

The use of administrative databases in vascular injury research has been increasing, but the validity of the diagnosis codes used in this research is uncertain. We assessed the positive predictive value (PPV) of International Classification of Diseases, tenth revision (ICD-10), vascular injury codes in administrative claims data in Ontario. We conducted a retrospective validation study using the Canadian Institute for Health Information Discharge Abstract Database, an administrative database that records all hospital admissions in Canada. We evaluated 380 randomly selected hospital discharge abstracts from the 2 main trauma centres in Toronto, Ont., St.Michael's Hospital and Sunnybrook Health Sciences Centre, between Apr. 1, 2002, and Mar. 31, 2010. We then compared these records with the corresponding patients' hospital charts to assess the level of agreement for procedure coding. We calculated the PPV and sensitivity to estimate the validity of vascular injury diagnosis coding. The overall PPV for vascular injury coding was estimated to be 95% (95% confidence interval [CI] 92.3-96.8). The PPV among code groups for neck, thorax, abdomen, upper extremity and lower extremity injuries ranged from 90.8 (95% CI 82.2-95.5) to 97.4 (95% CI 91.0-99.3), whereas sensitivity ranged from 90% (95% CI 81.5-94.8) to 98.7% (95% CI 92.9-99.8). Administrative claims hospital discharge data based on ICD-10 diagnosis codes have a high level of validity when identifying cases of vascular injury. Observational Study Level III.

Simulation codes and the impact of validation/uncertainty requirements

International Nuclear Information System (INIS)

Sills, H.E.

1995-01-01

Several of the OECD/CSNI members have adapted a proposed methodology for code validation and uncertainty assessment. Although the validation process adapted by members has a high degree of commonality, the uncertainty assessment processes selected are more variable, ranaing from subjective to formal. This paper describes the validation and uncertainty assessment process, the sources of uncertainty, methods of reducing uncertainty, and methods of assessing uncertainty.Examples are presented from the Ontario Hydro application of the validation methodology and uncertainty assessment to the system thermal hydraulics discipline and the TUF (1) system thermal hydraulics code. (author)

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.

Water evaporation over sump surface in nuclear containment studies: CFD and LP codes validation on TOSQAN tests

Energy Technology Data Exchange (ETDEWEB)

Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France); Degrees du Lou, O. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France); Arts et Métiers ParisTech, DynFluid Lab. EA92, 151, boulevard de l’Hôpital, 75013 Paris (France); Gelain, T. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France)

2013-10-15

Highlights: • Simulations of evaporative TOSQAN sump tests are performed. • These tests are under air–steam gas conditions with addition of He, CO{sub 2} and SF{sub 6}. • ASTEC-CPA LP and TONUS-CFD codes with UDF for sump model are used. • Validation of sump models of both codes show good results. • The code–experiment differences are attributed to turbulent gas mixing modeling. -- Abstract: During the course of a severe accident in a Nuclear Power Plant, water can be collected in the sump containment through steam condensation on walls and spray systems activation. The objective of this paper is to present code validation on evaporative sump tests performed on TOSQAN facility. The ASTEC-CPA code is used as a lumped-parameter code and specific user-defined-functions are developed for the TONUS-CFD code. The seven tests are air–steam tests, as well as tests with other non-condensable gases (He, CO{sub 2} and SF{sub 6}) under steady and transient conditions (two depressurization tests). The results show a good agreement between codes and experiments, indicating a good behavior of the sump models in both codes. The sump model developed as User-Defined Functions (UDF) for TONUS is considered as well validated and is ‘ready-to-use’ for all CFD codes in which such UDF can be added. The remaining discrepancies between codes and experiments are caused by turbulent transport and gas mixing, especially in the presence of non-condensable gases other than air, so that code validation on this important topic for hydrogen safety analysis is still recommended.

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)

Validation of models for the analysis of the transient behavior of metallic fast reactor fuel

International Nuclear Information System (INIS)

Kramer, J.M.; Hughes, T.H.; Gruber, E.E.

1989-01-01

The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in U-Pu-Zr metal alloys as a fuel for sodium-cooled fast reactors. Part of the attractiveness of the IFR concept is the improvement in reactor safety margins through inherent features of a metal-fueled LMR core. In order to demonstrate these safety margins it is necessary to have computer codes available to analyze the detailed response of metallic fuel to a wide range of accident initiators. Two of the codes that play a key role in assessing this response are the STARS fission gas behavior code and the FPIN2 fuel pin mechanics code. Verification and validation are two important components in the development of models and computer codes. Verification demonstrates through comparison of calculations with analytical solutions that the methodology and algorithms correctly solve the equations that govern the phenomena being modeled. Validation, on the other hand, demonstrates through comparison with data that the phenomena are being modeled correctly. Both components are necessary in order to have the confidence to extrapolate the calculations to reactor accident conditions. This paper presents the results of recent progress in the validation of models for the analysis of the behavior of metallic fast reactor fuel. 9 refs., 7 figs

Validation of ASTEC v1.0 computer code against FPT2 test

International Nuclear Information System (INIS)

Mladenov, I.; Tusheva, P.; Kalchev, B.; Dimov, D.; Ivanov, I.

2005-01-01

The aim of the work is by various nodalization schemes of the model to investigate the ASTEC v1.0 computer code sensitivity and to validate the code against PHEBUS - FPT2 experiment. This code is used for severe accident analysis. The aim corresponds to the main technical objective of the experiment which is to contribute to the validation of models and computer codes to be used for the calculation of the source term in case of a severe accident in a Light Water Reactor. The objective's scope of the FPT2 is large - separately for the bundle, the experimental circuit and the containment. Additional objectives are to characterize aerosol sizing and deposition processes, and also potential FP poisoning effects on hydrogen recombiner coupons exposed to containment atmospheric conditions representative of a LWR severe accident. The analyses of the results of the performed calculations show a good accordance with the reference case calculations, and then with the experimental data. Some differences in the calculations for the thermal behavior appear locally during the oxidation phase and the heat-up phase. There is very good confirmation regarding the volatile and semi-volatile fission products release from the fuel pellets. Important for analysis of the process is the final axial distribution of the mass of fuel relocation obtained at the end of the calculation

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Validation and verification plan for safety and PRA codes

International Nuclear Information System (INIS)

Ades, M.J.; Crowe, R.D.; Toffer, H.

1991-04-01

This report discusses a verification and validation (V ampersand V) plan for computer codes used for safety analysis and probabilistic risk assessment calculations. The present plan fulfills the commitments by Westinghouse Savannah River Company (WSRC) to the Department of Energy Savannah River Office (DOE-SRO) to bring the essential safety analysis and probabilistic risk assessment codes in compliance with verification and validation requirements

Validation of the reactor dynamics code TRAB

International Nuclear Information System (INIS)

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

1991-05-01

The one-dimensional reactor dynamics code TRAB (Transient Analysis code for BWRs) developed at VTT was originally designed for BWR analyses, but it can in its present version be used for various modelling purposes. The core model of TRAB can be used separately for LWR calculations. For PWR modelling the core model of TRAB has been coupled to circuit model SMABRE to form the SMATRA code. The versatile modelling capabilities of TRAB have been utilized also in analyses of e.g. the heating reactor SECURE and the RBMK-type reactor (Chernobyl). The report summarizes the extensive validation of TRAB. TRAB has been validated with benchmark problems, comparative calculations against independent analyses, analyses of start-up experiments of nuclear power plants and real plant transients. Comparative RBMES type reactor calculations have been made against Soviet simulations and the initial power excursion of the Chernobyl reactor accident has also been calculated with TRAB

A Systematic Method for Verification and Validation of Gyrokinetic Microstability Codes

Energy Technology Data Exchange (ETDEWEB)

Bravenec, Ronald [Fourth State Research, Austin, TX (United States)

2017-11-14

My original proposal for the period Feb. 15, 2014 through Feb. 14, 2017 called for an integrated validation and verification effort carried out by myself with collaborators. The validation component would require experimental profile and power-balance analysis. In addition, it would require running the gyrokinetic codes varying the input profiles within experimental uncertainties to seek agreement with experiment before discounting a code as invalidated. Therefore, validation would require a major increase of effort over my previous grant periods which covered only code verification (code benchmarking). Consequently, I had requested full-time funding. Instead, I am being funded at somewhat less than half time (5 calendar months per year). As a consequence, I decided to forego the validation component and to only continue the verification efforts.

Development of blow down and sodium-water reaction jet analysis codes-Validation by sodium-water reaction tests (SWAT-1R)

International Nuclear Information System (INIS)

Hiroshi Seino; Akikazu Kurihara; Isao Ono; Koji Jitsu

2005-01-01

Blow down analysis code (LEAP-BLOW) and sodium-water reaction jet analysis code (LEAP-JET) have been developed in order to improve the evaluation method on sodium-water reaction event in the steam generator (SG) of a sodium cooled fast breeder reactor (FBR). The validation analyses by these two codes were carried out using the data of Sodium-Water Reaction Test (SWAT-1R). The following main results have been obtained through this validation: (1) The calculational results by LEAP-BLOW such as internal pressure and water flow rate show good agreement with the results of the SWAT- 1R test. (2) The LEAP-JET code can qualitatively simulate the behavior of sodium-water reaction. However, it is found that the code has tendency to overestimate the maximum temperature of the reaction jet. (authors)

Application of advanced validation concepts to oxide fuel performance codes: LIFE-4 fast-reactor and FRAPCON thermal-reactor fuel performance codes

Energy Technology Data Exchange (ETDEWEB)

Unal, C., E-mail: cu@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Williams, B.J. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Yacout, A. [Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Higdon, D.M. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

2013-10-15

of MS/MP capabilities because these advanced tools have not yet reached sufficient maturity to support such an investigation. In an earlier paper (Unal et al., 2011), we proposed a methodology that potentially can be used to address these new challenges in the design and licensing of evolving nuclear technology. The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. An enhanced calibration concept was introduced and is accomplished through data assimilation. Since advanced MS/MP codes have not yet reached the level of maturity required for a comprehensive validation and calibration exercise, we considered two legacy fuel codes and apply parts of our methodology to these codes to demonstrate the benefits of the new calibration capabilities we recently developed as a part of the proposed framework. This effort does not directly support “born-assessed” validation for advanced MS/MP codes, but is useful to gain insight on legacy modeling deficiencies and to guide and develop recommendations on high and low priority directions for development of advanced codes and advanced experiments, so as to maximize the benefits of advanced validation and uncertainty quantification (VU) efforts involving the next generation of MS/MP code capabilities. This paper discusses the application of advanced validation techniques (sensitivity, calibration, and prediction) to nuclear fuel performance codes FRAPCON (Geelhood et al., 2011a,b) and LIFE-4 (Boltax et al., 1990). FRAPCON is used to predict oxide fuel behavior in light water reactors. LIFE-4 was developed in the 1980s to predict oxide fuel behavior in fast reactors. We introduce a sensitivity ranking methodology to narrow down the selected parameters for follow-up sensitivity and calibration analyses. We use screening methods with both codes and discuss the results. The number of selected modeling parameters was 61 for FRAPCON and 69 for LIFE-4. The screening

Improvements, verifications and validations of the BOW code

International Nuclear Information System (INIS)

Yu, S.D.; Tayal, M.; Singh, P.N.

1995-01-01

The BOW code calculates the lateral deflections of a fuel element consisting of sheath and pellets, due to temperature gradients, hydraulic drag and gravity. the fuel element is subjected to restraint from endplates, neighboring fuel elements and the pressure tube. Many new features have been added to the BOW code since its original release in 1985. This paper outlines the major improvements made to the code and verification/validation results. (author)

A proposed framework for computational fluid dynamics code calibration/validation

International Nuclear Information System (INIS)

Oberkampf, W.L.

1993-01-01

The paper reviews the terminology and methodology that have been introduced during the last several years for building confidence n the predictions from Computational Fluid Dynamics (CID) codes. Code validation terminology developed for nuclear reactor analyses and aerospace applications is reviewed and evaluated. Currently used terminology such as ''calibrated code,'' ''validated code,'' and a ''validation experiment'' is discussed along with the shortcomings and criticisms of these terms. A new framework is proposed for building confidence in CFD code predictions that overcomes some of the difficulties of past procedures and delineates the causes of uncertainty in CFD predictions. Building on previous work, new definitions of code verification and calibration are proposed. These definitions provide more specific requirements for the knowledge level of the flow physics involved and the solution accuracy of the given partial differential equations. As part of the proposed framework, categories are also proposed for flow physics research, flow modeling research, and the application of numerical predictions. The contributions of physical experiments, analytical solutions, and other numerical solutions are discussed, showing that each should be designed to achieve a distinctively separate purpose in building confidence in accuracy of CFD predictions. A number of examples are given for each approach to suggest methods for obtaining the highest value for CFD code quality assurance

Automation of RELAP5 input calibration and code validation using genetic algorithm

International Nuclear Information System (INIS)

Phung, Viet-Anh; Kööp, Kaspar; Grishchenko, Dmitry; Vorobyev, Yury; Kudinov, Pavel

2016-01-01

Highlights: • Automated input calibration and code validation using genetic algorithm is presented. • Predictions generally overlap experiments for individual system response quantities (SRQs). • It was not possible to predict simultaneously experimental maximum flow rate and oscillation period. • Simultaneous consideration of multiple SRQs is important for code validation. - Abstract: Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called “user effects”. The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation. In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function. In the code validation, we used maximum flow rate as the

Automation of RELAP5 input calibration and code validation using genetic algorithm

Energy Technology Data Exchange (ETDEWEB)

Phung, Viet-Anh, E-mail: vaphung@kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden); Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden); Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden); Vorobyev, Yury, E-mail: yura3510@gmail.com [National Research Center “Kurchatov Institute”, Kurchatov square 1, Moscow 123182 (Russian Federation); Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden)

2016-04-15

Highlights: • Automated input calibration and code validation using genetic algorithm is presented. • Predictions generally overlap experiments for individual system response quantities (SRQs). • It was not possible to predict simultaneously experimental maximum flow rate and oscillation period. • Simultaneous consideration of multiple SRQs is important for code validation. - Abstract: Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called “user effects”. The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation. In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function. In the code validation, we used maximum flow rate as the

Development of an Auto-Validation Program for MARS Code Assessments

International Nuclear Information System (INIS)

Lee, Young Jin; Chung, Bub Dong

2006-01-01

MARS (Multi-dimensional Analysis of Reactor Safety) code is a best-estimate thermal hydraulic system analysis code developed at KAERI. It is important for a thermal hydraulic computer code to be assessed against theoretical and experimental data to verify and validate the performance and the integrity of the structure, models and correlations of the code. The code assessment efforts for complex thermal hydraulics code such as MARS code can be tedious, time-consuming and require large amount of human intervention in data transfer to see the results in graphic forms. Code developers produce many versions of a code during development and each version need to be verified for integrity. Thus, for MARS code developers, it is desirable to have an automatic way of carrying out the code assessment calculations. In the present work, an Auto-Validation program that carries out the code assessment efforts has been developed. The program uses the user supplied configuration file (with '.vv' extension) which contain commands to read input file, to execute the user selected MARS program, and to generate result graphs. The program can be useful if a same set of code assessments is repeated with different versions of the code. The program is written with the Delphi program language. The program runs under the Microsoft Windows environment

Reliability and validity of a treatment fidelity assessment for motivational interviewing targeting sexual risk behaviors in people living with HIV/AIDS.

Science.gov (United States)

Seng, Elizabeth K; Lovejoy, Travis I

2013-12-01

This study psychometrically evaluates the Motivational Interviewing Treatment Integrity Code (MITI) to assess fidelity to motivational interviewing to reduce sexual risk behaviors in people living with HIV/AIDS. 74 sessions from a pilot randomized controlled trial of motivational interviewing to reduce sexual risk behaviors in people living with HIV were coded with the MITI. Participants reported sexual behavior at baseline, 3-month, and 6-months. Regarding reliability, excellent inter-rater reliability was achieved for measures of behavior frequency across the 12 sessions coded by both coders; global scales demonstrated poor intraclass correlations, but adequate percent agreement. Regarding validity, principle components analyses indicated that a two-factor model accounted for an adequate amount of variance in the data. These factors were associated with decreases in sexual risk behaviors after treatment. The MITI is a reliable and valid measurement of treatment fidelity for motivational interviewing targeting sexual risk behaviors in people living with HIV/AIDS.

Validation of Magnetic Reconstruction Codes for Real-Time Applications

International Nuclear Information System (INIS)

Mazon, D.; Murari, A.; Boulbe, C.; Faugeras, B.; Blum, J.; Svensson, J.; Quilichini, T.; Gelfusa, M.

2010-01-01

The real-time reconstruction of the plasma magnetic equilibrium in a tokamak is a key point to access high-performance regimes. Indeed, the shape of the plasma current density profile is a direct output of the reconstruction and has a leading effect for reaching a steady-state high-performance regime of operation. The challenge is thus to develop real-time methods and algorithms that reconstruct the magnetic equilibrium from the perspective of using these outputs for feedback control purposes. In this paper the validation of the JET real-time equilibrium reconstruction codes using both a Bayesian approach and a full equilibrium solver named Equinox will be detailed, the comparison being performed with the off-line equilibrium code EFIT (equilibrium fitting) or the real-time boundary reconstruction code XLOC (X-point local expansion). In this way a significant database, a methodology, and a strategy for the validation are presented. The validation of the results has been performed using a validated database of 130 JET discharges with a large variety of magnetic configurations. Internal measurements like polarimetry and motional Stark effect have been also used for the Equinox validation including some magnetohydrodynamic signatures for the assessment of the reconstructed safety profile and current density. (authors)

Error floor behavior study of LDPC codes for concatenated codes design

Science.gov (United States)

Chen, Weigang; Yin, Liuguo; Lu, Jianhua

2007-11-01

Error floor behavior of low-density parity-check (LDPC) codes using quantized decoding algorithms is statistically studied with experimental results on a hardware evaluation platform. The results present the distribution of the residual errors after decoding failure and reveal that the number of residual error bits in a codeword is usually very small using quantized sum-product (SP) algorithm. Therefore, LDPC code may serve as the inner code in a concatenated coding system with a high code rate outer code and thus an ultra low error floor can be achieved. This conclusion is also verified by the experimental results.

Validation of the Thermal-Hydraulic Model in the SACAP Code with the ISP Tests

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

In safety viewpoint, the pressure of the containment is the important parameter, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In Korea, there have been an extensive efforts to develop the computer code which can analyze the severe accident behavior of the pressurized water reactor. The development has been done in a modularized manner and SACAP(Severe Accident Containment Analysis Package) code is now under final stage of development. SACAP code adopts LP(Lumped Parameter) model and is applicable to analyze the synthetic behavior of the containment during severe accident occurred by thermal-hydraulic transient, combustible gas burn, direct containment heating by high pressure melt ejection, steam explosion and molten core-concrete interaction. The analyses of a number of ISP(International Standard Problem) experiments were done as a part of the SACAP code V and V(verification and validation). In this paper, the SACAP analysis results for ISP-35 NUPEC and ISP-47 TOSQAN are presented including comparison with other existing NPP simulation codes. In this paper, we selected and analyzed ISP-35 NUPEC, ISP-47 TOSQAN in order to confirm the computational performance of SACAP code currently under development. Now the multi-node analysis for the ISP-47 is under process. As a result of simulation, SACAP predicts well the thermal-hydraulic variables such as temperature, pressure, etc. Also, we verify that SACAP code is properly equipped to analyze the gas distribution and condensation.

Contributions to the validation of the ASTEC V1 code

International Nuclear Information System (INIS)

Constantin, Marin; Rizoiu, Andrei; Turcu, Ilie

2004-01-01

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

Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

International Nuclear Information System (INIS)

Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha

2016-01-01

The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

Energy Technology Data Exchange (ETDEWEB)

Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

Analysis of NSPP experiment with ART code for analyzing transport behavior of Aerosol and radionuclides

International Nuclear Information System (INIS)

Ishigami, Tsutomu; Kobayashi, Kensuke; Kajimoto, Mitsuhiro.

1989-01-01

The ART code calculates transport behavior of aerosols and radionuclides during core meltdown accidents in the light water reactors. Since aerosols play an important role in carrying fission products from the core region to the environment, the ART code includes detailed models of aerosol behavior. Aerosols including several radionuclides are classified into many groups according to the aerosol mass. The models of aerosol behavior include agglomeration processes caused by Brownian motion, aerosol settling velocity difference and turbulent flow, and natural deposition processes due to diffusion, thermophoresis, diffusiophoresis, gravitational settling and forced convection. In order to examine validity of the ART models, the NSPP aerosol experiment was analyzed. The ART calculated results showed good agreement with the experimental data. It was ascertained that aerosol growth due to agglomeration, gravitational settling, thermophoresis in an air atmosphere, and diffusiophoresis in an air-steam atmosphere were important physical phenomena in the aerosol behavior. (author)

Validation of the transportation computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND

International Nuclear Information System (INIS)

Maheras, S.J.; Pippen, H.K.

1995-05-01

The computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND were used to estimate radiation doses from the transportation of radioactive material in the Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement. HIGHWAY and INTERLINE were used to estimate transportation routes for truck and rail shipments, respectively. RADTRAN 4 was used to estimate collective doses from incident-free transportation and the risk (probability x consequence) from transportation accidents. RISKIND was used to estimate incident-free radiation doses for maximally exposed individuals and the consequences from reasonably foreseeable transportation accidents. The purpose of this analysis is to validate the estimates made by these computer codes; critiques of the conceptual models used in RADTRAN 4 are also discussed. Validation is defined as ''the test and evaluation of the completed software to ensure compliance with software requirements.'' In this analysis, validation means that the differences between the estimates generated by these codes and independent observations are small (i.e., within the acceptance criterion established for the validation analysis). In some cases, the independent observations used in the validation were measurements; in other cases, the independent observations used in the validation analysis were generated using hand calculations. The results of the validation analyses performed for HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND show that the differences between the estimates generated using the computer codes and independent observations were small. Based on the acceptance criterion established for the validation analyses, the codes yielded acceptable results; in all cases the estimates met the requirements for successful validation

ClinicalCodes: an online clinical codes repository to improve the validity and reproducibility of research using electronic medical records.

Science.gov (United States)

Springate, David A; Kontopantelis, Evangelos; Ashcroft, Darren M; Olier, Ivan; Parisi, Rosa; Chamapiwa, Edmore; Reeves, David

2014-01-01

Lists of clinical codes are the foundation for research undertaken using electronic medical records (EMRs). If clinical code lists are not available, reviewers are unable to determine the validity of research, full study replication is impossible, researchers are unable to make effective comparisons between studies, and the construction of new code lists is subject to much duplication of effort. Despite this, the publication of clinical codes is rarely if ever a requirement for obtaining grants, validating protocols, or publishing research. In a representative sample of 450 EMR primary research articles indexed on PubMed, we found that only 19 (5.1%) were accompanied by a full set of published clinical codes and 32 (8.6%) stated that code lists were available on request. To help address these problems, we have built an online repository where researchers using EMRs can upload and download lists of clinical codes. The repository will enable clinical researchers to better validate EMR studies, build on previous code lists and compare disease definitions across studies. It will also assist health informaticians in replicating database studies, tracking changes in disease definitions or clinical coding practice through time and sharing clinical code information across platforms and data sources as research objects.

Phase 1 Validation Testing and Simulation for the WEC-Sim Open Source Code

Science.gov (United States)

Ruehl, K.; Michelen, C.; Gunawan, B.; Bosma, B.; Simmons, A.; Lomonaco, P.

2015-12-01

WEC-Sim is an open source code to model wave energy converters performance in operational waves, developed by Sandia and NREL and funded by the US DOE. The code is a time-domain modeling tool developed in MATLAB/SIMULINK using the multibody dynamics solver SimMechanics, and solves the WEC's governing equations of motion using the Cummins time-domain impulse response formulation in 6 degrees of freedom. The WEC-Sim code has undergone verification through code-to-code comparisons; however validation of the code has been limited to publicly available experimental data sets. While these data sets provide preliminary code validation, the experimental tests were not explicitly designed for code validation, and as a result are limited in their ability to validate the full functionality of the WEC-Sim code. Therefore, dedicated physical model tests for WEC-Sim validation have been performed. This presentation provides an overview of the WEC-Sim validation experimental wave tank tests performed at the Oregon State University's Directional Wave Basin at Hinsdale Wave Research Laboratory. Phase 1 of experimental testing was focused on device characterization and completed in Fall 2015. Phase 2 is focused on WEC performance and scheduled for Winter 2015/2016. These experimental tests were designed explicitly to validate the performance of WEC-Sim code, and its new feature additions. Upon completion, the WEC-Sim validation data set will be made publicly available to the wave energy community. For the physical model test, a controllable model of a floating wave energy converter has been designed and constructed. The instrumentation includes state-of-the-art devices to measure pressure fields, motions in 6 DOF, multi-axial load cells, torque transducers, position transducers, and encoders. The model also incorporates a fully programmable Power-Take-Off system which can be used to generate or absorb wave energy. Numerical simulations of the experiments using WEC-Sim will be

Validation analysis of pool fire experiment (Run-F7) using SPHINCS code

International Nuclear Information System (INIS)

Yamaguchi, Akira; Tajima, Yuji

1998-04-01

SPHINCS (Sodium Fire Phenomenology IN multi-Cell System) code has been developed for the safety analysis of sodium fire accident in a Fast Breeder Reactor. The main features of the SPHINCS code with respect to the sodium pool fire phenomena are multi-dimensional modeling of the thermal behavior in sodium pool and steel liner, modeling of the extension of sodium pool area based on the sodium mass conservation, and equilibrium model for the chemical reaction of pool fire on the flame sheet at the surface of sodium pool during. Therefore, the SPHINCS code is capable of temperature evaluation of the steel liner in detail during the small and/or medium scale sodium leakage accidents. In this study, Run-F7 experiment in which the sodium leakage rate is 11.8 kg/hour has been analyzed. In the experiment the diameter of the sodium pool is approximately 60 cm and the maximum steel liner temperature was 616 degree C. The analytical results tell us the agreement between the SPHINCS analysis and the experiment is excellent with respect to the time history and spatial distribution of the liner temperature, sodium pool extension behavior, as well as atmosphere gas temperature. It is concluded that the pool fire modeling of the SPHINCS code has been validated for this experiment. The SPHINCS code is currently applicable to the sodium pool fire phenomena and the temperature evaluation of the steel liner. The experiment series are continued to check some parameters, i.e., sodium leakage rate and the height of sodium leakage. Thus, the author will analyze the subsequent experiments to check the influence of the parameters and applies SPHINCS to the sodium fire consequence analysis of fast reactor. (author)

Validation of the Classroom Behavior Inventory

Science.gov (United States)

Blunden, Dale; And Others

1974-01-01

Factor-analytic methods were used toassess contruct validity of the Classroom Behavior Inventory, a scale for rating behaviors associated with hyperactivity. The Classroom Behavior Inventory measures three dimensions of behavior: Hyperactivity, Hostility, and Sociability. Significant concurrent validity was obtained for only one Classroom Behavior…

WSRC approach to validation of criticality safety computer codes

International Nuclear Information System (INIS)

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

1991-01-01

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

COCOSYS: Status of development and validation of the German containment code system

International Nuclear Information System (INIS)

Allelein, H.-J.; Arndt, S.; Klein-Hessling, W.; Schwarz, S.; Spengler, C.; Weber, G.

2006-01-01

For the simulation of severe accident propagation in containments of nuclear power plants it is necessary to assess the efficiency of a severe accident measures under conditions as realistic as possible. Therefore the German containment code system COCOSYS is under development and validation at GRS. The main objective is to provide a code system on the basis of mostly mechanistic models for the comprehensive simulation of all relevant processes and plant states during severe accidents in the containment of light water reactors covering the design basis accidents, too. COCOSYS is being used for the identification of possible deficits in plant safety, qualification of the safety reserves of the entire system, assessment of damage-limiting or mitigating accident management measures, support of integral codes in PSA level 2 studies and safety evaluation of new plants. COCOSYS is composed for three main modules, which are separate executable files. The communication is realized via PVM (parallel virtual machine). The thermal hydraulic main module (THY) contains several specific models relevant for the simulation of severe accidents. Beside the usual capabilities to calculate the gas distribution and thermal behavior inside the containment, there are special models for the simulation of Hydrogen deflagration, pressure suppression systems etc. Further detailed models exist for the simulation of safety systems, like catalytic recombiners (PAR's), safety relief valves (used in WWR-440/V-230 type plants), ice condenser model, pump and spray system models for the complete simulation of cooling systems. The aerosol and fission product part (AFP) describes the aerosol behavior of nonsoluble and as well as hygroscopic aerosols, iodine chemistry and fission transport. Further the decay process of nuclides is considered using ORIGIN like routines. The corium concrete interaction (CCI) main module is based on an improved version of WECHSL extended by the ChemApp module for the

Development and validation of computer codes for analysis of PHWR containment behaviour

International Nuclear Information System (INIS)

Markandeya, S.G.; Haware, S.K.; Ghosh, A.K.; Venkat Raj, V.

1997-01-01

In order to ensure that the design intent of the containment of Indian Pressurised Heavy Water Reactors (IPHWRs) is met, both analytical and experimental studies are being pursued at BARC. As a part of analytical studies, computer codes for predicting the behaviour of containment under various accident scenarios are developed/adapted. These include codes for predicting 1) pressure, temperature transients in the containment following either Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB), 2) hydrogen behaviour in respect of its distribution, combustion and the performance of proposed mitigation systems, and 3) behaviour of fission product aerosols in the piping circuits of the primary heat transport system and in the containment. All these codes have undergone thorough validation using data obtained from in-house test facilities or from international sources. Participation in the International Standard Problem (ISP) exercises has also helped in validation of the codes. The present paper briefly describes some of these codes and the various exercises performed for their validation. (author)

Descriptive analysis of the verbal behavior of a therapist: a known-group validity analysis of the putative behavioral functions involved in clinical interaction.

Science.gov (United States)

Virues-Ortega, Javier; Montaño-Fidalgo, Montserrat; Froján-Parga, María Xesús; Calero-Elvira, Ana

2011-12-01

This study analyzes the interobserver agreement and hypothesis-based known-group validity of the Therapist's Verbal Behavior Category System (SISC-INTER). The SISC-INTER is a behavioral observation protocol comprised of a set of verbal categories representing putative behavioral functions of the in-session verbal behavior of a therapist (e.g., discriminative, reinforcing, punishing, and motivational operations). The complete therapeutic process of a clinical case of an individual with marital problems was recorded (10 sessions, 8 hours), and data were arranged in a temporal sequence using 10-min periods. Hypotheses based on the expected performance of the putative behavioral functions portrayed by the SISC-INTER codes across prevalent clinical activities (i.e., assessing, explaining, Socratic method, providing clinical guidance) were tested using autoregressive integrated moving average (ARIMA) models. Known-group validity analyses provided support to all hypotheses. The SISC-INTER may be a useful tool to describe therapist-client interaction in operant terms. The utility of reliable and valid protocols for the descriptive analysis of clinical practice in terms of verbal behavior is discussed. Copyright © 2011. Published by Elsevier Ltd.

Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

International Nuclear Information System (INIS)

Wren, D.J.; Popov, N.; Snell, V.G.

2004-01-01

Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

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.

Validation of containment thermal hydraulic computer codes for VVER reactor

Energy Technology Data Exchange (ETDEWEB)

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

2005-07-01

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

Verification and validation of the THYTAN code for the graphite oxidation analysis in the HTGR systems

International Nuclear Information System (INIS)

Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi

2014-12-01

The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V and V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement. (author)

Lawrence Livermore National Laboratory Probabilistic Seismic Hazard Codes Validation

International Nuclear Information System (INIS)

Savy, J B

2003-01-01

Probabilistic Seismic Hazard Analysis (PSHA) is a methodology that estimates the likelihood that various levels of earthquake-caused ground motion will be exceeded at a given location in a given future time-period. LLNL has been developing the methodology and codes in support of the Nuclear Regulatory Commission (NRC) needs for reviews of site licensing of nuclear power plants, since 1978. A number of existing computer codes have been validated and still can lead to ranges of hazard estimates in some cases. Until now, the seismic hazard community had not agreed on any specific method for evaluation of these codes. The Earthquake Engineering Research Institute (EERI) and the Pacific Engineering Earthquake Research (PEER) center organized an exercise in testing of existing codes with the aim of developing a series of standard tests that future developers could use to evaluate and calibrate their own codes. Seven code developers participated in the exercise, on a voluntary basis. Lawrence Livermore National laboratory participated with some support from the NRC. The final product of the study will include a series of criteria for judging of the validity of the results provided by a computer code. This EERI/PEER project was first planned to be completed by June of 2003. As the group neared completion of the tests, the managing team decided that new tests were necessary. As a result, the present report documents only the work performed to this point. It demonstrates that the computer codes developed by LLNL perform all calculations correctly and as intended. Differences exist between the results of the codes tested, that are attributed to a series of assumptions, on the parameters and models, that the developers had to make. The managing team is planning a new series of tests to help in reaching a consensus on these assumptions

Toward a CFD nose-to-tail capability - Hypersonic unsteady Navier-Stokes code validation

Science.gov (United States)

Edwards, Thomas A.; Flores, Jolen

1989-01-01

Computational fluid dynamics (CFD) research for hypersonic flows presents new problems in code validation because of the added complexity of the physical models. This paper surveys code validation procedures applicable to hypersonic flow models that include real gas effects. The current status of hypersonic CFD flow analysis is assessed with the Compressible Navier-Stokes (CNS) code as a case study. The methods of code validation discussed to beyond comparison with experimental data to include comparisons with other codes and formulations, component analyses, and estimation of numerical errors. Current results indicate that predicting hypersonic flows of perfect gases and equilibrium air are well in hand. Pressure, shock location, and integrated quantities are relatively easy to predict accurately, while surface quantities such as heat transfer are more sensitive to the solution procedure. Modeling transition to turbulence needs refinement, though preliminary results are promising.

The Validation of Macro and Micro Observations of Parent-Child Dynamics Using the Relationship Affect Coding System in Early Childhood.

Science.gov (United States)

Dishion, Thomas J; Mun, Chung Jung; Tein, Jenn-Yun; Kim, Hanjoe; Shaw, Daniel S; Gardner, Frances; Wilson, Melvin N; Peterson, Jenene

2017-04-01

This study examined the validity of micro social observations and macro ratings of parent-child interaction in early to middle childhood. Seven hundred and thirty-one families representing multiple ethnic groups were recruited and screened as at risk in the context of Women, Infant, and Children (WIC) Nutritional Supplement service settings. Families were randomly assigned to the Family Checkup (FCU) intervention or the control condition at age 2 and videotaped in structured interactions in the home at ages 2, 3, 4, and 5. Parent-child interaction videotapes were micro-coded using the Relationship Affect Coding System (RACS) that captures the duration of two mutual dyadic states: positive engagement and coercion. Macro ratings of parenting skills were collected after coding the videotapes to assess parent use of positive behavior support and limit setting skills (or lack thereof). Confirmatory factor analyses revealed that the measurement model of macro ratings of limit setting and positive behavior support was not supported by the data, and thus, were excluded from further analyses. However, there was moderate stability in the families' micro social dynamics across early childhood and it showed significant improvements as a function of random assignment to the FCU. Moreover, parent-child dynamics were predictive of chronic behavior problems as rated by parents in middle childhood, but not emotional problems. We conclude with a discussion of the validity of the RACS and on methodological advantages of micro social coding over the statistical limitations of macro rating observations. Future directions are discussed for observation research in prevention science.

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

Directory of Open Access Journals (Sweden)

Y. M. Tumanov

2010-03-01

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

Validation of computer codes used in the safety analysis of Canadian research reactors

International Nuclear Information System (INIS)

Bishop, W.E.; Lee, A.G.

1998-01-01

AECL has embarked on a validation program for the suite of computer codes that it uses in performing the safety analyses for its research reactors. Current focus is on codes used for the analysis of the two MAPLE reactors under construction at Chalk River but the program will be extended to include additional codes that will be used for the Irradiation Research Facility. The program structure is similar to that used for the validation of codes used in the safety analyses for CANDU power reactors. (author)

Validation matrix for the assessment of thermal-hydraulic codes for VVER LOCA and transients. A report by the OECD support group on the VVER thermal-hydraulic code validation matrix

International Nuclear Information System (INIS)

2001-06-01

This report deals with an internationally agreed experimental test facility matrix for the validation of best estimate thermal-hydraulic computer codes applied for the analysis of VVER reactor primary systems in accident and transient conditions. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities that supplement the CSNI CCVMs and are suitable for reproducing these aspects are selected. Secondly, a list of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. The construction of VVER Thermal-Hydraulic Code Validation Matrix follows the logic of the CSNI Code Validation Matrices (CCVM). Similar to the CCVM it is an attempt to collect together in a systematic way the best sets of available test data for VVER specific code validation, assessment and improvement, including quantitative assessment of uncertainties in the modelling of phenomena by the codes. In addition to this objective, it is an attempt to record information which has been generated in countries operating VVER reactors over the last 20 years so that it is more accessible to present and future workers in that field than would otherwise be the case. (authors)

CFD Code Validation against Stratified Air-Water Flow Experimental Data

International Nuclear Information System (INIS)

Terzuoli, F.; Galassi, M.C.; Mazzini, D.; D'Auria, F.

2008-01-01

Pressurized thermal shock (PTS) modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV) lifetime is the cold water emergency core cooling (ECC) injection into the cold leg during a loss of coolant accident (LOCA). Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM) Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs) code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mecanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX), and a research code (NEPTUNE CFD). The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling

CFD Code Validation against Stratified Air-Water Flow Experimental Data

Directory of Open Access Journals (Sweden)

F. Terzuoli

2008-01-01

Full Text Available Pressurized thermal shock (PTS modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV lifetime is the cold water emergency core cooling (ECC injection into the cold leg during a loss of coolant accident (LOCA. Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mécanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX, and a research code (NEPTUNE CFD. The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling.

Development and Verification of Behavior of Tritium Analytic Code (BOTANIC)

International Nuclear Information System (INIS)

Park, Min Young; Kim, Eung Soo

2014-01-01

VHTR, one of the Generation IV reactor concepts, has a relatively high operation temperature and is usually suggested as a heat source for many industrial processes, including hydrogen production process. Thus, it is vital to trace tritium behavior in the VHTR system and the potential permeation rate to the industrial process. In other words, tritium is a crucial issue in terms of safety in the fission reactor system. Therefore, it is necessary to understand the behavior of tritium and the development of the tool to enable this is vital.. In this study, a Behavior of Tritium Analytic Code (BOTANIC) an analytic tool which is capable of analyzing tritium behavior is developed using a chemical process code called gPROMS. BOTANIC was then further verified using the analytic solutions and benchmark codes such as Tritium Permeation Analysis Code (TPAC) and COMSOL. In this study, the Behavior of Tritium Analytic Code, BOTANIC, has been developed using a chemical process code called gPROMS. The code has several distinctive features including non-diluted assumption, flexible applications and adoption of distributed permeation model. Due to these features, BOTANIC has the capability to analyze a wide range of tritium level systems and has a higher accuracy as it has the capacity to solve distributed models. BOTANIC was successfully developed and verified using analytical solution and the benchmark code calculation result. The results showed very good agreement with the analytical solutions and the calculation results of TPAC and COMSOL. Future work will be focused on the total system verification

Experimental validation of the HARMONIE code

International Nuclear Information System (INIS)

Bernard, A.; Dorsselaere, J.P. van

1984-01-01

An experimental program of deformation, in air, of different groups of subassemblies (7 to 41 subassemblies), was performed on a scale 1 mock-up in the SPX1 geometry, in order to achieve a first experimental validation of the code HARMONIE. The agreement between tests and calculations was suitable, qualitatively for all the groups and quantitatively for regular groups of 19 subassemblies at most. The differences come mainly from friction between pads, and secondly from the foot gaps. (author)

Study of bifurcation behavior of two-dimensional turbo product code decoders

International Nuclear Information System (INIS)

He Yejun; Lau, Francis C.M.; Tse, Chi K.

2008-01-01

Turbo codes, low-density parity-check (LDPC) codes and turbo product codes (TPCs) are high performance error-correction codes which employ iterative algorithms for decoding. Under different conditions, the behaviors of the decoders are different. While the nonlinear dynamical behaviors of turbo code decoders and LDPC decoders have been reported in the literature, the dynamical behavior of TPC decoders is relatively unexplored. In this paper, we investigate the behavior of the iterative algorithm of a two-dimensional TPC decoder when the input signal-to-noise ratio (SNR) varies. The quantity to be measured is the mean square value of the posterior probabilities of the information bits. Unlike turbo decoders or LDPC decoders, TPC decoders do not produce a clear 'waterfall region'. This is mainly because the TPC decoding algorithm does not converge to 'indecisive' fixed points even at very low SNR values

Verification and Validation of Heat Transfer Model of AGREE Code

Energy Technology Data Exchange (ETDEWEB)

Tak, N. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seker, V.; Drzewiecki, T. J.; Downar, T. J. [Department of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Michigan (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington (United States)

2013-05-15

The AGREE code was originally developed as a multi physics simulation code to perform design and safety analysis of Pebble Bed Reactors (PBR). Currently, additional capability for the analysis of Prismatic Modular Reactor (PMR) core is in progress. Newly implemented fluid model for a PMR core is based on a subchannel approach which has been widely used in the analyses of light water reactor (LWR) cores. A hexagonal fuel (or graphite block) is discretized into triangular prism nodes having effective conductivities. Then, a meso-scale heat transfer model is applied to the unit cell geometry of a prismatic fuel block. Both unit cell geometries of multi-hole and pin-in-hole types of prismatic fuel blocks are considered in AGREE. The main objective of this work is to verify and validate the heat transfer model newly implemented for a PMR core in the AGREE code. The measured data in the HENDEL experiment were used for the validation of the heat transfer model for a pin-in-hole fuel block. However, the HENDEL tests were limited to only steady-state conditions of pin-in-hole fuel blocks. There exist no available experimental data regarding a heat transfer in multi-hole fuel blocks. Therefore, numerical benchmarks using conceptual problems are considered to verify the heat transfer model of AGREE for multi-hole fuel blocks as well as transient conditions. The CORONA and GAMMA+ codes were used to compare the numerical results. In this work, the verification and validation study were performed for the heat transfer model of the AGREE code using the HENDEL experiment and the numerical benchmarks of selected conceptual problems. The results of the present work show that the heat transfer model of AGREE is accurate and reliable for prismatic fuel blocks. Further validation of AGREE is in progress for a whole reactor problem using the HTTR safety test data such as control rod withdrawal tests and loss-of-forced convection tests.

Use of operational data for the validation of the SOPHT thermal-hydraulic code

Energy Technology Data Exchange (ETDEWEB)

Ho, S F; Martin, G; Shoukas, L; Siddiqui, Z; Phillips, B [Ontario Hydro, Bowmanville, ON (Canada). Darlington Nuclear Generating Station

1996-12-31

The primary objective of this paper is to describe the validation process of the SOPHT and MINI-SOPHT codes with the use of reactor operational data. The secondary objective is to illustrative the effectiveness of the code as a performance monitoring tool by discussing the discoveries that were made during the validation process. (author). 2 refs.

A proposed methodology for computational fluid dynamics code verification, calibration, and validation

Science.gov (United States)

Aeschliman, D. P.; Oberkampf, W. L.; Blottner, F. G.

Verification, calibration, and validation (VCV) of Computational Fluid Dynamics (CFD) codes is an essential element of the code development process. The exact manner in which code VCV activities are planned and conducted, however, is critically important. It is suggested that the way in which code validation, in particular, is often conducted--by comparison to published experimental data obtained for other purposes--is in general difficult and unsatisfactory, and that a different approach is required. This paper describes a proposed methodology for CFD code VCV that meets the technical requirements and is philosophically consistent with code development needs. The proposed methodology stresses teamwork and cooperation between code developers and experimentalists throughout the VCV process, and takes advantage of certain synergisms between CFD and experiment. A novel approach to uncertainty analysis is described which can both distinguish between and quantify various types of experimental error, and whose attributes are used to help define an appropriate experimental design for code VCV experiments. The methodology is demonstrated with an example of laminar, hypersonic, near perfect gas, 3-dimensional flow over a sliced sphere/cone of varying geometrical complexity.

Validation of the TAC/BLOOST code (Contract research)

International Nuclear Information System (INIS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

2005-06-01

Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30% (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test. (author)

Validity of an observation method for assessing pain behavior in individuals with multiple sclerosis.

Science.gov (United States)

Cook, Karon F; Roddey, Toni S; Bamer, Alyssa M; Amtmann, Dagmar; Keefe, Francis J

2013-09-01

Pain is a common and complex experience for individuals who live with multiple sclerosis (MS) and it interferes with physical, psychological, and social function. A valid and reliable tool for quantifying observed pain behaviors in MS is critical to understand how pain behaviors contribute to pain-related disability in this clinical population. To evaluate the reliability and validity of a pain behavioral observation protocol in individuals who have MS. Community-dwelling volunteers with MS (N=30), back pain (N=5), or arthritis (N=8) were recruited based on clinician referrals, advertisements, fliers, web postings, and participation in previous research. Participants completed the measures of pain severity, pain interference, and self-reported pain behaviors and were videotaped doing typical activities (e.g., walking and sitting). Two coders independently recorded frequencies of pain behaviors by category (e.g., guarding and bracing) and interrater reliability statistics were calculated. Naïve observers reviewed videotapes of individuals with MS and rated their pain. The Spearman's correlations were calculated between pain behavior frequencies and self-reported pain and pain ratings by naïve observers. Interrater reliability estimates indicated the reliability of pain codes in the MS sample. Kappa coefficients ranged from moderate (sighing=0.40) to substantial agreements (guarding=0.83). These values were comparable with those obtained in the combined back pain and arthritis sample. Concurrent validity was supported by correlations with self-reported pain (0.46-0.53) and with self-reports of pain behaviors (0.58). Construct validity was supported by a finding of 0.87 correlation between total pain behaviors observed by coders and mean pain ratings by naïve observers. Results support the use of the pain behavior observation protocol for assessing pain behaviors of individuals with MS. Valid assessments of pain behaviors of individuals with MS could lead to

Study of bifurcation behavior of two-dimensional turbo product code decoders

Energy Technology Data Exchange (ETDEWEB)

He Yejun [Department of Electronic and Information Engineering, Hong Kong Polytechnic University, Hunghom, Hong Kong (China); Lau, Francis C.M. [Department of Electronic and Information Engineering, Hong Kong Polytechnic University, Hunghom, Hong Kong (China)], E-mail: encmlau@polyu.edu.hk; Tse, Chi K. [Department of Electronic and Information Engineering, Hong Kong Polytechnic University, Hunghom, Hong Kong (China)

2008-04-15

Turbo codes, low-density parity-check (LDPC) codes and turbo product codes (TPCs) are high performance error-correction codes which employ iterative algorithms for decoding. Under different conditions, the behaviors of the decoders are different. While the nonlinear dynamical behaviors of turbo code decoders and LDPC decoders have been reported in the literature, the dynamical behavior of TPC decoders is relatively unexplored. In this paper, we investigate the behavior of the iterative algorithm of a two-dimensional TPC decoder when the input signal-to-noise ratio (SNR) varies. The quantity to be measured is the mean square value of the posterior probabilities of the information bits. Unlike turbo decoders or LDPC decoders, TPC decoders do not produce a clear 'waterfall region'. This is mainly because the TPC decoding algorithm does not converge to 'indecisive' fixed points even at very low SNR values.

Pre-engineering Spaceflight Validation of Environmental Models and the 2005 HZETRN Simulation Code

Science.gov (United States)

Nealy, John E.; Cucinotta, Francis A.; Wilson, John W.; Badavi, Francis F.; Dachev, Ts. P.; Tomov, B. T.; Walker, Steven A.; DeAngelis, Giovanni; Blattnig, Steve R.; Atwell, William

2006-01-01

The HZETRN code has been identified by NASA for engineering design in the next phase of space exploration highlighting a return to the Moon in preparation for a Mars mission. In response, a new series of algorithms beginning with 2005 HZETRN, will be issued by correcting some prior limitations and improving control of propagated errors along with established code verification processes. Code validation processes will use new/improved low Earth orbit (LEO) environmental models with a recently improved International Space Station (ISS) shield model to validate computational models and procedures using measured data aboard ISS. These validated models will provide a basis for flight-testing the designs of future space vehicles and systems of the Constellation program in the LEO environment.

Developing and modifying behavioral coding schemes in pediatric psychology: a practical guide.

Science.gov (United States)

Chorney, Jill MacLaren; McMurtry, C Meghan; Chambers, Christine T; Bakeman, Roger

2015-01-01

To provide a concise and practical guide to the development, modification, and use of behavioral coding schemes for observational data in pediatric psychology. This article provides a review of relevant literature and experience in developing and refining behavioral coding schemes. A step-by-step guide to developing and/or modifying behavioral coding schemes is provided. Major steps include refining a research question, developing or refining the coding manual, piloting and refining the coding manual, and implementing the coding scheme. Major tasks within each step are discussed, and pediatric psychology examples are provided throughout. Behavioral coding can be a complex and time-intensive process, but the approach is invaluable in allowing researchers to address clinically relevant research questions in ways that would not otherwise be possible. © The Author 2014. Published by Oxford University Press on behalf of the Society of Pediatric Psychology. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Verification and Validation of the Tritium Transport Code TMAP7

International Nuclear Information System (INIS)

Longhurst, Glen R.; Ambrosek, James

2005-01-01

The TMAP code has been upgraded to version 7, which includes radioactive decay along with many features implemented in prior versions. Pursuant to acceptance and release for distribution, the code was exercised in a variety of problem types to demonstrate that it provides results in agreement with theoretical results for cases where those are available. It has also been used to model certain experimental results. In this paper, the capabilities of the TMAP7 code are demonstrated by presenting some of the results from the verification and validation process

Validation of thermal hydraulic computer codes for advanced light water reactor

International Nuclear Information System (INIS)

Macek, J.

2001-01-01

The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

Preliminary Validation of the MATRA-LMR Code Using Existing Sodium-Cooled Experimental Data

International Nuclear Information System (INIS)

Choi, Sun Rock; Kim, Sangji

2014-01-01

The main objective of the SFR prototype plant is to verify TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. The fuel design limit is highly dependent on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, an accurate temperature calculation in each subassembly is highly important to assure a safe and reliable operation of the reactor systems. The current core thermalhydraulic design is mainly performed using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code, which has been already validated using the existing sodium-cooled experimental data. In addition to the SLTHEN code, a detailed analysis is performed using the MATRA-LMR (Multichannel Analyzer for Transient and steady-state in Rod Array-Liquid Metal Reactor) code. In this work, the MATRA-LMR code is validated for a single subassembly evaluation using the previous experimental data. The MATRA-LMR code has been validated using existing sodium-cooled experimental data. The results demonstrate that the design code appropriately predicts the temperature distributions compared with the experimental values. Major differences are observed in the experiments with the large pin number due to the radial-wise mixing difference

Measurement of reactivity coefficients for code validation

International Nuclear Information System (INIS)

Nuding, Matthias; Loetsch, Thomas

2005-01-01

In the year 2003 measurements in the cold reactor state have been performed at the NPP KKI 2 in order to validate the codes that are used for reactor core calculations and especially for the proof of the shutdown margin that is produced by calculations only. For full power states code verification is quite easy because the calculations can be compared with different measured values, e.g. with the activation values determined by the aeroball system. For cold reactor states, however the data base is smaller, especially for reactor cores that are quite 'inhomogeneous' and have rather high Pu-fiss-and 235 U-contents. At the same time the cold reactor state is important regarding the shutdown margin. For these reasons the measurements mentioned above have been performed in order to check the accuracy of the codes that are used by the operator and by our organization for many years. Basically, boron concentrations and control rod worths for different configurations have been measured. The results of the calculation show a very good agreement with the measured values. Therefore, it can be stated that the operator's as well as our code system is suitable for routine use, e.g. during licensing procedures (Authors)

Experimental validation of the containment codes ASTARTE and SEURBNUK

International Nuclear Information System (INIS)

Kendall, K.C.; Arnold, L.A.; Broadhouse, B.J.; Jones, A.; Yerkess, A.; Benuzzi, A.

1979-10-01

The fast reactor containment codes ASTARTE and SEURBNUK are being validated against data from the COVA series of small scale experiments being performed jointly by the UKAEA and JRC Ispra. The experimental programme is nearly complete, and data are given. (U.K.)

Development and preliminary validation of flux map processing code MAPLE

International Nuclear Information System (INIS)

Li Wenhuai; Zhang Xiangju; Dang Zhen; Chen Ming'an; Lu Haoliang; Li Jinggang; Wu Yuanbao

2013-01-01

The self-reliant flux map processing code MAPLE was developed by China General Nuclear Power Corporation (CGN). Weight coefficient method (WCM), polynomial expand method (PEM) and thin plane spline (TPS) method were applied to fit the deviation between measured and predicted detector signal results for two-dimensional radial plane, to interpolate or extrapolate the non-instrumented location deviation. Comparison of results in the test cases shows that the TPS method can better capture the information of curved fitting lines than the other methods. The measured flux map data of the Lingao Nuclear Power Plant were processed using MAPLE as validation test cases, combined with SMART code. Validation results show that the calculation results of MAPLE are reasonable and satisfied. (authors)

Gamma streaming experiments for validation of Monte Carlo code

International Nuclear Information System (INIS)

Thilagam, L.; Mohapatra, D.K.; Subbaiah, K.V.; Iliyas Lone, M.; Balasubramaniyan, V.

2012-01-01

In-homogeneities in shield structures lead to considerable amount of leakage radiation (streaming) increasing the radiation levels in accessible areas. Development works on experimental as well as computational methods for quantifying this streaming radiation are still continuing. Monte Carlo based radiation transport code, MCNP is usually a tool for modeling and analyzing such problems involving complex geometries. In order to validate this computational method for streaming analysis, it is necessary to carry out some experimental measurements simulating these inhomogeneities like ducts and voids present in the bulk shields for typical cases. The data thus generated will be analysed by simulating the experimental set up employing MCNP code and optimized input parameters for the code in finding solutions for similar radiation streaming problems will be formulated. Comparison of experimental data obtained from radiation streaming experiments through ducts will give a set of thumb rules and analytical fits for total radiation dose rates within and outside the duct. The present study highlights the validation of MCNP code through the gamma streaming experiments carried out with the ducts of various shapes and dimensions. Over all, the present study throws light on suitability of MCNP code for the analysis of gamma radiation streaming problems for all duct configurations considered. In the present study, only dose rate comparisons have been made. Studies on spectral comparison of streaming radiation are in process. Also, it is planned to repeat the experiments with various shield materials. Since the penetrations and ducts through bulk shields are unavoidable in an operating nuclear facility the results on this kind of radiation streaming simulations and experiments will be very useful in the shield structure optimization without compromising the radiation safety

Validation and application of the system code ATHLET-CD for BWR severe accident analyses

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

2016-10-15

Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

SGN III code conversion from Apollo to HP

International Nuclear Information System (INIS)

Lee, Hae Cho

1996-04-01

SGN III computer code is used to analyze transient behavior of reactor coolant system, pressurizer and steam generators in the event of main steam line break (MSLB), and to calculate mass and energy release for containment design. This report firstly describes detailed work carried out for installation of SFN III on Apollo DN 10000 and code validation results after installation. Secondly, a series of work is also describes in relation to installation of SGN III on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 8 refs. (Author) .new

Validation of system codes RELAP5 and SPECTRA for natural convection boiling in narrow channels

Energy Technology Data Exchange (ETDEWEB)

Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; Slootman, M.L.F.; Wiersema, H.T.

2016-10-15

Highlights: • Computer codes RELAP5/Mod3.3 and SPECTRA 3.61 validated for boiling in narrow channels. • Validated codes can be used for LOCA analyses in research reactors. • Code validation based on natural convection boiling in narrow channels experiments. - Abstract: Safety analyses of LOCA scenarios in nuclear power plants are performed with so called thermal–hydraulic system codes, such as RELAP5. Such codes are validated for typical fuel geometries applied in nuclear power plants. The question considered by this article is if the codes can be applied for LOCA analyses in research reactors, in particular exceeding CHF in very narrow channels. In order to answer this question, validation calculations were performed with two thermal–hydraulic system codes: RELAP and SPECTRA. The validation was based on natural convection boiling in narrow channels experiments, performed by Prof. Monde et al. in the years 1990–2000. In total 42 vertical tube and annulus experiments were simulated with both codes. A good agreement of the calculated values with the measured data was observed. The main conclusions are: • The computer codes RELAP5/Mod 3.3 (US NRC version) and SPECTRA 3.61 have been validated for natural convection boiling in narrow channels using experiments of Monde. The dimensions applied in the experiments were performed for a range that covers the values observed in typical research reactors. Therefore it is concluded that both codes are validated and can be used for LOCA analyses in research reactors, including natural convection boiling. The applicability range of the present validation is: hydraulic diameters of 1.1 ⩽ D{sub hyd} ⩽ 9.0 mm, heated lengths of 0.1 ⩽ L ⩽ 1.0 m, pressures of 0.10 ⩽ P ⩽ 0.99 MPa. In most calculations the burnout was predicted to occur at lower power than that observed in the experiments. In several cases the burnout was observed at higher power. The overprediction was not larger than 16% in RELAP and 15% in

Assessment of US NRC fuel rod behavior codes to extended burnup

International Nuclear Information System (INIS)

Laats, E.T.; Croucher, D.W.; Haggag, F.M.

1982-01-01

The purpose of this paper is to report the status of assessing the capabilities of the NRC fuel rod performance codes for calculating extended burnup rod behavior. As part of this effort, a large spectrum of fuel rod behavior phenomena was examined, and the phenomena deemed as being influential during extended burnup operation were identified. Then, the experiment data base addressing these identified phenomena was examined for availability and completeness at extended burnups. Calculational capabilities of the NRC's steady state FRAPCON-2 and transient FRAP-T6 fuel rod behavior codes were examined for each of the identified phenomenon. Parameters calculated by the codes were compared with the available data base, and judgments were made regarding model performance. Overall, the FRAPCON-2 code was found to be moderately well assessed to extended burnups, but the FRAP-T6 code cannot be adequately assessed until more transient high burnup data are available

System verification and validation report for the TMAD code

International Nuclear Information System (INIS)

Finfrock, S.H.

1995-01-01

This document serves as the Verification and Validation Report for the TMAD code system, which includes the TMAD code and the LIBMAKR Code. The TMAD code was commissioned to facilitate the interpretation of moisture probe measurements in the Hanford Site waste tanks. In principle, the code is an interpolation routine that acts over a library of benchmark data based on two independent variables, typically anomaly size and moisture content. Two additional variables, anomaly type and detector type, can also be considered independent variables, but no interpolation is done over them. The dependent variable is detector response. The intent is to provide the code with measured detector responses from two or more detectors. The code will then interrogate (and interpolate upon) the benchmark data library and find the anomaly-type/anomaly-size/moisture-content combination that provides the closest match to the measured data. The primary purpose of this document is to provide the results of the system testing and the conclusions based thereon. The results of the testing process are documented in the body of the report. Appendix A gives the test plan, including test procedures, used in conducting the tests. Appendix B lists the input data required to conduct the tests, and Appendices C and 0 list the numerical results of the tests

Civil Behavior, Safe-School Planning, and Dress Codes

Science.gov (United States)

Studak, Cathryn M.; Workman, Jane E.

2007-01-01

This research examined news reports in order to identify incidents that precipitated dress code revisions. News reports were examined within the framework of rules for civil behavior. Using key words "school dress codes" and "violence," LEXIS/NEXIS was used to access 104 articles from 44 U.S. newspapers from December 3, 2004 to December 2, 2005.…

Verification and validation of the PLTEMP/ANL code for thermal hydraulic analysis of experimental and test reactors

International Nuclear Information System (INIS)

Kalimullah, M.; Olson, A.O.; Feldman, E.E.; Hanan, N.; Dionne, B.

2012-01-01

The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

Energy Technology Data Exchange (ETDEWEB)

Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-04-07

The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

Development and validation of a nodal code for core calculation

International Nuclear Information System (INIS)

Nowakowski, Pedro Mariano

2004-01-01

The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency

Social Workers and the NASW "Code of Ethics": Belief, Behavior, Disjuncture

Science.gov (United States)

DiFranks, Nikki Nelson

2008-01-01

A quantitative descriptive survey of a national sample of social workers (N = 206) examined discrepancies between belief in the NASW "Code of Ethics" and behavior in implementing the code and social workers' disjunctive distress (disjuncture) when belief and behavior are discordant. Relationships between setting and disjuncture and ethics…

Independent verification and validation testing of the FLASH computer code, Versiion 3.0

International Nuclear Information System (INIS)

Martian, P.; Chung, J.N.

1992-06-01

Independent testing of the FLASH computer code, Version 3.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at various Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Verification tests, and validation tests, were used to determine the operational status of the FLASH computer code. These tests were specifically designed to test: correctness of the FORTRAN coding, computational accuracy, and suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: blind testing, independent applications, and graduated difficulty of test cases. Both quantitative and qualitative testing was performed through evaluating relative root mean square values and graphical comparisons of the numerical, analytical, and experimental data. Four verification test were used to check the computational accuracy and correctness of the FORTRAN coding, and three validation tests were used to check the suitability to simulating actual conditions. These tests cases ranged in complexity from simple 1-D saturated flow to 2-D variably saturated problems. The verification tests showed excellent quantitative agreement between the FLASH results and analytical solutions. The validation tests showed good qualitative agreement with the experimental data. Based on the results of this testing, it was concluded that the FLASH code is a versatile and powerful two-dimensional analysis tool for fluid flow. In conclusion, all aspects of the code that were tested, except for the unit gradient bottom boundary condition, were found to be fully operational and ready for use in hydrological and environmental studies

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

International Nuclear Information System (INIS)

2004-08-01

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

Verification and validation of XSDRNPM code for tank waste calculations

International Nuclear Information System (INIS)

ROGERS, C.A.

1999-01-01

This validation study demonstrates that the XSDRNPM computer code accurately calculates the infinite neutron multiplication for water-moderated systems of low enriched uranium, plutonium, and iron. Calculations are made on a 200 MHz Brvo MS 5200M personal

COAST code conversion from Cyber to HP

International Nuclear Information System (INIS)

Lee, Hae Cho

1996-04-01

The transient thermal hydraulic behavior of reactor coolant system in a nuclear power plant following loss of coolant flow is analyzed by use of COAST digital computer code. COAST calculates individual loop flow rates and steam generator pressure drops is a function of time following coast-down of any number of reactor coolant pumps. This report firstly describes detailed work carried out for installation of COAST on HP 9000/700 series and code validation results after installation. Secondly, a series of work is also describes in relation to installation of COAST on Apollo DN10000 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

Evaporation over sump surface in containment studies: code validation on TOSQAN tests

International Nuclear Information System (INIS)

Malet, J.; Gelain, T.; Degrees du Lou, O.; Daru, V.

2011-01-01

During the course of a severe accident in a Nuclear Power Plant, water can be collected in the sump containment through steam condensation on walls and spray systems activation. The objective of this paper is to present code validation on evaporative sump tests performed on the TOSQAN facility. The ASTEC-CPA code is used as a lumped-parameter code and specific user-defined-functions are developed for the TONUS-CFD code. The tests are air-steam tests, as well as tests with other non-condensable gases (He, CO 2 and SF 6 ) under steady and transient conditions. The results show a good agreement between codes and experiments, indicating a good behaviour of the sump models in both codes. (author)

Development and validation of ALEPH Monte Carlo burn-up code

International Nuclear Information System (INIS)

Stankovskiy, A.; Van den Eynde, G.; Vidmar, T.

2011-01-01

The Monte-Carlo burn-up code ALEPH is being developed in SCK-CEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and 'deterministic' depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH. (authors)

Assessment of heat transfer correlations for supercritical water in the frame of best-estimate code validation

International Nuclear Information System (INIS)

Jaeger, Wadim; Espinoza, Victor H. Sanchez; Schneider, Niko; Hurtado, Antonio

2009-01-01

Within the frame of the Generation IV international forum six innovative reactor concepts are the subject of comprehensive investigations. In some projects supercritical water will be considered as coolant, moderator (as for the High Performance Light Water Reactor) or secondary working fluid (one possible option for Liquid Metal-cooled Fast Reactors). Supercritical water is characterized by a pronounced change of the thermo-physical properties when crossing the pseudo-critical line, which goes hand in hand with a change in the heat transfer (HT) behavior. Hence, it is essential to estimate, in a proper way, the heat-transfer coefficient and subsequently the wall temperature. The scope of this paper is to present and discuss the activities at the Institute for Reactor Safety (IRS) related to the implementation of correlations for wall-to-fluid HT at supercritical conditions in Best-Estimate codes like TRACE as well as its validation. It is important to validate TRACE before applying it to safety analyses of HPLWR or of other reactor systems. In the past 3 decades various experiments have been performed all over the world to reveal the peculiarities of wall-to-fluid HT at supercritical conditions. Several different heat transfer phenomena such as HT enhancement (due to higher Prandtl numbers in the vicinity of the pseudo-critical point) or HT deterioration (due to strong property variations) were observed. Since TRACE is a component based system code with a finite volume method the resolution capabilities are limited and not all physical phenomena can be modeled properly. But Best -Estimate system codes are nowadays the preferred option for safety related investigations of full plants or other integral systems. Thus, the increase of the confidence in such codes is of high priority. In this paper, the post-test analysis of experiments with supercritical parameters will be presented. For that reason various correlations for the HT, which considers the characteristics

The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes

Energy Technology Data Exchange (ETDEWEB)

Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Jeltsov, Marti, E-mail: marti@safety.sci.kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Karbojian, Aram, E-mail: karbojan@kth.se; Villanueva, Walter, E-mail: walter@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

2015-08-15

Highlights: • Design of a heavy liquid thermal-hydraulic loop for CFD/STH code validation. • Description of the loop instrumentation and assessment of measurement error. • Experimental data from forced to natural circulation transient. - Abstract: Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

An approach to verification and validation of MHD codes for fusion applications

Energy Technology Data Exchange (ETDEWEB)

Smolentsev, S., E-mail: sergey@fusion.ucla.edu [University of California, Los Angeles (United States); Badia, S. [Centre Internacional de Mètodes Numèrics en Enginyeria, Barcelona (Spain); Universitat Politècnica de Catalunya – Barcelona Tech (Spain); Bhattacharyay, R. [Institute for Plasma Research, Gandhinagar, Gujarat (India); Bühler, L. [Karlsruhe Institute of Technology (Germany); Chen, L. [University of Chinese Academy of Sciences, Beijing (China); Huang, Q. [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui (China); Jin, H.-G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Krasnov, D. [Technische Universität Ilmenau (Germany); Lee, D.-W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Mas de les Valls, E. [Centre Internacional de Mètodes Numèrics en Enginyeria, Barcelona (Spain); Universitat Politècnica de Catalunya – Barcelona Tech (Spain); Mistrangelo, C. [Karlsruhe Institute of Technology (Germany); Munipalli, R. [HyPerComp, Westlake Village (United States); Ni, M.-J. [University of Chinese Academy of Sciences, Beijing (China); Pashkevich, D. [St. Petersburg State Polytechnical University (Russian Federation); Patel, A. [Universitat Politècnica de Catalunya – Barcelona Tech (Spain); Pulugundla, G. [University of California, Los Angeles (United States); Satyamurthy, P. [Bhabha Atomic Research Center (India); Snegirev, A. [St. Petersburg State Polytechnical University (Russian Federation); Sviridov, V. [Moscow Power Engineering Institute (Russian Federation); Swain, P. [Bhabha Atomic Research Center (India); and others

2015-11-15

Highlights: • Review of status of MHD codes for fusion applications. • Selection of five benchmark problems. • Guidance for verification and validation of MHD codes for fusion applications. - Abstract: We propose a new activity on verification and validation (V&V) of MHD codes presently employed by the fusion community as a predictive capability tool for liquid metal cooling applications, such as liquid metal blankets. The important steps in the development of MHD codes starting from the 1970s are outlined first and then basic MHD codes, which are currently in use by designers of liquid breeder blankets, are reviewed. A benchmark database of five problems has been proposed to cover a wide range of MHD flows from laminar fully developed to turbulent flows, which are of interest for fusion applications: (A) 2D fully developed laminar steady MHD flow, (B) 3D laminar, steady developing MHD flow in a non-uniform magnetic field, (C) quasi-two-dimensional MHD turbulent flow, (D) 3D turbulent MHD flow, and (E) MHD flow with heat transfer (buoyant convection). Finally, we introduce important details of the proposed activities, such as basic V&V rules and schedule. The main goal of the present paper is to help in establishing an efficient V&V framework and to initiate benchmarking among interested parties. The comparison results computed by the codes against analytical solutions and trusted experimental and numerical data as well as code-to-code comparisons will be presented and analyzed in companion paper/papers.

NAUAHYGROS - A code for calculating aerosol behavior in nuclear power plant containments following a severe accident

Energy Technology Data Exchange (ETDEWEB)

Sher, R. [Rudolph Sher Associates, Stanford, CA (United States); Li, J. [Polestar Applied Technology, Inc., Los Altos, CA (United States)

1995-02-01

NAUAHYGROS is a computer code to calculate the behavior of fission product and other aerosol particles in the containment of a nuclear reactor following a severe accident. It is an extension of the German code NAUA, which has been in widespread use for many years. Early versions of NAUA treated various aerosol phenomena in dry atmospheres, including aerosol agglomeration, diffusion (plateout), and settling processes. Later versions added treatments of steam condensation on particles in saturated or supersaturated containment atmospheres. The importance of these condensation effects on aerosol removal rates was demonstrated in large scale simulated containment tests. The additional features incorporated in NAUAHYGROS include principally a treatment of steam condensation on hygroscopic aerosols, which can grow as a result of steam condensation even in superheated atmospheres, and improved modelling of steam condensation on the walls of the containment. The code has been validated against the LACE experiments.

The WINCON programme - validation of fast reactor primary containment codes

International Nuclear Information System (INIS)

Sidoli, J.E.A.; Kendall, K.C.

1988-01-01

In the United Kingdom safety studies for the Commercial Demonstration Fast Reactor (CDFR) include an assessment of the capability of the primary containment in providing an adequate containment for defence against the hazards resulting from a hypothetical Whole Core Accident (WCA). The assessment is based on calculational estimates using computer codes supported by measured evidence from small-scale experiments. The hydrodynamic containment code SEURBNUK-EURDYN is capable of representing a prescribed energy release, the sodium coolant and cover gas, and the main containment and safety related internal structures. Containment loadings estimated using SEURBNUK-EURDYN are used in the structural dynamic code EURDYN-03 for the prediction of the containment response. The experiments serve two purposes, they demonstrate the response of the CDFR containment to accident loadings and provide data for the validation of the codes. This paper summarises the recently completed WINfrith CONtainment (WINCON) experiments that studied the response of specific features of current CDFR design options to WCA loadings. The codes have been applied to some of the experiments and a satisfactory prediction of the global response of the model containment is obtained. This provides confidence in the use of the codes in reactor assessments. (author)

The Initial Atmospheric Transport (IAT) Code: Description and Validation

Energy Technology Data Exchange (ETDEWEB)

Morrow, Charles W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bartel, Timothy James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-10-01

The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34D accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.

Further validation of liquid metal MHD code for unstructured grid based on OpenFOAM

Energy Technology Data Exchange (ETDEWEB)

Feng, Jingchao; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; He, Qingyun; Ye, Minyou

2015-11-15

Highlights: • Specific correction scheme has been adopted to revise the calculating result for non-orthogonal meshes. • The developed MHD code based on OpenFOAM platform has been validated by benchmark cases under uniform and non-uniform magnetic field in round and rectangular ducts. • ALEX experimental results have been used to validate the MHD code based on OpenFOAM. - Abstract: In fusion liquid metal blankets, complex geometries involving contractions, expansions, bends, manifolds are very common. The characteristics of liquid metal flow in these geometries are significant. In order to extend the magnetohydrodynamic (MHD) solver developed on OpenFOAM platform to be applied in the complex geometry, the MHD solver based on unstructured meshes has been implemented. The adoption of non-orthogonal correction techniques in the solver makes it possible to process the non-orthogonal meshes in complex geometries. The present paper focused on the validation of the code under critical conditions. An analytical solution benchmark case and two experimental benchmark cases were conducted to validate the code. Benchmark case I is MHD flow in a circular pipe with arbitrary electric conductivity of the walls in a uniform magnetic field. Benchmark cases II and III are experimental cases of 3D laminar steady MHD flow under fringing magnetic field. In all these cases, the numerical results match well with the benchmark cases.

Further validation of liquid metal MHD code for unstructured grid based on OpenFOAM

International Nuclear Information System (INIS)

Feng, Jingchao; Chen, Hongli; He, Qingyun; Ye, Minyou

2015-01-01

Highlights: • Specific correction scheme has been adopted to revise the calculating result for non-orthogonal meshes. • The developed MHD code based on OpenFOAM platform has been validated by benchmark cases under uniform and non-uniform magnetic field in round and rectangular ducts. • ALEX experimental results have been used to validate the MHD code based on OpenFOAM. - Abstract: In fusion liquid metal blankets, complex geometries involving contractions, expansions, bends, manifolds are very common. The characteristics of liquid metal flow in these geometries are significant. In order to extend the magnetohydrodynamic (MHD) solver developed on OpenFOAM platform to be applied in the complex geometry, the MHD solver based on unstructured meshes has been implemented. The adoption of non-orthogonal correction techniques in the solver makes it possible to process the non-orthogonal meshes in complex geometries. The present paper focused on the validation of the code under critical conditions. An analytical solution benchmark case and two experimental benchmark cases were conducted to validate the code. Benchmark case I is MHD flow in a circular pipe with arbitrary electric conductivity of the walls in a uniform magnetic field. Benchmark cases II and III are experimental cases of 3D laminar steady MHD flow under fringing magnetic field. In all these cases, the numerical results match well with the benchmark cases.

Analysis of Quality of Proxy Questions in Health Surveys by Behavior Coding

NARCIS (Netherlands)

Benitez, I.; Padilla, J.L.; Ongena, Yfke

2012-01-01

The aim of this study is to show how to analyze the quality of questions for proxy informants by means of behavior coding. Proxy questions can undermine survey data quality because of the fact that proxies respond to questions on behalf of other people. Behavior coding can improve questions by

Radiant Energy Measurements from a Scaled Jet Engine Axisymmetric Exhaust Nozzle for a Baseline Code Validation Case

Science.gov (United States)

Baumeister, Joseph F.

1994-01-01

A non-flowing, electrically heated test rig was developed to verify computer codes that calculate radiant energy propagation from nozzle geometries that represent aircraft propulsion nozzle systems. Since there are a variety of analysis tools used to evaluate thermal radiation propagation from partially enclosed nozzle surfaces, an experimental benchmark test case was developed for code comparison. This paper briefly describes the nozzle test rig and the developed analytical nozzle geometry used to compare the experimental and predicted thermal radiation results. A major objective of this effort was to make available the experimental results and the analytical model in a format to facilitate conversion to existing computer code formats. For code validation purposes this nozzle geometry represents one validation case for one set of analysis conditions. Since each computer code has advantages and disadvantages based on scope, requirements, and desired accuracy, the usefulness of this single nozzle baseline validation case can be limited for some code comparisons.

Validation of full core geometry model of the NODAL3 code in the PWR transient Benchmark problems

International Nuclear Information System (INIS)

T-M Sembiring; S-Pinem; P-H Liem

2015-01-01

The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16 % occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4 % for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. (author)

Aztheca Code

International Nuclear Information System (INIS)

Quezada G, S.; Espinosa P, G.; Centeno P, J.; Sanchez M, H.

2017-09-01

This paper presents the Aztheca code, which is formed by the mathematical models of neutron kinetics, power generation, heat transfer, core thermo-hydraulics, recirculation systems, dynamic pressure and level models and control system. The Aztheca code is validated with plant data, as well as with predictions from the manufacturer when the reactor operates in a stationary state. On the other hand, to demonstrate that the model is applicable during a transient, an event occurred in a nuclear power plant with a BWR reactor is selected. The plant data are compared with the results obtained with RELAP-5 and the Aztheca model. The results show that both RELAP-5 and the Aztheca code have the ability to adequately predict the behavior of the reactor. (Author)

Testing the Validity of a Cognitive Behavioral Model for Gambling Behavior.

Science.gov (United States)

Raylu, Namrata; Oei, Tian Po S; Loo, Jasmine M Y; Tsai, Jung-Shun

2016-06-01

Currently, cognitive behavioral therapies appear to be one of the most studied treatments for gambling problems and studies show it is effective in treating gambling problems. However, cognitive behavior models have not been widely tested using statistical means. Thus, the aim of this study was to test the validity of the pathways postulated in the cognitive behavioral theory of gambling behavior using structural equation modeling (AMOS 20). Several questionnaires assessing a range of gambling specific variables (e.g., gambling urges, cognitions and behaviors) and gambling correlates (e.g., psychological states, and coping styles) were distributed to 969 participants from the community. Results showed that negative psychological states (i.e., depression, anxiety and stress) only directly predicted gambling behavior, whereas gambling urges predicted gambling behavior directly as well as indirectly via gambling cognitions. Avoidance coping predicted gambling behavior only indirectly via gambling cognitions. Negative psychological states were significantly related to gambling cognitions as well as avoidance coping. In addition, significant gender differences were also found. The results provided confirmation for the validity of the pathways postulated in the cognitive behavioral theory of gambling behavior. It also highlighted the importance of gender differences in conceptualizing gambling behavior.

Large leak sodium-water reaction code SWACS and its validation

International Nuclear Information System (INIS)

Miyake, O.; Shindo, Y.; Hiroi, H.; Tanabe, H.; Sato, M.

1982-01-01

A computer code SWACS for analyzing the large leak accident of an LMFBR steam generators has been developed and validated. Five tests data obtained by SWAT-3 test facility were compared with code results. In each of SWAT-3 tests, a double-ended guillotine rupture of one tube was simulated in a helical coil steam generator model with 1/2.5 scaled test vessel to the prototype SG. The analytical results, including an initial pressure spike, a propagated pressure in a secondary system, and a quasi-steady pressure, indicate that the overall large-leak event could be predicted in reasonably good agreement

Validity of International Classification of Diseases (ICD) coding for dengue infections in hospital discharge records in Malaysia.

Science.gov (United States)

Woon, Yuan-Liang; Lee, Keng-Yee; Mohd Anuar, Siti Fatimah Zahra; Goh, Pik-Pin; Lim, Teck-Onn

2018-04-20

Hospitalization due to dengue illness is an important measure of dengue morbidity. However, limited studies are based on administrative database because the validity of the diagnosis codes is unknown. We validated the International Classification of Diseases, 10th revision (ICD) diagnosis coding for dengue infections in the Malaysian Ministry of Health's (MOH) hospital discharge database. This validation study involves retrospective review of available hospital discharge records and hand-search medical records for years 2010 and 2013. We randomly selected 3219 hospital discharge records coded with dengue and non-dengue infections as their discharge diagnoses from the national hospital discharge database. We then randomly sampled 216 and 144 records for patients with and without codes for dengue respectively, in keeping with their relative frequency in the MOH database, for chart review. The ICD codes for dengue were validated against lab-based diagnostic standard (NS1 or IgM). The ICD-10-CM codes for dengue had a sensitivity of 94%, modest specificity of 83%, positive predictive value of 87% and negative predictive value 92%. These results were stable between 2010 and 2013. However, its specificity decreased substantially when patients manifested with bleeding or low platelet count. The diagnostic performance of the ICD codes for dengue in the MOH's hospital discharge database is adequate for use in health services research on dengue.

Development and Construct Validation of the Mentor Behavior Scale

Science.gov (United States)

Brodeur, Pascale; Larose, Simon; Tarabulsy, George; Feng, Bei; Forget-Dubois, Nadine

2015-01-01

Researchers suggest that certain supportive behaviors of mentors could increase the benefits of school-based mentoring for youth. However, the literature contains few validated instruments to measure these behaviors. In our present study, we aimed to construct and validate a tool to measure the supportive behaviors of mentors participating in…

Nuclear data to support computer code validation

International Nuclear Information System (INIS)

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

1997-04-01

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

San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses - Revision 1

International Nuclear Information System (INIS)

Hermann, O.W.

2000-01-01

The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotopes) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data, usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, was considered to be of sufficient quality for depletion code validation

Validation and uncertainty analysis of the Athlet thermal-hydraulic computer code

International Nuclear Information System (INIS)

Glaeser, H.

1995-01-01

The computer code ATHLET is being developed by GRS as an advanced best-estimate code for the simulation of breaks and transients in Pressurized Water Reactor (PWRs) and Boiling Water Reactor (BWRs) including beyond design basis accidents. A systematic validation of ATHLET is based on a well balanced set of integral and separate effects tests emphasizing the German combined Emergency Core Cooling (ECC) injection system. When using best estimate codes for predictions of reactor plant states during assumed accidents, qualification of the uncertainty in these calculations is highly desirable. A method for uncertainty and sensitivity evaluation has been developed by GRS where the computational effort is independent of the number of uncertain parameters. (author)

Examples of Use of SINBAD Database for Nuclear Data and Code Validation

Science.gov (United States)

Kodeli, Ivan; Žerovnik, Gašper; Milocco, Alberto

2017-09-01

The SINBAD database currently contains compilations and evaluations of over 100 shielding benchmark experiments. The SINBAD database is widely used for code and data validation. Materials covered include: Air, N. O, H2O, Al, Be, Cu, graphite, concrete, Fe, stainless steel, Pb, Li, Ni, Nb, SiC, Na, W, V and mixtures thereof. Over 40 organisations from 14 countries and 2 international organisations have contributed data and work in support of SINBAD. Examples of the use of the database in the scope of different international projects, such as the Working Party on Evaluation Cooperation of the OECD and the European Fusion Programme demonstrate the merit and possible usage of the database for the validation of modern nuclear data evaluations and new computer codes.

Installation of aerosol behavior model into multi-dimensional thermal hydraulic analysis code AQUA

International Nuclear Information System (INIS)

Kisohara, Naoyuki; Yamaguchi, Akira

1997-12-01

The safety analysis of FBR plant system for sodium leak phenomena needs to evaluate the deposition of the aerosol particle to the components in the plant, the chemical reaction of aerosol to humidity in the air and the effect of the combustion heat through aerosol to the structural component. For this purpose, ABC-INTG (Aerosol Behavior in Containment-INTeGrated Version) code has been developed and used until now. This code calculates aerosol behavior in the gas area of uniform temperature and pressure by 1 cell-model. Later, however, more detailed calculation of aerosol behavior requires the installation of aerosol model into multi-cell thermal hydraulic analysis code AQUA. AQUA can calculate the carrier gas flow, temperature and the distribution of the aerosol spatial concentration. On the other hand, ABC-INTG can calculate the generation, deposition to the wall and flower, agglomeration of aerosol particle and figure out the distribution of the aerosol particle size. Thus, the combination of these two codes enables to deal with aerosol model coupling the distribution of the aerosol spatial concentration and that of the aerosol particle size. This report describes aerosol behavior model, how to install the aerosol model to AQUA and new subroutine equipped to the code. Furthermore, the test calculations of the simple structural model were executed by this code, appropriate results were obtained. Thus, this code has prospect to predict aerosol behavior by the introduction of coupling analysis with multi-dimensional gas thermo-dynamics for sodium combustion evaluation. (J.P.N.)

Simulation of ROCOM Experiment using CUPID Code

Energy Technology Data Exchange (ETDEWEB)

Cho, Yun Je; Lee, Jae Ryong; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of)

2016-10-15

KAERI has developed CUPID, which is a three dimensional thermal hydraulics code for the transient analysis of two-phase flows in nuclear reactor components. To validate the capability of CUPID for simulation of multi-dimensional flow mixing behavior, ROCOM (ROssenforf COolant Mixing) test was simulated. ROCOM test has been conducted in the OECD PKL2 Project to investigate in more detail the thermal hydraulic behavior inside the RPV. Thus far, many researchers used the ROCOM data to validate the CFD code capability of thermal mixing behavior. In this study, a hybrid grid was generated using SALOME software and the ROCOM simulation was performed using CUPID. In addition, the effect of turbulence model was also investigated. Test ROCOM 2.1 and 1.2 cases were simulated using the CUPID code. It was shown that CUPID had capabilities to properly simulate the thermal mixing behavior in the case where the cold water is injected asymmetrically. As the result of calculations, it was found that the mixing efficiency in the downcomer and lower plenum was varied according to the turbulence model. In particular, the calculation results showed that the low Reynolds number turbulence model resulted in good agreement with the experimental data. The further works may involve the finer grid generation and the test of other turbulence models.

Design validation of the ITER EC upper launcher according to codes and standards

Energy Technology Data Exchange (ETDEWEB)

Spaeh, Peter, E-mail: peter.spaeh@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Gagliardi, Mario [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); F4E, Fusion for Energy, Joint Undertaking, Barcelona (Spain); Grossetti, Giovanni; Meier, Andreas; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Weinhorst, Bastian [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany)

2015-10-15

Highlights: • A set of applicable codes and standards has been chosen for the ITER EC upper launcher. • For a particular component load combinations, failure modes and stress categorizations have been determined. • The design validation was performed in accordance with the “design by analysis”-approach of the ASME boiler and pressure vessel code section III. - Abstract: The ITER electron cyclotron (EC) upper launcher has passed the CDR (conceptual design review) in 2005 and the PDR (preliminary design review) in 2009 and is in its final design phase now. The final design will be elaborated by the European consortium ECHUL-CA with contributions from several research institutes in Germany, Italy, the Netherlands and Switzerland. Within this consortium KIT is responsible for the design of the structural components (the upper port plug, UPP) and also the design integration of the launcher. As the selection of applicable codes and standards was under discussion for the past decade, the conceptual and the preliminary design of the launcher structure were not elaborated in straight accordance with a particular code but with a variety of well-acknowledged engineering practices. For the final design it is compulsory to validate the design with respect to a typical engineering code in order to be compliant with the ITER quality and nuclear requirements and to get acceptance from the French regulator. This paper presents typical design validation of the closure plate, which is the vacuum and Tritium barrier and thus a safety relevant component of the upper port plug (UPP), performed with the ASME boiler and pressure vessel code. Rationales for choosing this code are given as well as a comparison between different design methods, like the “design by rule” and the “design by analysis” approach. Also the selections of proper load specifications and the identification of potential failure modes are covered. In addition to that stress categorizations, analyses

Design validation of the ITER EC upper launcher according to codes and standards

International Nuclear Information System (INIS)

Spaeh, Peter; Aiello, Gaetano; Gagliardi, Mario; Grossetti, Giovanni; Meier, Andreas; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro; Weinhorst, Bastian

2015-01-01

Highlights: • A set of applicable codes and standards has been chosen for the ITER EC upper launcher. • For a particular component load combinations, failure modes and stress categorizations have been determined. • The design validation was performed in accordance with the “design by analysis”-approach of the ASME boiler and pressure vessel code section III. - Abstract: The ITER electron cyclotron (EC) upper launcher has passed the CDR (conceptual design review) in 2005 and the PDR (preliminary design review) in 2009 and is in its final design phase now. The final design will be elaborated by the European consortium ECHUL-CA with contributions from several research institutes in Germany, Italy, the Netherlands and Switzerland. Within this consortium KIT is responsible for the design of the structural components (the upper port plug, UPP) and also the design integration of the launcher. As the selection of applicable codes and standards was under discussion for the past decade, the conceptual and the preliminary design of the launcher structure were not elaborated in straight accordance with a particular code but with a variety of well-acknowledged engineering practices. For the final design it is compulsory to validate the design with respect to a typical engineering code in order to be compliant with the ITER quality and nuclear requirements and to get acceptance from the French regulator. This paper presents typical design validation of the closure plate, which is the vacuum and Tritium barrier and thus a safety relevant component of the upper port plug (UPP), performed with the ASME boiler and pressure vessel code. Rationales for choosing this code are given as well as a comparison between different design methods, like the “design by rule” and the “design by analysis” approach. Also the selections of proper load specifications and the identification of potential failure modes are covered. In addition to that stress categorizations, analyses

The Mistra experiment for field containment code validation first results

International Nuclear Information System (INIS)

Caron-Charles, M.; Blumenfeld, L.

2001-01-01

The MISTRA facility is a large scale experiment, designed for the purpose of thermal-hydraulics multi-D codes validation. A short description of the facility, the set up of the instrumentation and the test program are presented. Then, the first experimental results, studying helium injection in the containment and their calculations are detailed. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

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

International Nuclear Information System (INIS)

Panayotov, Dobromir

2008-01-01

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

Validation Study of CODES Dragonfly Network Model with Theta Cray XC System

Energy Technology Data Exchange (ETDEWEB)

Mubarak, Misbah [Argonne National Lab. (ANL), Argonne, IL (United States); Ross, Robert B. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-05-31

This technical report describes the experiments performed to validate the MPI performance measurements reported by the CODES dragonfly network simulation with the Theta Cray XC system at the Argonne Leadership Computing Facility (ALCF).

Uncertainty propagation applied to multi-scale thermal-hydraulics coupled codes. A step towards validation

Energy Technology Data Exchange (ETDEWEB)

Geffray, Clotaire Clement

2017-03-20

The work presented here constitutes an important step towards the validation of the use of coupled system thermal-hydraulics and computational fluid dynamics codes for the simulation of complex flows in liquid metal cooled pool-type facilities. First, a set of methods suited for uncertainty and sensitivity analysis and validation activities with regards to the specific constraints of the work with coupled and expensive-to-run codes is proposed. Then, these methods are applied to the ATHLET - ANSYS CFX model of the TALL-3D facility. Several transients performed at this latter facility are investigated. The results are presented, discussed and compared to the experimental data. Finally, assessments of the validity of the selected methods and of the quality of the model are offered.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Initial verification and validation of RAZORBACK - A research reactor transient analysis code

Energy Technology Data Exchange (ETDEWEB)

Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2015-09-01

This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

Research on verification and validation strategy of detonation fluid dynamics code of LAD2D

Science.gov (United States)

Wang, R. L.; Liang, X.; Liu, X. Z.

2017-07-01

The verification and validation (V&V) is an important approach in the software quality assurance of code in complex engineering application. Reasonable and efficient V&V strategy can achieve twice the result with half the effort. This article introduces the software-Lagrangian adaptive hydrodynamics code in 2D space (LAD2D), which is self-developed software in detonation CFD with plastic-elastic structure. The V&V strategy of this detonation CFD code is presented based on the foundation of V&V methodology for scientific software. The basic framework of the module verification and the function validation is proposed, composing the detonation fluid dynamics model V&V strategy of LAD2D.

Validity of the Italian Code of Ethics for everyday nursing practice.

Science.gov (United States)

Gobbi, Paola; Castoldi, Maria Grazia; Alagna, Rosa Anna; Brunoldi, Anna; Pari, Chiara; Gallo, Annamaria; Magri, Miriam; Marioni, Lorena; Muttillo, Giovanni; Passoni, Claudia; Torre, Anna La; Rosa, Debora; Carnevale, Franco A

2016-12-07

The research question for this study was as follows: Is the Code of Ethics for Nurses in Italy (Code) a valid or useful decision-making instrument for nurses faced with ethical problems in their daily clinical practice? Focus groups were conducted to analyze specific ethical problems through 11 case studies. The analysis was conducted using sections of the Code as well as other relevant documents. Each focus group had a specific theme and nurses participated freely in the discussions according to their respective clinical competencies. The executive administrative committee of the local nursing licensing council provided approval for conducting this project. Measures were taken to protect the confidentiality of consenting participants. The answer to the research question posed for this investigation was predominantly positive. Many sections of the Code were useful for discussion and identifying possible solutions for the ethical problems presented in the 11 cases. We concluded that the Code of Ethics for Nurses in Italy can be a valuable aid in daily practice in most clinical situations that can give rise to ethical problems. © The Author(s) 2016.

CSNI Integral Test Facility Matrices for Validation of Best-Estimate Thermal-Hydraulic Computer Codes

International Nuclear Information System (INIS)

Glaeser, H.

2008-01-01

Internationally agreed Integral Test Facility (ITF) matrices for validation of realistic thermal hydraulic system computer codes were established. ITF development is mainly for Pressurised Water Reactors (PWRs) and Boiling Water Reactors (BWRs). A separate activity was for Russian Pressurised Water-cooled and Water-moderated Energy Reactors (WWER). Firstly, the main physical phenomena that occur during considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a list of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. In this paper some specific examples from the ITF matrices will also be provided. The matrices will be a guide for code validation, will be a basis for comparisons of code predictions performed with different system codes, and will contribute to the quantification of the uncertainty range of code model predictions. In addition to this objective, the construction of such a matrix is an attempt to record information which has been generated around the world over the last years, so that it is more accessible to present and future workers in that field than would otherwise be the case.

Guide to Using the WIND Toolkit Validation Code

Energy Technology Data Exchange (ETDEWEB)

Lieberman-Cribbin, W.; Draxl, C.; Clifton, A.

2014-12-01

In response to the U.S. Department of Energy's goal of using 20% wind energy by 2030, the Wind Integration National Dataset (WIND) Toolkit was created to provide information on wind speed, wind direction, temperature, surface air pressure, and air density on more than 126,000 locations across the United States from 2007 to 2013. The numerical weather prediction model output, gridded at 2-km and at a 5-minute resolution, was further converted to detail the wind power production time series of existing and potential wind facility sites. For users of the dataset it is important that the information presented in the WIND Toolkit is accurate and that errors are known, as then corrective steps can be taken. Therefore, we provide validation code written in R that will be made public to provide users with tools to validate data of their own locations. Validation is based on statistical analyses of wind speed, using error metrics such as bias, root-mean-square error, centered root-mean-square error, mean absolute error, and percent error. Plots of diurnal cycles, annual cycles, wind roses, histograms of wind speed, and quantile-quantile plots are created to visualize how well observational data compares to model data. Ideally, validation will confirm beneficial locations to utilize wind energy and encourage regional wind integration studies using the WIND Toolkit.

Nuclear Energy Knowledge and Validation Center (NEKVaC) Needs Workshop Summary Report

Energy Technology Data Exchange (ETDEWEB)

Gougar, Hans [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-02-01

The Department of Energy (DOE) has made significant progress developing simulation tools to predict the behavior of nuclear systems with greater accuracy and of increasing our capability to predict the behavior of these systems outside of the standard range of applications. These analytical tools require a more complex array of validation tests to accurately simulate the physics and multiple length and time scales. Results from modern simulations will allow experiment designers to narrow the range of conditions needed to bound system behavior and to optimize the deployment of instrumentation to limit the breadth and cost of the campaign. Modern validation, verification and uncertainty quantification (VVUQ) techniques enable analysts to extract information from experiments in a systematic manner and provide the users with a quantified uncertainty estimate. Unfortunately, the capability to perform experiments that would enable taking full advantage of the formalisms of these modern codes has progressed relatively little (with some notable exceptions in fuels and thermal-hydraulics); the majority of the experimental data available today is the "historic" data accumulated over the last decades of nuclear systems R&D. A validated code-model is a tool for users. An unvalidated code-model is useful for code developers to gain understanding, publish research results, attract funding, etc. As nuclear analysis codes have become more sophisticated, so have the measurement and validation methods and the challenges that confront them. A successful yet cost-effective validation effort requires expertise possessed only by a few, resources possessed only by the well-capitalized (or a willing collective), and a clear, well-defined objective (validating a code that is developed to satisfy the need(s) of an actual user). To that end, the Idaho National Laboratory established the Nuclear Energy Knowledge and Validation Center to address the challenges of modern code validation and to

Independent validation testing of the FLAME computer code, Version 1.0

International Nuclear Information System (INIS)

Martian, P.; Chung, J.N.

1992-07-01

Independent testing of the FLAME computer code, Version 1.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Validation tests, (i.e., tests which compare field data to the computer generated solutions) were used to determine the operational status of the FLAME computer code and were done on a qualitative basis through graphical comparisons of the experimental and numerical data. These tests were specifically designed to check: (1) correctness of the FORTRAN coding, (2) computational accuracy, and (3) suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: (1) independent applications, and (2) graduated difficulty of test cases. Three tests ranging in complexity from simple one-dimensional steady-state flow field problems under near-saturated conditions to two-dimensional transient flow problems with very dry initial conditions

System code improvements for modelling passive safety systems and their validation

Energy Technology Data Exchange (ETDEWEB)

Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-11-15

GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

Development and validation of the Pediatric Anesthesia Behavior score--an objective measure of behavior during induction of anesthesia.

Science.gov (United States)

Beringer, Richard M; Greenwood, Rosemary; Kilpatrick, Nicky

2014-02-01

Measuring perioperative behavior changes requires validated objective rating scales. We developed a simple score for children's behavior during induction of anesthesia (Pediatric Anesthesia Behavior score) and assessed its reliability, concurrent validity, and predictive validity. Data were collected as part of a wider observational study of perioperative behavior changes in children undergoing general anesthesia for elective dental extractions. One-hundred and two healthy children aged 2-12 were recruited. Previously validated behavioral scales were used as follows: the modified Yale Preoperative Anxiety Scale (m-YPAS); the induction compliance checklist (ICC); the Pediatric Anesthesia Emergence Delirium scale (PAED); and the Post-Hospitalization Behavior Questionnaire (PHBQ). Pediatric Anesthesia Behavior (PAB) score was independently measured by two investigators, to allow assessment of interobserver reliability. Concurrent validity was assessed by examining the correlation between the PAB score, the m-YPAS, and the ICC. Predictive validity was assessed by examining the association between the PAB score, the PAED scale, and the PHBQ. The PAB score correlated strongly with both the m-YPAS (P risk of developing postoperative behavioral disturbance. This study provides evidence for its reliability and validity. © 2013 John Wiley & Sons Ltd.

Development, validation and application of NAFA 2D-CFD code

International Nuclear Information System (INIS)

Vaidya, A.M.; Maheshwari, N.K.; Vijayan, P.K.; Saha, D.

2010-01-01

A 2D axi-symmetric code named NAFA (Version 1.0) is developed for studying the pipe flow under various conditions. It can handle laminar/ turbulent flows, with or without heat transfer, under sub-critical/super-critical conditions. The code solves for momentum, energy equations with standard k-ε turbulence model (with standard wall functions). It solves pipe flow subjected to 'velocity inlet', 'wall', 'axis' and 'pressure outlet' boundary conditions. It is validated for several cases by comparing its results with experimental data/analytical solutions/correlations. The code has excellent convergence characteristics as verified from fall of equation residual in each case. It has proven capability of generating mesh independent results for laminar as well as turbulent flows. The code is applied to supercritical flows. For supercritical flows, the effect of mesh size on prediction of heat transfer coefficient is studied. With grid refinement, the Y + reduces and reaches the limiting value of 11.63. Hence the accuracy is found to increase with grid refinement. NAFA is able to qualitatively predict the effect of heat flux and operating pressure on heat transfer coefficient. The heat transfer coefficient matches well with experimental values under various conditions. (author)

From Symbolic to Substantive Documents: When Business Codes of Ethics Impact Unethical Behavior in the Workplace

OpenAIRE

Kaptein, S.P.

2009-01-01

textabstractA business code of ethics is widely regarded as an important instrument to curb unethical behavior in the workplace. However, little is empirically known about the factors that determine the impact of a code on unethical behavior. Besides the existence of a code, this study proposes five determining factors: the content of the code, the frequency of communication activities surrounding the code, the quality of the communication activities, and the embedment of the code in the orga...

Modification and validation of the natural heat convection and subcooled void formation models in the code PARET

International Nuclear Information System (INIS)

Hainoun, A.; Alhabit, F.; Ghazi, N.

2008-01-01

Two new modifications have been included in the current PARET code that is widely applied in the dynamic and safety analysis of research reactors. A new model was implemented for the simulation of void formation in the subcooled boiling regime, the other modification dealt with the implementation of a new approach to improve the prediction of heat transfer coefficient under natural circulation condition. The modified code was successfully validated using adequate single effect tests covering the physical phenomena of interest for both natural circulation and subcooled void formation at low pressure and low heat flux. The validation results indicate significant improvement of the code compared to the default version. Additionally, to simplify the code application an interactive user interface was developed enabling pre and post-processing of the code predictions. (author)

Development and validation of the ENIGMA code for MOX fuel performance modelling

International Nuclear Information System (INIS)

Palmer, I.; Rossiter, G.; White, R.J.

2000-01-01

The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO 2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO 2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)

Test and validation of the iterative code for the neutrons spectrometry and dosimetry: NSDUAZ

International Nuclear Information System (INIS)

Reyes H, A.; Ortiz R, J. M.; Reyes A, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.

2014-08-01

In this work was realized the test and validation of an iterative code for neutronic spectrometry known as Neutron Spectrometry and Dosimetry of the Universidad Autonoma de Zacatecas (NSDUAZ). This code was designed in a user graph interface, friendly and intuitive in the environment programming of LabVIEW using the iterative algorithm known as SPUNIT. The main characteristics of the program are: the automatic selection of the initial spectrum starting from the neutrons spectra catalog compiled by the International Atomic Energy Agency, the possibility to generate a report in HTML format that shows in graph and numeric way the neutrons flowing and calculates the ambient dose equivalent with base to this. To prove the designed code, the count rates of a spectrometer system of Bonner spheres were used with a detector of 6 LiI(Eu) with 7 polyethylene spheres with diameter of 0, 2, 3, 5, 8, 10 and 12. The count rates measured with two neutron sources: 252 Cf and 239 PuBe were used to validate the code, the obtained results were compared against those obtained using the BUNKIUT code. We find that the reconstructed spectra present an error that is inside the limit reported in the literature that oscillates around 15%. Therefore, it was concluded that the designed code presents similar results to those techniques used at the present time. (Author)

Validation and verification of the ORNL Monte Carlo codes for nuclear safety analysis

International Nuclear Information System (INIS)

Emmett, M.B.

1993-01-01

The process of ensuring the quality of computer codes can be very time consuming and expensive. The Oak Ridge National Laboratory (ORNL) Monte Carlo codes all predate the existence of quality assurance (QA) standards and configuration control. The number of person-years and the amount of money spent on code development make it impossible to adhere strictly to all the current requirements. At ORNL, the Nuclear Engineering Applications Section of the Computing Applications Division is responsible for the development, maintenance, and application of the Monte Carlo codes MORSE and KENO. The KENO code is used for doing criticality analyses; the MORSE code, which has two official versions, CGA and SGC, is used for radiation transport analyses. Because KENO and MORSE were very thoroughly checked out over the many years of extensive use both in the United States and in the international community, the existing codes were open-quotes baselined.close quotes This means that the versions existing at the time the original configuration plan is written are considered to be validated and verified code systems based on the established experience with them

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

International Nuclear Information System (INIS)

Cevolani, S.

1995-01-01

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

Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

International Nuclear Information System (INIS)

Scheglov, A.; Proselkov, V.

2001-01-01

This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

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

International Nuclear Information System (INIS)

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

2001-08-01

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

Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Chakir, E.; El Bakkari, B.; El Younoussi, C.

2015-01-01

Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

Validation of the containment code Sirius: interpretation of an explosion experiment on a scale model

International Nuclear Information System (INIS)

Blanchet, Y.; Obry, P.; Louvet, J.; Deshayes, M.; Phalip, C.

1979-01-01

The explicit 2-D axisymmetric Langrangian code SIRIUS, developed at the CEA/DRNR, Cadarache, deals with transient compressive flows in deformable primary tanks with more or less complex internal component geometries. This code has been subjected to a two-year intensive validation program on scale model experiments and a number of improvements have been incorporated. This paper presents a recent calculation of one of these experiments using the SIRIUS code, and the comparison with experimental results shows the encouraging possibilities of this Lagrangian code

From Symbolic to Substantive Documents: When Business Codes of Ethics Impact Unethical Behavior in the Workplace

NARCIS (Netherlands)

S.P. Kaptein (Muel)

2009-01-01

textabstractA business code of ethics is widely regarded as an important instrument to curb unethical behavior in the workplace. However, little is empirically known about the factors that determine the impact of a code on unethical behavior. Besides the existence of a code, this study proposes five

Validation and testing of the VAM2D computer code

International Nuclear Information System (INIS)

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

1991-10-01

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

Computer code ENDSAM for random sampling and validation of the resonance parameters covariance matrices of some major nuclear data libraries

International Nuclear Information System (INIS)

Plevnik, Lucijan; Žerovnik, Gašper

2016-01-01

Highlights: • Methods for random sampling of correlated parameters. • Link to open-source code for sampling of resonance parameters in ENDF-6 format. • Validation of the code on realistic and artificial data. • Validation of covariances in three major contemporary nuclear data libraries. - Abstract: Methods for random sampling of correlated parameters are presented. The methods are implemented for sampling of resonance parameters in ENDF-6 format and a link to the open-source code ENDSAM is given. The code has been validated on realistic data. Additionally, consistency of covariances of resonance parameters of three major contemporary nuclear data libraries (JEFF-3.2, ENDF/B-VII.1 and JENDL-4.0u2) has been checked.

Validation of two-phase flow code THYC on VATICAN experiment

International Nuclear Information System (INIS)

Maurel, F.; Portesse, A.; Rimbert, P.; Thomas, B.

1997-01-01

As part of a comprehensive program for THYC validation (THYC is a 3-dimensional two-phase flow computer code for PWR core configuration), an experimental project > has been initiated by the Direction des Etudes et Recherches of Electricite de France. Two mock-ups tested in Refrigerant-114, VATICAN-1 (with simple space grids) and VATICAN-2 (with mixing grids) were set up to investigate void fraction distributions using a single beam gamma densitometer. First, experiments were conducted with the VATICAN-1 mock-up. A set of constitutive laws to be used in rod bundles was determined but some doubts still remain for friction losses closure laws for oblique flow over tubes. From VATICAN-2 tests, calculations were performed using the standard set of correlations. Comparison with the experimental data shows an underprediction of void fraction by THYC in disturbed regions. Analyses highlight the poor treatment of axial relative velocity in these regions. A fitting of the radial and axial relative velocity values in the disturbed region improves the prediction of void fraction by the code but without any physical explanation. More analytical experiments should be carried out to validate friction losses closure laws for oblique flows and relative velocity downstream of a mixing grid. (author)

Validation of two-phase flow code THYC on VATICAN experiment

Energy Technology Data Exchange (ETDEWEB)

Maurel, F.; Portesse, A.; Rimbert, P.; Thomas, B. [EDF/DER, Dept. TTA, 78 - Chatou (France)

1997-12-31

As part of a comprehensive program for THYC validation (THYC is a 3-dimensional two-phase flow computer code for PWR core configuration), an experimental project <> has been initiated by the Direction des Etudes et Recherches of Electricite de France. Two mock-ups tested in Refrigerant-114, VATICAN-1 (with simple space grids) and VATICAN-2 (with mixing grids) were set up to investigate void fraction distributions using a single beam gamma densitometer. First, experiments were conducted with the VATICAN-1 mock-up. A set of constitutive laws to be used in rod bundles was determined but some doubts still remain for friction losses closure laws for oblique flow over tubes. From VATICAN-2 tests, calculations were performed using the standard set of correlations. Comparison with the experimental data shows an underprediction of void fraction by THYC in disturbed regions. Analyses highlight the poor treatment of axial relative velocity in these regions. A fitting of the radial and axial relative velocity values in the disturbed region improves the prediction of void fraction by the code but without any physical explanation. More analytical experiments should be carried out to validate friction losses closure laws for oblique flows and relative velocity downstream of a mixing grid. (author)

A validation study of the BURNUP and associated options of the MONTE CARLO neutronics code MONK5W

International Nuclear Information System (INIS)

Howard, E.A.

1985-11-01

This is a report on the validation of the burnup option of the Monte Carlo Neutronics Code MONK5W, together with the associated facilities which allow for control rod movements and power changes. The validation uses reference solutions produced by the Deterministic Neutronics Code LWR-WIMS for a 2D model which represents a whole reactor calculation with control rod movements. (author)

CACTI: free, open-source software for the sequential coding of behavioral interactions.

Science.gov (United States)

Glynn, Lisa H; Hallgren, Kevin A; Houck, Jon M; Moyers, Theresa B

2012-01-01

The sequential analysis of client and clinician speech in psychotherapy sessions can help to identify and characterize potential mechanisms of treatment and behavior change. Previous studies required coding systems that were time-consuming, expensive, and error-prone. Existing software can be expensive and inflexible, and furthermore, no single package allows for pre-parsing, sequential coding, and assignment of global ratings. We developed a free, open-source, and adaptable program to meet these needs: The CASAA Application for Coding Treatment Interactions (CACTI). Without transcripts, CACTI facilitates the real-time sequential coding of behavioral interactions using WAV-format audio files. Most elements of the interface are user-modifiable through a simple XML file, and can be further adapted using Java through the terms of the GNU Public License. Coding with this software yields interrater reliabilities comparable to previous methods, but at greatly reduced time and expense. CACTI is a flexible research tool that can simplify psychotherapy process research, and has the potential to contribute to the improvement of treatment content and delivery.

Development and Validation of A Nuclear Fuel Cycle Analysis Tool: A FUTURE Code

Energy Technology Data Exchange (ETDEWEB)

Kim, S. K.; Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yoon Hee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2013-10-15

This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy) code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

First steps towards a validation of the new burnup and depletion code TNT

Energy Technology Data Exchange (ETDEWEB)

Herber, S.C.; Allelein, H.J. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6); Friege, N. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Kasselmann, S. [Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6)

2012-11-01

In the frame of the fusion of the core design calculation capabilities, represented by V.S.O.P., and the accident calculation capabilities, represented by MGT(-3D), the successor of the TINTE code, difficulties were observed in defining an interface between a program backbone and the ORIGEN code respectively the ORIGENJUEL code. The estimation of the effort of refactoring the ORIGEN code or to write a new burnup code from scratch, led to the decision that it would be more efficient writing a new code, which could benefit from existing programming and software engineering tools from the computer code side and which can use the latest knowledge of nuclear reactions, e.g. consider all documented reaction channels. Therefore a new code with an object-oriented approach was developed at IEK-6. Object-oriented programming is currently state of the art and provides mostly an improved extensibility and maintainability. The new code was named TNT which stands for Topological Nuclide Transformation, since the code makes use of the real topology of the nuclear reactions. Here we want to present some first validation results from code to code benchmarks with the codes ORIGEN V2.2 and FISPACT2005 and whenever possible analytical results also used for the comparison. The 2 reference codes were chosen due to their high reputation in the field of fission reactor analysis (ORIGEN) and fusion facilities (FISPACT). (orig.)

Aztheca Code; Codigo Aztheca

Energy Technology Data Exchange (ETDEWEB)

Quezada G, S.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Centeno P, J.; Sanchez M, H., E-mail: sequga@gmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, Circuito Exterior s/n, 04510 Ciudad de Mexico (Mexico)

2017-09-15

This paper presents the Aztheca code, which is formed by the mathematical models of neutron kinetics, power generation, heat transfer, core thermo-hydraulics, recirculation systems, dynamic pressure and level models and control system. The Aztheca code is validated with plant data, as well as with predictions from the manufacturer when the reactor operates in a stationary state. On the other hand, to demonstrate that the model is applicable during a transient, an event occurred in a nuclear power plant with a BWR reactor is selected. The plant data are compared with the results obtained with RELAP-5 and the Aztheca model. The results show that both RELAP-5 and the Aztheca code have the ability to adequately predict the behavior of the reactor. (Author)

Implementation of QR Code and Digital Signature to Determine the Validity of KRS and KHS Documents

Directory of Open Access Journals (Sweden)

Fatich Fazlur Rochman

2017-05-01

Full Text Available Universitas Airlangga students often find it difficult to verify the mark that came out in the Kartu Hasil Studi (KHS is called Study Result Card or courses taken in the Kartu Rencana Studi (KRS is called Study Plan Card, if there are changes to the data on the system used Universitas Airlangga. This complicated KRS and KHS verification process happened because the KRS and KHS documents that owned by student is easier to counterfeit than the data in the system. Implementation digital signature and QR Code technology as a solution that can prove the validity of KRS or KHS. The KRS and KHS validation system developed by Digital Signature and QR Code. QR Code is a type of matrix code that was developed as a code that allows its contents to be decoded at high speed while the Digital Signature has a function as a marker on the data to ensure that the data is the original data. The verification process was divided into two types are reading the Digital Signature and printing document that works by scanning the data from QR Code. The application of the system is carried out were the addition of the QR Code on KRS and KHS, required a readiness of human resources. 

TASS code topical report. V.2 TASS code validation report for the non-LOCA transient analysis of the CE and Westinghouse type plants

International Nuclear Information System (INIS)

Sim, Suk K.; Chang, W. P.; Kim, K. D.; Lee, S. J.; Kim, H. C.; Yoon, H. Y.

1997-02-01

The development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the CE and Westinghouse type operating reactors as well as the PWR plants under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the FSAR transients, loss of AC power transient plant data, load rejection and startup test data for the reference plants as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best FSAR transient and shows its capability in simulating plant transient analyses. (author). 12 refs., 32 tabs., 132 figs

Preliminary validation of the MATRA-LMR-FB code for the flow blockage in a subassembly

International Nuclear Information System (INIS)

Jeong, H. Y.; Ha, K. S.; Kwon, Y. M.; Chang, W. P.; Lee, Y. B.; Heo, S.

2005-01-01

To analyze the flow blockage in a subassembly of a Liquid Metal-cooled Reactor (LMR), the MATRA-LMR-FB code has been developed and validated for the existing experimental data. Compared to the MATRA-LMR code, which had been successfully applied for the core thermal-hydraulic design of KALIMER, the MATRA-LMR-FB code includes some advanced modeling features. Firstly, the Distributed Resistance Model (DRM), which enables a very accurate description of the effects of wire-wrap and blockage in a flow path, is developed for the MATRA-LMR-FB code. Secondly, the hybrid difference method is used to minimize the numerical diffusion especially at the low flow region such as recirculating wakes after blockage. In addition, the code is equipped with various turbulent mixing models to describe the active mixing due to the turbulent motions as accurate as possible. For the validation of the MATRA-LMR-FB code the ORNL THORS test and KOS 169-pin test are analyzed. Based on the analysis results for the temperature data, the accuracy of the code is evaluated quantitatively. The MATRA-LMR-FB code predicts very accurately the exit temperatures measured in the subassembly with wire-wrap. However, the predicted temperatures for the experiment with spacer grid show some deviations from the measured. To enhance the accuracy of the MATRA-LMR-FB for the flow path with grid spacers, it is suggested to improve the models for pressure loss due to spacer grid and the modeling method for blockage itself. The developed MATRA-LMR-FB code is evaluated to be applied to the flow blockage analysis of KALIMER-600 which adopts the wire-wrapped subassemblies

Validation of a CFD code for Unsteady Flows with cyclic boundary Conditions

International Nuclear Information System (INIS)

Kim, Jong-Tae; Kim, Sang-Baik; Lee, Won-Jae

2006-01-01

Currently Lilac code is under development to analyze thermo-hydraulics of a high-temperature gas-cooled reactor (GCR). Interesting thermo-hydraulic phenomena in a nuclear reactor are usually unsteady and turbulent. The analysis of the unsteady flows by using a three dimension CFD code is time-consuming if the flow domain is very large. Hopefully, flow domains commonly encountered in the nuclear thermo-hydraulics is periodic. So it is better to use the geometrical characteristics in order to reduce the computational resources. To get the benefits from reducing the computation domains especially for the calculations of unsteady flows, the cyclic boundary conditions are implemented in the parallelized CFD code LILAC. In this study, the parallelized cyclic boundary conditions are validated by solving unsteady laminar and turbulent flows past a circular cylinder

DEVELOPMENT AND VALIDATION OF A NUCLEAR FUEL CYCLE ANALYSIS TOOL: A FUTURE CODE

Directory of Open Access Journals (Sweden)

S.K. KIM

2013-10-01

Full Text Available This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C# and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.

Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

Science.gov (United States)

Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

International Nuclear Information System (INIS)

Suyama, K.; Mochizuki, H.; Okuno, H.; Miyoshi, Y.

2004-01-01

This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models. (authors)

Validity of Principal Diagnoses in Discharge Summaries and ICD-10 Coding Assessments Based on National Health Data of Thailand.

Science.gov (United States)

Sukanya, Chongthawonsatid

2017-10-01

This study examined the validity of the principal diagnoses on discharge summaries and coding assessments. Data were collected from the National Health Security Office (NHSO) of Thailand in 2015. In total, 118,971 medical records were audited. The sample was drawn from government hospitals and private hospitals covered by the Universal Coverage Scheme in Thailand. Hospitals and cases were selected using NHSO criteria. The validity of the principal diagnoses listed in the "Summary and Coding Assessment" forms was established by comparing data from the discharge summaries with data obtained from medical record reviews, and additionally, by comparing data from the coding assessments with data in the computerized ICD (the data base used for reimbursement-purposes). The summary assessments had low sensitivities (7.3%-37.9%), high specificities (97.2%-99.8%), low positive predictive values (9.2%-60.7%), and high negative predictive values (95.9%-99.3%). The coding assessments had low sensitivities (31.1%-69.4%), high specificities (99.0%-99.9%), moderate positive predictive values (43.8%-89.0%), and high negative predictive values (97.3%-99.5%). The discharge summaries and codings often contained mistakes, particularly the categories "Endocrine, nutritional, and metabolic diseases", "Symptoms, signs, and abnormal clinical and laboratory findings not elsewhere classified", "Factors influencing health status and contact with health services", and "Injury, poisoning, and certain other consequences of external causes". The validity of the principal diagnoses on the summary and coding assessment forms was found to be low. The training of physicians and coders must be strengthened to improve the validity of discharge summaries and codings.

Validation of the DRAGON/DONJON code package for MNR using the IAEA 10 MW benchmark problem

International Nuclear Information System (INIS)

Day, S.E.; Garland, W.J.

2000-01-01

The first step in developing a framework for reactor physics analysis is to establish the appropriate and proven reactor physics codes. The chosen code package is tested, by executing a benchmark problem and comparing the results to the accepted standards. The IAEA 10 MW Benchmark problem is suitable for static reactor physics calculations on plate-fueled research reactor systems and has been used previously to validate codes for the McMaster Nuclear (MNR). The flexible and advanced geometry capabilities of the DRAGON transport theory code make it a desirable tool, and the accompanying DONJON diffusion theory code also has useful features applicable to safety analysis work at MNR. This paper describes the methodology used to benchmark the DRAGON/DONJON code package against this problem and the results herein extend the domain of validation of this code package. The results are directly applicable to MNR and are relevant to a reduced-enrichment fuel program. The DRAGON transport code models, used in this study, are based on the 1-D infinite slab approximation whereas the DONJON diffusion code models are defined in 3-D Cartesian geometry. The cores under consideration are composed of HEU (93% enrichment), MEU (45% enrichment) and LEU (20% enrichment) fuel and are examined in a fresh state, as well as at beginning-of-life (BOL) and end-of-life (EOL) exposures. The required flux plots and flux-ratio plots are included, as are transport theory code k∞and diffusion theory code k eff results. In addition to this, selected isotope atom densities are charted as a function of fuel burnup. Results from this analysis are compared to and are in good agreement with previously published results. (author)

Validation of a pre-coded food record for infants and young children

DEFF Research Database (Denmark)

Gondolf, Ulla Holmboe; Tetens, Inge; Hills, A. P.

2012-01-01

Background/Objectives:To assess the validity of a 7-day pre-coded food record (PFR) method in 9-month-old infants against metabolizable energy intake (ME(DLW)) measured by doubly labeled water (DLW); additionally to compare PFR with a 7-day weighed food record (WFR) in 9-month-old infants and 36...

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

International Nuclear Information System (INIS)

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

2003-02-01

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

Development and validation of corium oxidation model for the VAPEX code

International Nuclear Information System (INIS)

Blinkov, V.N.; Melikhov, V.I.; Davydov, M.V.; Melikhov, O.I.; Borovkova, E.M.

2011-01-01

In light water reactor core melt accidents, the molten fuel (corium) can be brought into contact with coolant water in the course of the melt relocation in-vessel and ex-vessel as well as in an accident mitigation action of water addition. Mechanical energy release from such an interaction is of interest in evaluating the structural integrity of the reactor vessel as well as of the containment. Usually, the source for the energy release is considered to be the rapid transfer of heat from the molten fuel to the water ('vapor explosion'). When the fuel contains a chemically reactive metal component, there could be an additional source for the energy release, which is the heat release and hydrogen production due to the metal-water chemical reaction. In Electrogorsk Research and Engineering Center the computer code VAPEX (VAPor EXplosion) has been developed for analysis of the molten fuel coolant interaction. Multifield approach is used for modeling of dynamics of following phases: water, steam, melt jet, melt droplets, debris. The VAPEX code was successfully validated on FARO experimental data. Hydrogen generation was observed in FARO tests even though corium didn't contain metal component. The reason for hydrogen generation was not clear, so, simplified empirical model of hydrogen generation was implemented in the VAPEX code to take into account input of hydrogen into pressure increase. This paper describes new more detailed model of hydrogen generation due to the metal-water chemical reaction and results of its validation on ZREX experiments. (orig.)

Caregiver person-centeredness and behavioral symptoms during mealtime interactions: development and feasibility of a coding scheme.

Science.gov (United States)

Gilmore-Bykovskyi, Andrea L

2015-01-01

Mealtime behavioral symptoms are distressing and frequently interrupt eating for the individual experiencing them and others in the environment. A computer-assisted coding scheme was developed to measure caregiver person-centeredness and behavioral symptoms for nursing home residents with dementia during mealtime interactions. The purpose of this pilot study was to determine the feasibility, ease of use, and inter-observer reliability of the coding scheme, and to explore the clinical utility of the coding scheme. Trained observers coded 22 observations. Data collection procedures were acceptable to participants. Overall, the coding scheme proved to be feasible, easy to execute and yielded good to very good inter-observer agreement following observer re-training. The coding scheme captured clinically relevant, modifiable antecedents to mealtime behavioral symptoms, but would be enhanced by the inclusion of measures for resident engagement and consolidation of items for measuring caregiver person-centeredness that co-occurred and were difficult for observers to distinguish. Published by Elsevier Inc.

Development and Validation of the Behavioral Tendencies Questionnaire.

Directory of Open Access Journals (Sweden)

Nicholas T Van Dam

Full Text Available At a fundamental level, taxonomy of behavior and behavioral tendencies can be described in terms of approach, avoid, or equivocate (i.e., neither approach nor avoid. While there are numerous theories of personality, temperament, and character, few seem to take advantage of parsimonious taxonomy. The present study sought to implement this taxonomy by creating a questionnaire based on a categorization of behavioral temperaments/tendencies first identified in Buddhist accounts over fifteen hundred years ago. Items were developed using historical and contemporary texts of the behavioral temperaments, described as "Greedy/Faithful", "Aversive/Discerning", and "Deluded/Speculative". To both maintain this categorical typology and benefit from the advantageous properties of forced-choice response format (e.g., reduction of response biases, binary pairwise preferences for items were modeled using Latent Class Analysis (LCA. One sample (n1 = 394 was used to estimate the item parameters, and the second sample (n2 = 504 was used to classify the participants using the established parameters and cross-validate the classification against multiple other measures. The cross-validated measure exhibited good nomothetic span (construct-consistent relationships with related measures that seemed to corroborate the ideas present in the original Buddhist source documents. The final 13-block questionnaire created from the best performing items (the Behavioral Tendencies Questionnaire or BTQ is a psychometrically valid questionnaire that is historically consistent, based in behavioral tendencies, and promises practical and clinical utility particularly in settings that teach and study meditation practices such as Mindfulness Based Stress Reduction (MBSR.

Typology of Couples Entering Alcohol Behavioral Couple Therapy: An Empirical Approach and Test of Predictive Validity on Treatment Response.

Science.gov (United States)

Ladd, Benjamin O; McCrady, Barbara S

2016-01-01

This study aimed to examine whether classification of couples in which one partner has an alcohol problem is similar to that reported in the general couples literature. Typologies of couples seeking alcohol behavioral couple therapy (ABCT) were developed via hierarchical cluster analysis using behavioral codes of couple interactions during their first ABCT session. Four couples types based on in-session behavior were established reliably, labeled avoider, validator, hostile, and ambivalent-detached. These couple types resembled couples types found in previous research. Couple type was associated with baseline relationship satisfaction, but not alcohol use. Results suggest heterogeneity in couples with alcohol problems presenting to treatment; further study is needed to investigate the function of alcohol within these different types. © 2015 American Association for Marriage and Family Therapy.

Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Escalante, Javier Jimenez; Espinoza, Victor Sanchez

2015-07-15

Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.

Cracking the Neural Code for Sensory Perception by Combining Statistics, Intervention, and Behavior.

Science.gov (United States)

Panzeri, Stefano; Harvey, Christopher D; Piasini, Eugenio; Latham, Peter E; Fellin, Tommaso

2017-02-08

The two basic processes underlying perceptual decisions-how neural responses encode stimuli, and how they inform behavioral choices-have mainly been studied separately. Thus, although many spatiotemporal features of neural population activity, or "neural codes," have been shown to carry sensory information, it is often unknown whether the brain uses these features for perception. To address this issue, we propose a new framework centered on redefining the neural code as the neural features that carry sensory information used by the animal to drive appropriate behavior; that is, the features that have an intersection between sensory and choice information. We show how this framework leads to a new statistical analysis of neural activity recorded during behavior that can identify such neural codes, and we discuss how to combine intersection-based analysis of neural recordings with intervention on neural activity to determine definitively whether specific neural activity features are involved in a task. Copyright © 2017 Elsevier Inc. All rights reserved.

Conception and validation of the Behavioral Intentions Scale of Organizational Citizenship (BISOC

Directory of Open Access Journals (Sweden)

Ana Cristina Passos Gomes Menezes

2016-01-01

Full Text Available Abstract This study aimed to construct and validate the Behavioral Intentions of Organizational Citizenship Scale (BISOC. Organizational citizenship consists of measures of voluntary behaviors, which are beneficial to organizations and are not explicit in employment contracts. To investigate the psychometric properties of BISOC, we selected 767 employees in different cities from the states of Bahia and Pernambuco (Brazil. The validation procedures adopted, which used techniques from both Classical Test Theory and Item Response Theory, showed that the BISOC has a unidimensional structure. From the initial set of 42 items, 35 items met the validation criteria. By presenting suitable psychometric parameters, BISOC is the first measure of organizational citizenship behaviors developed and validated to assess behavioral intentions.

Off-take Model of the SPACE Code and Its Validation

International Nuclear Information System (INIS)

Oh, Myung Taek; Park, Chan Eok; Sohn, Jong Joo

2011-01-01

Liquid entrainment and vapor pull-through models of horizontal pipe have been implemented in the SPACE code. The model of SPACE accounts for the phase separation phenomena and computes the flux of mass and energy through an off-take attached to a horizontal pipe when stratified conditions occur in the horizontal pipe. This model is referred to as the off-take model. The importance of predicting the fluid conditions through an off-take in a small-break LOCA has been well known. In this case, the occurrence of the stratification can affect the break node void fraction and thus the break flow discharged from the primary system. In order to validate the off-take model newly developed for the SPACE code, a simulation of the HDU experiments has been performed. The main feature of the off-take model and its application results will be presented in this paper

Calibrating and validating a FE model for long-term behavior of RC beams

Directory of Open Access Journals (Sweden)

Tošić Nikola D.

2014-01-01

Full Text Available This study presents the research carried out in finding an optimal finite element (FE model for calculating the long-term behavior of reinforced concrete (RC beams. A multi-purpose finite element software DIANA was used. A benchmark test in the form of a simply supported beam loaded in four point bending was selected for model calibration. The result was the choice of 3-node beam elements, a multi-directional fixed crack model with constant stress cut-off, nonlinear tension softening and constant shear retention and a creep and shrinkage model according to CEB-FIP Model Code 1990. The model was then validated on 14 simply supported beams and 6 continuous beams. Good agreement was found with experimental results (within ±15%.

Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

Energy Technology Data Exchange (ETDEWEB)

Chang, H. Oh, PhD; Cliff Davis; Richard Moore

2004-11-01

The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

Calculation of Sodium Fire Test-I (Run-E6) using sodium combustion analysis code ASSCOPS version 2.0

Energy Technology Data Exchange (ETDEWEB)

Nakagiri, Toshio; Ohno, Shuji; Miyake, Osamu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1997-11-01

The calculation of Sodium Fire Test-I (Run-E6) was performed using the ASSCOPS (Analysis of Simultaneous Sodium Combustions in Pool and Spray) code version 2.0 in order to determine the parameters used in the code for the calculations of sodium combustion behavior of small or medium scale sodium leak, and to validate the applicability of the code. The parameters used in the code were determined and the validation of the code was confirmed because calculated temperatures, calculated oxygen concentration and other calculated values almost agreed with the test results. (author)

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

International Nuclear Information System (INIS)

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

2016-09-01

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

Validation of ICD-9 Codes for Stable Miscarriage in the Emergency Department.

Science.gov (United States)

Quinley, Kelly E; Falck, Ailsa; Kallan, Michael J; Datner, Elizabeth M; Carr, Brendan G; Schreiber, Courtney A

2015-07-01

International Classification of Disease, Ninth Revision (ICD-9) diagnosis codes have not been validated for identifying cases of missed abortion where a pregnancy is no longer viable but the cervical os remains closed. Our goal was to assess whether ICD-9 code "632" for missed abortion has high sensitivity and positive predictive value (PPV) in identifying patients in the emergency department (ED) with cases of stable early pregnancy failure (EPF). We studied females ages 13-50 years presenting to the ED of an urban academic medical center. We approached our analysis from two perspectives, evaluating both the sensitivity and PPV of ICD-9 code "632" in identifying patients with stable EPF. All patients with chief complaints "pregnant and bleeding" or "pregnant and cramping" over a 12-month period were identified. We randomly reviewed two months of patient visits and calculated the sensitivity of ICD-9 code "632" for true cases of stable miscarriage. To establish the PPV of ICD-9 code "632" for capturing missed abortions, we identified patients whose visits from the same time period were assigned ICD-9 code "632," and identified those with actual cases of stable EPF. We reviewed 310 patient records (17.6% of 1,762 sampled). Thirteen of 31 patient records assigned ICD-9 code for missed abortion correctly identified cases of stable EPF (sensitivity=41.9%), and 140 of the 142 patients without EPF were not assigned the ICD-9 code "632"(specificity=98.6%). Of the 52 eligible patients identified by ICD-9 code "632," 39 cases met the criteria for stable EPF (PPV=75.0%). ICD-9 code "632" has low sensitivity for identifying stable EPF, but its high specificity and moderately high PPV are valuable for studying cases of stable EPF in epidemiologic studies using administrative data.

The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

Energy Technology Data Exchange (ETDEWEB)

Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-09-01

MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input that describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of

VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS

Directory of Open Access Journals (Sweden)

Tagor Malem Sembiring

2015-10-01

Full Text Available ABSTRACT VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS. The coupled neutronic and thermal-hydraulic (T/H code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR ejection at peripheral core using a full core geometry model, the C1 and C2 cases.  By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM and the improved quasistatic method (IQS. All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16% occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4% for C2 case.  All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. Keywords: nodal method, coupled neutronic and thermal-hydraulic code, PWR, transient case, control rod ejection.   ABSTRAK VALIDASI MODEL GEOMETRI TERAS PENUH PAKET PROGRAM NODAL3 DALAM PROBLEM BENCHMARK GAYUT WAKTU PWR. Paket program kopel neutronik dan termohidraulika (T/H, NODAL3, telah divalidasi dengan beberapa kasus benchmark statis PWR dan kasus benchmark gayut waktu PWR NEACRP.  Akan tetapi, paket program NODAL3 belum divalidasi dalam kasus benchmark gayut waktu akibat penarikan sebuah perangkat batang kendali (CR di tepi teras menggunakan model geometri teras penuh, yaitu kasus C1 dan C2. Dengan penelitian ini, akurasi paket program

Preliminary validation of RELAP5/Mod4.0 code for LBE cooled NACIE facility

Energy Technology Data Exchange (ETDEWEB)

Kumari, Indu; Khanna, Ashok, E-mail: akhanna@iitk.ac.in

2017-04-01

Highlights: • Detail discussion of thermo physical properties of Lead Bismuth Eutectic incorporated in the code RELAP5/Mod4.0 included. • Benchmarking of LBE properties in RELAP5/Mod4.0 against literature. • NACIE facility for three different power levels (10.8, 21.7 and 32.5 kW) under natural circulation considered for benchmarking. • Preliminary validation of the LBE properties against experimental data. • NACIE facility for power level 22.5 kW considered for validation. - Abstract: The one-dimensional thermal hydraulic computer code RELAP5 was developed for thermal hydraulic study of light water reactor as well as for nuclear research reactors. The purpose of this work is to evaluate the code RELAP5/Mod4.0 for analysis of research reactors. This paper consists of three major sections. The first section presents detailed discussions on thermo-physical properties of Lead Bismuth Eutectic (LBE) incorporated in RELAP5/Mod4.0 code. In the second section, benchmarking of RELAP5/Mod4.0 has been done with the Natural Circulation Experimental (NACIE) facility in comparison with Barone’s simulations using RELAP5/Mod3.3. Three different power levels (10.8 kW, 21.7 kW and 32.5 kW) under natural circulation conditions are considered. Results obtained for LBE temperatures, temperature difference across heat section, pin surface temperatures, mass flow rates and heat transfer coefficients in heat section heat exchanger are in agreement with Barone’s simulation results within 7% of average relative error. Third section presents validation of RELAP5/Mod4.0 against the experimental data of NACIE facility performed by Tarantino et al. test number 21 at power of 22.5 kW comparing the profiles of temperatures, mass flow rate and velocity of LBE. Simulation and experimental results agree within 7% of average relative error.

Preliminary Validity of the Eyberg Child Behavior Inventory With Filipino Immigrant Parents.

Science.gov (United States)

Coffey, Dean M; Javier, Joyce R; Schrager, Sheree M

Filipinos are an understudied minority affected by significant behavioral health disparities. We evaluate evidence for the reliability, construct validity, and convergent validity of the Eyberg Child Behavior Inventory (ECBI) in 6- to 12- year old Filipino children ( N = 23). ECBI scores demonstrated high internal consistency, supporting a single-factor model (pre-intervention α =.91; post-intervention α =.95). Results document convergent validity with the Child Behavior Checklist Externalizing scale at pretest ( r = .54, p Filipino children.

National assessment of validity of coding of acute mastoiditis: a standardised reassessment of 1966 records.

Science.gov (United States)

Stalfors, J; Enoksson, F; Hermansson, A; Hultcrantz, M; Robinson, Å; Stenfeldt, K; Groth, A

2013-04-01

To investigate the internal validity of the diagnosis code used at discharge after treatment of acute mastoiditis. Retrospective national re-evaluation study of patient records 1993-2007 and make comparison with the original ICD codes. All ENT departments at university hospitals and one large county hospital department in Sweden. A total of 1966 records were reviewed for patients with ICD codes for in-patient treatment of acute (529), chronic (44) and unspecified mastoiditis (21) and acute otitis media (1372). ICD codes were reviewed by the authors with a defined protocol for the clinical diagnosis of acute mastoiditis. Those not satisfying the diagnosis were given an alternative diagnosis. Of 529 records with ICD coding for acute mastoiditis, 397 (75%) were found to meet the definition of acute mastoiditis used in this study, while 18% were not diagnosed as having any type of mastoiditis after review. Review of the in-patients treated for acute media otitis identified an additional 60 cases fulfilling the definition of acute mastoiditis. Overdiagnosis was common, and many patients with a diagnostic code indicating acute mastoiditis had been treated for external otitis or otorrhoea with transmyringeal drainage. The internal validity of the diagnosis acute mastoiditis is dependent on the use of standardised, well-defined criteria. Reliability of diagnosis is fundamental for the comparison of results from different studies. Inadequate reliability in the diagnosis of acute mastoiditis also affects calculations of incidence rates and statistical power and may also affect the conclusions drawn from the results. © 2013 Blackwell Publishing Ltd.

Validity of the coding for herpes simplex encephalitis in the Danish National Patient Registry

DEFF Research Database (Denmark)

Jørgensen, Laura Krogh; Dalgaard, Lars Skov; Østergaard, Lars Jørgen

2016-01-01

BACKGROUND: Large health care databases are a valuable source of infectious disease epidemiology if diagnoses are valid. The aim of this study was to investigate the accuracy of the recorded diagnosis coding of herpes simplex encephalitis (HSE) in the Danish National Patient Registry (DNPR...... (7.3%) as probable cases providing an overall PPV of 58.0% (95% confidence interval [CI]: 53.0-62.9). For "Encephalitis due to herpes simplex virus" (ICD-10 code B00.4), the PPV was 56.6% (95% CI: 51.1-62.0). Similarly, the PPV for "Meningoencephalitis due to herpes simplex virus" (ICD-10 code B00.4A......) was 56.8% (95% CI: 39.5-72.9). "Herpes viral encephalitis" (ICD-10 code G05.1E) had a PPV of 75.9% (95% CI: 56.5-89.7), thereby representing the highest PPV. The estimated sensitivity was 95.5%. CONCLUSION: The PPVs of the ICD-10 diagnosis coding for adult HSE in the DNPR were relatively low. Hence...

Validation of the 3D finite element transport theory code EVENT for shielding applications

International Nuclear Information System (INIS)

Warner, Paul; Oliveira, R.E. de

2000-01-01

This paper is concerned with the validation of the 3D deterministic neutral-particle transport theory code EVENT for shielding applications. The code is based on the finite element-spherical harmonics (FE-P N ) method which has been extensively developed over the last decade. A general multi-group, anisotropic scattering formalism enables the code to address realistic steady state and time dependent, multi-dimensional coupled neutron/gamma radiation transport problems involving high scattering and deep penetration alike. The powerful geometrical flexibility and competitive computational effort makes the code an attractive tool for shielding applications. In recognition of this, EVENT is currently in the process of being adopted by the UK nuclear industry. The theory behind EVENT is described and its numerical implementation is outlined. Numerical results obtained by the code are compared with predictions of the Monte Carlo code MCBEND and also with the results from benchmark shielding experiments. In particular, results are presented for the ASPIS experimental configuration for both neutron and gamma ray calculations using the BUGLE 96 nuclear data library. (author)

A comparison of oxide thickness predictability from the perspective of codes

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Validation of thermal hydraulic codes for fusion reactors safety

International Nuclear Information System (INIS)

Sardain, P.; Gulden, W.; Massaut, V.; Takase, K.; Merill, B.; Caruso, G.

2006-01-01

A significant effort has been done worldwide on the validation of thermal hydraulic codes, which can be used for the safety assessment of fusion reactors. This work is an item of an implementing agreement under the umbrella of the International Energy Agency. The European part is supported by EFDA. Several programmes related to transient analysis in water-cooled fusion reactors were run in order to assess the capabilities of the codes to treat the main physical phenomena governing the accidental sequences related to water/steam discharge into the vacuum vessel or the cryostat. The typical phenomena are namely the pressurization of a volume at low initial pressure, the critical flow, the flashing, the relief into an expansion volume, the condensation of vapor in a pressure suppression system, the formation of ice on a cryogenic structure, the heat transfer between walls and fluid in various thermodynamic conditions. · A benchmark exercise has been done involving different types of codes, from homogeneous equilibrium to six equations non-equilibrium models. Several cases were defined, each one focusing on a particular phenomenon. · The ICE (Ingress of Coolant Event) facility has been operated in Japan. It has simulated an in-vessel LOCA and the discharge of steam into a pressure suppression system. · The EVITA (European Vacuum Impingement Test Apparatus) facility has been operated in France. It has simulated ingress of coolant into the cryostat, i.e. into a volume at low initial pressure containing surfaces at cryogenic temperature. This paper gives the main lessons gained from these programs, in particular the possibilities for the improvement of the computer codes, extending their capabilities. For example, the water properties have been extended below the triple point. Ice formation models have been implemented. Work has also been done on condensation models. The remaining needs for R-and-D are also highlighted. (author)

Contributions to the validation of advanced codes for accident analysis calculations with 3-dimensional neutron kinetics. STC with the Ukraine. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Rohde, U.; Khalimonchuk, V.; Kuchin, A.; Seidel, A.

2000-10-01

In the frame of a project of scientific-technical cooperation funded by BMBF/BMWi, the coupled code ATHLET-DYN3D has been transferred to the Scientific and Technical Centre on Nuclear and Radiation Safety Kiev (Ukraine). This program code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermohydraulics code system ATHLET. For the purpose of validation of this coupled code, a measurement data base has been generated. In the data base suitable experimental data for operational transients from NPPs are collected. The data collection and documentation was performed in accordance with a directive about requirements to measurement data for code validation, which has been elaborated within the project. The validation calculations have been performed for two selected transients. The results of these calculations were compared with measurement values from the data base. The function of the code DYN3D was expanded with a subroutine for reactivity coefficients calculation. Using this modification of the code DYN3D, investigations of reactivity contributions on different operational processes can be performed. (orig.) [de

Study of the microstructure of neutron irradiated beryllium for the validation of the ANFIBE code

International Nuclear Information System (INIS)

Rabaglino, E.; Ferrero, C.; Reimann, J.; Ronchi, C.; Schulenberg, T.

2002-01-01

The behaviour of beryllium under fast neutron irradiation is a key issue of the helium cooled pebble bed tritium breeding blanket, due to the production of large quantities of helium and of a non-negligible amount of tritium. To optimise the design, a reliable prediction of swelling due to helium bubbles and of tritium inventory during normal and off-normal operation of a fusion power reactor is needed. The ANFIBE code (ANalysis of Fusion Irradiated BEryllium) is being developed to meet this need. The code has to be applied in a range of irradiation conditions where no experimental data are available, therefore a detailed gas kinetics model, and a specific and particularly careful validation strategy are needed. The validation procedure of the first version of the code was based on macroscopic data of swelling and tritium release. This approach is, however, incomplete, since a verification of the microscopic behaviour of the gas in the metal is necessary to obtain a reliable description of swelling. This paper discusses a general strategy for a thorough validation of the gas kinetics models in ANFIBE. The microstructure characterisation of weakly irradiated beryllium pebbles, with different visual examination techniques, is then presented as an example of the application of this strategy. In particular, the advantage of developing 3D techniques, such as X-ray microtomography, is demonstrated

Automation and adaptation: Nurses’ problem-solving behavior following the implementation of bar coded medication administration technology

Science.gov (United States)

Holden, Richard J.; Rivera-Rodriguez, A. Joy; Faye, Héléne; Scanlon, Matthew C.; Karsh, Ben-Tzion

2012-01-01

The most common change facing nurses today is new technology, particularly bar coded medication administration technology (BCMA). However, there is a dearth of knowledge on how BCMA alters nursing work. This study investigated how BCMA technology affected nursing work, particularly nurses’ operational problem-solving behavior. Cognitive systems engineering observations and interviews were conducted after the implementation of BCMA in three nursing units of a freestanding pediatric hospital. Problem-solving behavior, associated problems, and goals, were specifically defined and extracted from observed episodes of care. Three broad themes regarding BCMA’s impact on problem solving were identified. First, BCMA allowed nurses to invent new problem-solving behavior to deal with pre-existing problems. Second, BCMA made it difficult or impossible to apply some problem-solving behaviors that were commonly used pre-BCMA, often requiring nurses to use potentially risky workarounds to achieve their goals. Third, BCMA created new problems that nurses were either able to solve using familiar or novel problem-solving behaviors, or unable to solve effectively. Results from this study shed light on hidden hazards and suggest three critical design needs: (1) ecologically valid design; (2) anticipatory control; and (3) basic usability. Principled studies of the actual nature of clinicians’ work, including problem solving, are necessary to uncover hidden hazards and to inform health information technology design and redesign. PMID:24443642

Validation of coding algorithms for the identification of patients hospitalized for alcoholic hepatitis using administrative data.

Science.gov (United States)

Pang, Jack X Q; Ross, Erin; Borman, Meredith A; Zimmer, Scott; Kaplan, Gilaad G; Heitman, Steven J; Swain, Mark G; Burak, Kelly W; Quan, Hude; Myers, Robert P

2015-09-11

Epidemiologic studies of alcoholic hepatitis (AH) have been hindered by the lack of a validated International Classification of Disease (ICD) coding algorithm for use with administrative data. Our objective was to validate coding algorithms for AH using a hospitalization database. The Hospital Discharge Abstract Database (DAD) was used to identify consecutive adults (≥18 years) hospitalized in the Calgary region with a diagnosis code for AH (ICD-10, K70.1) between 01/2008 and 08/2012. Medical records were reviewed to confirm the diagnosis of AH, defined as a history of heavy alcohol consumption, elevated AST and/or ALT (34 μmol/L, and elevated INR. Subgroup analyses were performed according to the diagnosis field in which the code was recorded (primary vs. secondary) and AH severity. Algorithms that incorporated ICD-10 codes for cirrhosis and its complications were also examined. Of 228 potential AH cases, 122 patients had confirmed AH, corresponding to a positive predictive value (PPV) of 54% (95% CI 47-60%). PPV improved when AH was the primary versus a secondary diagnosis (67% vs. 21%; P codes for ascites (PPV 75%; 95% CI 63-86%), cirrhosis (PPV 60%; 47-73%), and gastrointestinal hemorrhage (PPV 62%; 51-73%) had improved performance, however, the prevalence of these diagnoses in confirmed AH cases was low (29-39%). In conclusion the low PPV of the diagnosis code for AH suggests that caution is necessary if this hospitalization database is used in large-scale epidemiologic studies of this condition.

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Validation of the XLACS code related to contribution of resolved and unresolved resonances and background cross sections

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1990-01-01

The procedures for calculating contributions of resolved and unresolved resonances and background cross sections, in XLACS code, were revised. Constant weighting function and zero Kelvin temperature were considered. Discrepancies found were corrected and now the validated XLACS code generates results that are correct and in accordance with its originally established procedures. (author)

PIV Uncertainty Methodologies for CFD Code Validation at the MIR Facility

Energy Technology Data Exchange (ETDEWEB)

Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Skifton, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stoots, Carl [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kim, Eung Soo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Conder, Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-12-01

Currently, computational fluid dynamics (CFD) is widely used in the nuclear thermal hydraulics field for design and safety analyses. To validate CFD codes, high quality multi dimensional flow field data are essential. The Matched Index of Refraction (MIR) Flow Facility at Idaho National Laboratory has a unique capability to contribute to the development of validated CFD codes through the use of Particle Image Velocimetry (PIV). The significance of the MIR facility is that it permits non intrusive velocity measurement techniques, such as PIV, through complex models without requiring probes and other instrumentation that disturb the flow. At the heart of any PIV calculation is the cross-correlation, which is used to estimate the displacement of particles in some small part of the image over the time span between two images. This image displacement is indicated by the location of the largest peak. In the MIR facility, uncertainty quantification is a challenging task due to the use of optical measurement techniques. Currently, this study is developing a reliable method to analyze uncertainty and sensitivity of the measured data and develop a computer code to automatically analyze the uncertainty/sensitivity of the measured data. The main objective of this study is to develop a well established uncertainty quantification method for the MIR Flow Facility, which consists of many complicated uncertainty factors. In this study, the uncertainty sources are resolved in depth by categorizing them into uncertainties from the MIR flow loop and PIV system (including particle motion, image distortion, and data processing). Then, each uncertainty source is mathematically modeled or adequately defined. Finally, this study will provide a method and procedure to quantify the experimental uncertainty in the MIR Flow Facility with sample test results.

Interviewer and respondent interaction in survey interviews : Empirical evidence from behavior coding studies and question wording experiments

NARCIS (Netherlands)

Ongena, Yfke

2010-01-01

This book sheds light on verbal interaction problems in survey interviews. It is shown how behavior coding, i.e., coding the utterances of interviewer and respondent while they are answering survey questions, can be used to detect interactional problems. Several empirical studies using behavior

Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

Energy Technology Data Exchange (ETDEWEB)

Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

2004-08-01

thermal-hydraulic feedback effects. Thus, in VALCO work package 3 (WP 3) stand-alone three-dimensional neutron-kinetic codes have been validated. Measurements carried out in an original-size VVER-1000 mock-up (V-1000 facility, Kurchatov Institute Moscow) were used for the validation of the codes DYN3D, HEXTRAN, KIKO3D and BIPR-8, which are chiefly designed for VVER safety calculations. The significant neutron flux tilt measured in the V-1000 core, which is caused only by radial-reflector asymmetries, was successfully modelled. A good agreement between calculated and measured steady-state powers has been achieved, for relative assembly powers and inner-assembly pin power distributions. Calculated effective multiplication factors exceed unity in all cases. (orig.)

Code Cactus; Code Cactus

Energy Technology Data Exchange (ETDEWEB)

Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1966-09-01

This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

Validation of the ORIGEN-S code for predicting radionuclide inventories in used CANDU fuel

International Nuclear Information System (INIS)

Tait, J.C.; Gauld, I.; Kerr, A.H.

1995-01-01

The safety assessment being conducted by AECL Research for the concept of deep geological disposal of used CANDU UO 2 fuel requires the calculation of radionuclide inventories in the fuel to provide source terms for radionuclide release. This report discusses the validation of selected actinide and fission-product inventories calculated using the ORIGEN-S code coupled with the WIMS-AECL lattice code, using data from analytical measurements of radioisotope inventories in Pickering CANDU reactor fuel. The recent processing of new ENDF/B-VI cross-section data has allowed the ORIGEN-S calculations to be performed using the most up-to-date nuclear data available. The results indicate that the code is reliably predicting actinide and the majority of fission-product inventories to within the analytical uncertainty. ((orig.))

Classification and Validation of Behavioral Subtypes of Learning-Disabled Children.

Science.gov (United States)

Speece, Deborah L.; And Others

1985-01-01

Using the Classroom Behavior Inventory, teachers rated the behaviors of 63 school-identified, learning-disabled first and second graders. Hierarchical cluster analysis techniques identified seven distinct behavioral subtypes. Internal validation techniques indicated that the subtypes were replicable and had profile patterns different from a sample…

Needs of thermal-hydraulic codes for analyzing hydrogen behavior of future chinese NPPs

International Nuclear Information System (INIS)

Zhiwei Zhou; Jianjun Xiao; Mengjia Yang

2005-01-01

severe accident management guidelines are therefore needed for dealing with both the in-vessel and ex-vessel phenomena, including hydrogen generation, diffusion/convection and deflagration/detonation. To develop the sophisticated thermalhydraulic codes for analyzing severe accident related hydrogen behavior of a light water reactor system is quite expensive and rather unrealistic for China along to bear the cost. Therefore, the most effective way for China to establish the design capability of analyzing severe accident for new nuclear power plant projects is to participate the international or multi-national R and D program, such as EUROATOM cost-sharing program and GEN-IV program, etc. By international cooperation, China can not only gain in most extent the successful experience of the countries with advanced technology in developing nuclear power plants, but also contribute itself most effectively in keeping the momentum of enlarging the peaceful utilization of nuclear energy in the world. Certainly, the future Chinese nuclear power market will be a significant industrial driver for developing the-state-of-the-art thermal-hydraulic codes, including hydrogen behavior analysis codes. This paper also reports some computational study on hydrogen diffusion/convection behavior in the containment related to Daya Bay NPP severe accident analysis with CFD code GASFLOW. The code validation were largely carried out in past few years in Germany and had been applied to EPR and other German NPPs. (authors)

Validation of Mean Drift Forces Computed with the BEM Code NEMOH

DEFF Research Database (Denmark)

Thomsen, Jonas Bjerg

This report covers a simple investigation of mean drift forces found by use of the boundary element method code NEMOH. The results from NEMOH are compared to analytical results from literature and to numerical values found from the commercial software package WADAM by DNV-GL. The work was conduct...... under the project ”Mooring Solutions for Large Wave Energy Converters”, during", Work Package 4: Full Dynamic Analysis". The validation compares results from a simple sphere and from a vertical cylinder....

Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

International Nuclear Information System (INIS)

El bakkari, B.; El Bardouni, T.; Merroun, O.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Chakir, E.

2009-01-01

The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k ∞ ) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.

Validation of the MCNP-DSP Monte Carlo code for calculating source-driven noise parameters of subcritical systems

International Nuclear Information System (INIS)

Valentine, T.E.; Mihalczo, J.T.

1995-01-01

This paper describes calculations performed to validate the modified version of the MCNP code, the MCNP-DSP, used for: the neutron and photon spectra of the spontaneous fission of californium 252; the representation of the detection processes for scattering detectors; the timing of the detection process; and the calculation of the frequency analysis parameters for the MCNP-DSP code

Validation of the WIMSD4M cross-section generation code with benchmark results

International Nuclear Information System (INIS)

Deen, J.R.; Woodruff, W.L.; Leal, L.E.

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D 2 O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented

Validation of the WIMSD4M cross-section generation code with benchmark results

Energy Technology Data Exchange (ETDEWEB)

Deen, J.R.; Woodruff, W.L. [Argonne National Lab., IL (United States); Leal, L.E. [Oak Ridge National Lab., TN (United States)

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.

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

International Nuclear Information System (INIS)

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

2010-07-01

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

Construct Validation of the Behavior and Instructional Management Scale

Science.gov (United States)

Martin, Nancy K.; Sass, Daniel A.

2010-01-01

Beliefs related to classroom management vary among teachers and play an important role in classrooms. Despite the importance of this construct, valid measures have proven difficult to develop. This study evaluated the psychometric properties of the Behavior and Instructional Management Scale (BIMS), a short but valid measure of teachers'…

Improvement of a combustion model in MELCOR code

International Nuclear Information System (INIS)

Ogino, Masao; Hashimoto, Takashi

1999-01-01

NUPEC has been improving a hydrogen combustion model in MELCOR code for severe accident analysis. In the proposed combustion model, the flame velocity in a node was predicted using five different flame front shapes of fireball, prism, bubble, spherical jet, and plane jet. For validation of the proposed model, the results of the Battelle multi-compartment hydrogen combustion test were used. The selected test cases for the study were Hx-6, 13, 14, 20 and Ix-2 which had two, three or four compartments under homogeneous hydrogen concentration of 5 to 10 vol%. The proposed model could predict well the combustion behavior in multi-compartment containment geometry on the whole. MELCOR code, incorporating the present combustion model, can simulate combustion behavior during severe accident with acceptable computing time and some degree of accuracy. The applicability study of the improved MELCOR code to the actual reactor plants will be further continued. (author)

AEEW comments on the NNC/CEGB LOCA code validation report RX 440-A

International Nuclear Information System (INIS)

Brittain, I.; Bryce, W.M.; O'Mahoney, R.; Richards, C.G.; Gibson, I.H.; Porter, W.H.L.; Fell, J.

1984-03-01

Comments are made on the NNC/CEGB report PWR/RX 440-A, Review of Validation for the ECCS Evaluation Model Codes, by K.T. Routledge et al, 1982. This set out to review methods and models used in the LOCA safety case for Sizewell B. These methods are embodied in the Evaluation Model Computer codes SATAN-VI, WREFLOOD, WFLASH, LOCTA-IV and COCO. The main application of these codes is the determination of peak clad temperature and overall containment pressure. The comments represent the views of a group which has been involved for a number of years in the development and application of Best-Estimate methods for LOCA analysis. It is the judgement of this group that, overall, the EM methods can be used to make an acceptable safety case, but there are a number of points of detail still to be resolved. (U.K.)

Validation of Code ASTEC with LIVE-L1 Experimental Results

International Nuclear Information System (INIS)

Bachrata, Andrea

2008-01-01

The severe accidents with core melting are considered at the design stage of project at Generation 3+ of Nuclear Power Plants (NPP). Moreover, there is an effort to apply the severe accident management to the operated NPP. The one of main goals of severe accidents mitigation is corium localization and stabilization. The two strategies that fulfil this requirement are: the in-vessel retention (e.g. AP-600, AP- 1000) and the ex-vessel retention (e.g. EPR). To study the scenario of in-vessel retention, a large experimental program and the integrated codes have been developed. The LIVE-L1 experimental facility studied the formation of melt pools and the melt accumulation in the lower head using different cooling conditions. Nowadays, a new European computer code ASTEC is being developed jointly in France and Germany. One of the important steps in ASTEC development in the area of in-vessel retention of corium is its validation with LIVE-L1 experimental results. Details of the experiment are reported. Results of the ASTEC (module DIVA) application to the analysis of the test are presented. (author)

Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

International Nuclear Information System (INIS)

Wright, A.L.; Wilson, J.H.; Arwood, P.C.

1986-01-01

The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) and ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models

Phenomenological modeling of critical heat flux: The GRAMP code and its validation

International Nuclear Information System (INIS)

Ahmad, M.; Chandraker, D.K.; Hewitt, G.F.; Vijayan, P.K.; Walker, S.P.

2013-01-01

Highlights: ► Assessment of CHF limits is vital for LWR optimization and safety analysis. ► Phenomenological modeling is a valuable adjunct to pure empiricism. ► It is based on empirical representations of the (several, competing) phenomena. ► Phenomenological modeling codes making ‘aggregate’ predictions need careful assessment against experiments. ► The physical and mathematical basis of a phenomenological modeling code GRAMP is presented. ► The GRAMP code is assessed against measurements from BARC (India) and Harwell (UK), and the Look Up Tables. - Abstract: Reliable knowledge of the critical heat flux is vital for the design of light water reactors, for both safety and optimization. The use of wholly empirical correlations, or equivalently “Look Up Tables”, can be very effective, but is generally less so in more complex cases, and in particular cases where the heat flux is axially non-uniform. Phenomenological models are in principle more able to take into account of a wider range of conditions, with a less comprehensive coverage of experimental measurements. These models themselves are in part based upon empirical correlations, albeit of the more fundamental individual phenomena occurring, rather than the aggregate behaviour, and as such they too require experimental validation. In this paper we present the basis of a general-purpose phenomenological code, GRAMP, and then use two independent ‘direct’ sets of measurement, from BARC in India and from Harwell in the United Kingdom, and the large dataset embodied in the Look Up Tables, to perform a validation exercise on it. Very good agreement between predictions and experimental measurements is observed, adding to the confidence with which the phenomenological model can be used. Remaining important uncertainties in the phenomenological modeling of CHF, namely the importance of the initial entrained fraction on entry to annular flow, and the influence of the heat flux on entrainment rate

Improvements and validation of the transient analysis code MOREL for molten salt reactors

International Nuclear Information System (INIS)

Zhuang Kun; Zheng Youqi; Cao Liangzhi; Hu Tianliang; Wu Hongchun

2017-01-01

The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective. (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)

Validation of the ORIGEN-S code for predicting radionuclide inventories in used CANDU Fuel

International Nuclear Information System (INIS)

Tait, J.C.; Gauld, I.; Kerr, A.H.

1994-10-01

The safety assessment being conducted by AECL Research for the concept of deep geological disposal of used CANDU UO 2 fuel requires the calculation of radionuclide inventories in the fuel to provide source terms for radionuclide release. This report discusses the validation of selected actinide and fission-product inventories calculated using the ORIGEN-S code coupled with the WIMS-AECL lattice code, using data from analytical measurements of radioisotope inventories in Pickering CANDU reactor fuel. The recent processing of new ENDF/B-VI cross-section data has allowed the ORIGEN-S calculations to be performed using the most up-to-date nuclear data available. The results indicate that the code is reliably predicting actinide and the majority of fission-product inventories to within the analytical uncertainty. 38 refs., 4 figs., 5 tabs

Validation of the assert subchannel code: Prediction of CHF in standard and non-standard Candu bundle geometries

International Nuclear Information System (INIS)

Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.

1993-01-01

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of prediting CHF at these local conditions, makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries

Development of in-vessel source term analysis code, tracer

International Nuclear Information System (INIS)

Miyagi, K.; Miyahara, S.

1996-01-01

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

Development and Validation of Children's Responsible Environmental Behavior Scale

Science.gov (United States)

Erdogan, Mehmet; Ok, Ahmet; Marcinkowski, Thomas Joseph

2012-01-01

Though environmentally responsible behavior (ERB) has been a focus of many studies in the field of environmental education, very few scales have been developed to assess children's ERB. In this regard, this article focuses on the development and validation of Children's Responsible Environmental Behavior Scale (CREBS) and also reports the…

Evaluation and validation of criticality codes for fuel dissolver calculations

International Nuclear Information System (INIS)

Santamarina, A.; Smith, H.J.; Whitesides, G.E.

1991-01-01

During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. The spread of the results in the international calculation amounted to ± 12,000 pcm in the realistic fuel dissolver exercise n degrees 19 proposed by BNFL, and to ± 25,000 pcm in the benchmark n degrees 20 in which fissile material in solid form is surrounded by fissile material in solution. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat latter effect, permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates solicited from the participants. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism (NITAWL in the international SCALE package) to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. Improvements in the up-dated 1990 contributions, as do recent complementary reference calculations (MCNP, VIM, ultrafine slowing-down CGM calculation), confirm the need to use rigorous self-shielding methods in criticality design-oriented codes. 6 refs., 11 figs., 3 tabs

Behavioral correlates of the distributed coding of spatial context.

Science.gov (United States)

Anderson, Michael I; Killing, Sarah; Morris, Caitlin; O'Donoghue, Alan; Onyiagha, Dikennam; Stevenson, Rosemary; Verriotis, Madeleine; Jeffery, Kathryn J

2006-01-01

Hippocampal place cells respond heterogeneously to elemental changes of a compound spatial context, suggesting that they form a distributed code of context, whereby context information is shared across a population of neurons. The question arises as to what this distributed code might be useful for. The present study explored two possibilities: one, that it allows contexts with common elements to be disambiguated, and the other, that it allows a given context to be associated with more than one outcome. We used two naturalistic measures of context processing in rats, rearing and thigmotaxis (boundary-hugging), to explore how rats responded to contextual novelty and to relate this to the behavior of place cells. In experiment 1, rats showed dishabituation of rearing to a novel reconfiguration of familiar context elements, suggesting that they perceived the reconfiguration as novel, a behavior that parallels that of place cells in a similar situation. In experiment 2, rats were trained in a place preference task on an open-field arena. A change in the arena context triggered renewed thigmotaxis, and yet navigation continued unimpaired, indicating simultaneous representation of both the altered contextual and constant spatial cues. Place cells similarly exhibited a dual population of responses, consistent with the hypothesis that their activity underlies spatial behavior. Together, these experiments suggest that heterogeneous context encoding (or "partial remapping") by place cells may function to allow the flexible assignment of associations to contexts, a faculty that could be useful in episodic memory encoding. Copyright (c) 2006 Wiley-Liss, Inc.

Development validation and use of computer codes for inelastic analysis

International Nuclear Information System (INIS)

Jobson, D.A.

1983-01-01

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

Predicting risk behaviors: development and validation of a diagnostic scale.

Science.gov (United States)

Witte, K; Cameron, K A; McKeon, J K; Berkowitz, J M

1996-01-01

The goal of this study was to develop and validate the Risk Behavior Diagnosis (RBD) Scale for use by health care providers and practitioners interested in promoting healthy behaviors. Theoretically guided by the Extended Parallel Process Model (EPPM; a fear appeal theory), the RBD scale was designed to work in conjunction with an easy-to-use formula to determine which types of health risk messages would be most appropriate for a given individual or audience. Because some health risk messages promote behavior change and others backfire, this type of scale offers guidance to practitioners on how to develop the best persuasive message possible to motivate healthy behaviors. The results of the study demonstrate the RBD scale to have a high degree of content, construct, and predictive validity. Specific examples and practical suggestions are offered to facilitate use of the scale for health practitioners.

A Correlational Study: Code of Ethics in Testing and EFL Instructors' Professional Behavior

Science.gov (United States)

Ashraf, Hamid; Kafi, Zahra; Saeedan, Azaam

2018-01-01

The present study has aimed at delving the code of ethics in testing in English language institutions to see how far adhering to these ethical codes will result in EFL teachers' professional behavior. Therefore, 300 EFL instructors teaching at English language schools in Khorasan Razavi Province, Zabansara Language School, as well as Khorasan…

Under-coding of secondary conditions in coded hospital health data: Impact of co-existing conditions, death status and number of codes in a record.

Science.gov (United States)

Peng, Mingkai; Southern, Danielle A; Williamson, Tyler; Quan, Hude

2017-12-01

This study examined the coding validity of hypertension, diabetes, obesity and depression related to the presence of their co-existing conditions, death status and the number of diagnosis codes in hospital discharge abstract database. We randomly selected 4007 discharge abstract database records from four teaching hospitals in Alberta, Canada and reviewed their charts to extract 31 conditions listed in Charlson and Elixhauser comorbidity indices. Conditions associated with the four study conditions were identified through multivariable logistic regression. Coding validity (i.e. sensitivity, positive predictive value) of the four conditions was related to the presence of their associated conditions. Sensitivity increased with increasing number of diagnosis code. Impact of death on coding validity is minimal. Coding validity of conditions is closely related to its clinical importance and complexity of patients' case mix. We recommend mandatory coding of certain secondary diagnosis to meet the need of health research based on administrative health data.

Validation and optimisation of an ICD-10-coded case definition for sepsis using administrative health data

Science.gov (United States)

Jolley, Rachel J; Jetté, Nathalie; Sawka, Keri Jo; Diep, Lucy; Goliath, Jade; Roberts, Derek J; Yipp, Bryan G; Doig, Christopher J

2015-01-01

Objective Administrative health data are important for health services and outcomes research. We optimised and validated in intensive care unit (ICU) patients an International Classification of Disease (ICD)-coded case definition for sepsis, and compared this with an existing definition. We also assessed the definition's performance in non-ICU (ward) patients. Setting and participants All adults (aged ≥18 years) admitted to a multisystem ICU with general medicosurgical ICU care from one of three tertiary care centres in the Calgary region in Alberta, Canada, between 1 January 2009 and 31 December 2012 were included. Research design Patient medical records were randomly selected and linked to the discharge abstract database. In ICU patients, we validated the Canadian Institute for Health Information (CIHI) ICD-10-CA (Canadian Revision)-coded definition for sepsis and severe sepsis against a reference standard medical chart review, and optimised this algorithm through examination of other conditions apparent in sepsis. Measures Sensitivity (Sn), specificity (Sp), positive predictive value (PPV) and negative predictive value (NPV) were calculated. Results Sepsis was present in 604 of 1001 ICU patients (60.4%). The CIHI ICD-10-CA-coded definition for sepsis had Sn (46.4%), Sp (98.7%), PPV (98.2%) and NPV (54.7%); and for severe sepsis had Sn (47.2%), Sp (97.5%), PPV (95.3%) and NPV (63.2%). The optimised ICD-coded algorithm for sepsis increased Sn by 25.5% and NPV by 11.9% with slightly lowered Sp (85.4%) and PPV (88.2%). For severe sepsis both Sn (65.1%) and NPV (70.1%) increased, while Sp (88.2%) and PPV (85.6%) decreased slightly. Conclusions This study demonstrates that sepsis is highly undercoded in administrative data, thus under-ascertaining the true incidence of sepsis. The optimised ICD-coded definition has a higher validity with higher Sn and should be preferentially considered if used for surveillance purposes. PMID:26700284

Understanding Engagement in Dementia Through Behavior. The Ethographic and Laban-Inspired Coding System of Engagement (ELICSE and the Evidence-Based Model of Engagement-Related Behavior (EMODEB

Directory of Open Access Journals (Sweden)

Giulia Perugia

2018-05-01

Full Text Available Engagement in activities is of crucial importance for people with dementia. State of the art assessment techniques rely exclusively on behavior observation to measure engagement in dementia. These techniques are either too general to grasp how engagement is naturally expressed through behavior or too complex to be traced back to an overall engagement state. We carried out a longitudinal study to develop a coding system of engagement-related behavior that could tackle these issues and to create an evidence-based model of engagement to make meaning of such a coding system. Fourteen elderlies with mild to moderate dementia took part in the study. They were involved in two activities: a game-based cognitive stimulation and a robot-based free play. The coding system was developed with a mixed approach: ethographic and Laban-inspired. First, we developed two ethograms to describe the behavior of participants in the two activities in detail. Then, we used Laban Movement Analysis (LMA to identify a common structure to the behaviors in the two ethograms and unify them in a unique coding system. The inter-rater reliability (IRR of the coding system proved to be excellent for cognitive games (kappa = 0.78 and very good for robot play (kappa = 0.74. From the scoring of the videos, we developed an evidence-based model of engagement. This was based on the most frequent patterns of body part organization (i.e., the way body parts are connected in movement observed during activities. Each pattern was given a meaning in terms of engagement by making reference to the literature. The model was tested using structural equation modeling (SEM. It achieved an excellent goodness of fit and all the hypothesized relations between variables were significant. We called the coding system that we developed the Ethographic and Laban-Inspired Coding System of Engagement (ELICSE and the model the Evidence-based Model of Engagement-related Behavior (EMODEB. To the best of our

Evaluation and validation of criticality codes for fuel dissolver calculations

International Nuclear Information System (INIS)

Santamarina, A.; Smith, H.J.; Whitesides, G.E.

1991-01-01

During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)

RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

Energy Technology Data Exchange (ETDEWEB)

Talley, Darren G.

2017-04-01

This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.

Drugs, Guns, and Disadvantaged Youths: Co-Occurring Behavior and the Code of the Street

Science.gov (United States)

Allen, Andrea N.; Lo, Celia C.

2012-01-01

Guided by Anderson's theory of the code of the street, this study explored social mechanisms linking individual-level disadvantage factors with the adoption of beliefs grounded in the code of the street and with drug trafficking and gun carrying--the co-occurring behavior shaping violence among young men in urban areas. Secondary data were…

FEMAXI-III: a computer code for the analysis of thermal and mechanical behavior of fuel rods

International Nuclear Information System (INIS)

Nakajima, Tetsuo; Ichikawa, Michio; Iwano, Yoshihiko; Ito, Kenichi; Saito, Hiroaki; Kashima, Koichi; Kinoshita, Motoyasu; Okubo, Tadatsune.

1985-12-01

FEMAXI-III is a computer code to predict the thermal and mechanical behavior of a light water fuel rod during its irradiation life. It can analyze the integral behavior of a whole fuel rod throughout its life, as well as the localized behavior of a small part of fuel rod. The localized mechanical behavior such as the cladding ridge deformation is analyzed by the two-dimensional axisymmetric finite element method. FEMAXI-III calculates, in particular, the temperature distribution, the radial deformation, the fission gas release, and the inner gas pressure as a function of irradiation time and axial position, and the stresses and strains in the fuel and cladding at a small part of fuel rod as a function of irradiation time. For this purpose, Elasto-plasticity, creep, thermal expansion, fuel cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, and fuel-cladding mechanical interaction are modelled and their interconnected effects are considered in the code. Efforts have been made to improve the accuracy and stability of finite element solution and to minimize the computer memory and running time. This report describes the outline of the code and the basic models involved, and also includes the application of the code and its input manual. (author)

VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

Directory of Open Access Journals (Sweden)

NAM-IL TAK

2013-11-01

Full Text Available For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR, intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI and the AGREE code of the University of Michigan (U of M. One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

Development of CAP code for nuclear power plant containment: Lumped model

Energy Technology Data Exchange (ETDEWEB)

Hong, Soon Joon, E-mail: sjhong90@fnctech.com [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Ha, Sang Jun [Central Research Institute, Korea Hydro & Nuclear Power Company, Ltd., 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)

2015-09-15

Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP.

Development of CAP code for nuclear power plant containment: Lumped model

International Nuclear Information System (INIS)

Hong, Soon Joon; Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul; Ha, Sang Jun

2015-01-01

Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP

Development of Creative Behavior Observation Form: A Study on Validity and Reliability

Science.gov (United States)

Dere, Zeynep; Ömeroglu, Esra

2018-01-01

This study, Creative Behavior Observation Form was developed to assess creativity of the children. While the study group on the reliability and validity of Creative Behavior Observation Form was being developed, 257 children in total who were at the ages of 5-6 were used as samples with stratified sampling method. Content Validity Index (CVI) and…

Development and validation of gui based input file generation code for relap

International Nuclear Information System (INIS)

Anwar, M.M.; Khan, A.A.; Chughati, I.R.; Chaudri, K.S.; Inyat, M.H.; Hayat, T.

2009-01-01

Reactor Excursion and Leak Analysis Program (RELAP) is a widely acceptable computer code for thermal hydraulics modeling of Nuclear Power Plants. It calculates thermal- hydraulic transients in water-cooled nuclear reactors by solving approximations to the one-dimensional, two-phase equations of hydraulics in an arbitrarily connected system of nodes. However, the preparation of input file and subsequent analysis of results in this code is a tedious task. The development of a Graphical User Interface (GUI) for preparation of the input file for RELAP-5 is done with the validation of GUI generated Input File. The GUI is developed in Microsoft Visual Studio using Visual C Sharp (C) as programming language. The Nodalization diagram is drawn graphically and the program contains various component forms along with the starting data form, which are launched for properties assignment to generate Input File Cards serving as GUI for the user. The GUI is provided with Open / Save function to store and recall the Nodalization diagram along with Components' properties. The GUI generated Input File is validated for several case studies and individual component cards are compared with the originally required format. The generated Input File of RELAP is found consistent with the requirement of RELAP. The GUI provided a useful platform for simulating complex hydrodynamic problems efficiently with RELAP. (author)

Decay heat experiment and validation of calculation code systems for fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki

1999-10-01

Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)

Decay heat experiment and validation of calculation code systems for fusion reactor

International Nuclear Information System (INIS)

Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki

1999-10-01

Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)

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

Validation of the TRACR3D code for soil water flow under saturated/unsaturated conditions in three experiments

International Nuclear Information System (INIS)

Perkins, B.; Travis, B.; DePoorter, G.

1985-01-01

Validation of the TRACR3D code in a one-dimensional form was obtained for flow of soil water in three experiments. In the first experiment, a pulse of water entered a crushed-tuff soil and initially moved under conditions of saturated flow, quickly followed by unsaturated flow. In the second experiment, steady-state unsaturated flow took place. In the final experiment, two slugs of water entered crushed tuff under field conditions. In all three experiments, experimentally measured data for volumetric water content agreed, within experimental errors, with the volumetric water content predicted by the code simulations. The experiments and simulations indicated the need for accurate knowledge of boundary and initial conditions, amount and duration of moisture input, and relevant material properties as input into the computer code. During the validation experiments, limitations on monitoring of water movement in waste burial sites were also noted. 5 references, 34 figures, 9 tables

Brief implicit association test: Validity and utility in prediction of voting behavior

Directory of Open Access Journals (Sweden)

Pavlović Maša D.

2013-01-01

Full Text Available We employed the Brief Implicit Association Test (a recently developed short version of IAT to measure implicit political attitudes toward four political parties running for Serbian parliament. To test its criterion validity, we measured voting intention and actual voting behavior. In addition, we introduced political involvement as a potential moderator of the BIAT’s predictive and incremental validity. The BIAT demonstrated good internal and predictive validity, but lacked incremental validity over self-report measures. Predictive power of the BIAT was moderated by political involvement - the BIAT scores were stronger predictors of voting intention and behavior among voters highly involved in politics. [Projekat Ministarstva nauke Republike Srbije, br. 179018

Validation of the ASSERT subchannel code for MAPLE-X10 reactor conditions

International Nuclear Information System (INIS)

Carver, M.B.; Kiteley, J.C.; Junop, S.V.; Wasilewicz, J.F.

1993-01-01

The ASSERT subchannel analysis code has been developed specifically to model flow and phase distributions within CANDU fuel channels. Recently, ASSERT has been adapted for use in simulating the MAPLE-X10 reactor. ASSERT uses an advanced drift-flux model, which permits the phases to have unequal velocities and unequal temperatures (UVUT), and thus can model non-equilibrium effects such as phase separation tendencies and subcooled boiling. Modelling subcooled boiling accurately is particularly important for MAPLE-X10. This paper briefly summarizes the non-equilibrium model used in the ASSERT code, the equations used to represent these models, and the algorithms used to solve the equations numerically. Very few modifications to the ASSERT models were needed to address MAPLE conditions. These centered on the manner in which finned fuel rods are treated, and they are discussed in the paper. The paper also gives results from validation exercises, in which the ASSERT code predictions of subcooled boiling void-fraction and critical heat flux were compared to experiments using MAPLE-X10 finned fuel elements in annuli and various bundles. 18 refs., 13 figs., 3 tabs

Validation of CATHARE 3D code against UPTF TRAM C3 transients

International Nuclear Information System (INIS)

Glantz, Tony; Freitas, Roberto

2007-01-01

Within the nuclear reactor safety analysis, one of the events that could potentially lead to a recriticality accident in case of a Small Break LOCA (SBLOCA) in a pressurized water reactor (PWR) is a boron dilution scenario followed by a coolant mixing transient. Some UPTF experiments can be interpreted as generic boron dilution experiments. In fact, the UPTF experiments were originally designed to conduct separate effects studies focused on multi-dimensional thermal hydraulic phenomena. But, in the case of experimental program TRAM, some studies are realized on the boron mixing: tests C3. Some of these tests have been used for the validation and assessment of the 3D module of CATHARE code. Results are very satisfying; CATHARE 3D code is able to reproduce correctly the main features of the UPTF TRAM C3 tests, the temperature mixing in the cold leg, the formation of a strong stratification in the upper downcomer, the perfect mixing temperature in the lower downcomer and the strong stratification in the lower plenum. These results are also compared with the CFX-5 and TRIO-U codes results on these tests. (author)

Validations of BWR nuclear design code using ABWR MOX numerical benchmark problems

International Nuclear Information System (INIS)

Takano, Shou; Sasagawa, Masaru; Yamana, Teppei; Ikehara, Tadashi; Yanagisawa, Naoki

2017-01-01

BWR core design code package (the HINES assembly code and the PANACH core simulator), being used for full MOX-ABWR core design, has been benchmarked against the high-fidelity numerical solutions as references, for the purpose of validating its capability of predicting the BWR core design parameters systematically from UO 2 to 100% MOX cores. The reference solutions were created by whole core critical calculations using MCNPs with the precisely modeled ABWR cores both in hot and cold conditions at BOC and EOC of the equilibrium cycle. A Doppler-Broadening Rejection Correction (DCRB) implemented MCNP5-1.4 with ENDF/B-VII.0 was mainly used to evaluate the core design parameters, except for effective delayed neutron fraction (β eff ) and prompt neutron lifetime (l) with MCNP6.1. The discrepancies in the results between the design codes HINES-PANACH and MCNPs for the core design parameters such as the bundle powers, hot pin powers, control rod worth, boron worth, void reactivity, Doppler reactivity, β eff and l, are almost within target accuracy, leading to the conclusion that HINES-PANACH has sufficient fidelity for application to full MOX-ABWR core design. (author)

MISTRA facility for containment lumped parameter and CFD codes validation. Example of the International Standard Problem ISP47

International Nuclear Information System (INIS)

Tkatschenko, I.; Studer, E.; Paillere, H.

2005-01-01

During a severe accident in a Pressurized Water Reactor (PWR), the formation of a combustible gas mixture in the complex geometry of the reactor depends on the understanding of hydrogen production, the complex 3D thermal-hydraulics flow due to gas/steam injection, natural convection, heat transfer by condensation on walls and effect of mitigation devices. Numerical simulation of such flows may be performed either by Lumped Parameter (LP) or by Computational Fluid Dynamics (CFD) codes. Advantages and drawbacks of LP and CFD codes are well-known. LP codes are mainly developed for full size containment analysis but they need improvements, especially since they are not able to accurately predict the local gas mixing within the containment. CFD codes require a process of validation on well-instrumented experimental data before they can be used with a high degree of confidence. The MISTRA coupled effect test facility has been built at CEA to fulfil this validation objective: with numerous measurement points in the gaseous volume - temperature, gas concentration, velocity and turbulence - and with well controlled boundary conditions. As illustration of both experimental and simulation areas of this topic, a recent example in the use of MISTRA test data is presented for the case of the International Standard Problem ISP47. The proposed experimental work in the MISTRA facility provides essential data to fill the gaps in the modelling/validation of computational tools. (author)

Validation of the computer code system ATHLET / ATHLET-CD. Final report; Validierung des Rechenprogrammsystems ATHLET / ATHLET-CD. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Austregesilo, H.; Bals, C.; Erdmann, W.; Horche, W.; Krzykacz-Hausmann, B.; Pointner, W.; Schoeffel, P.; Skorek, T.; Weber, S.; Wielenberg, A.

2010-04-15

In the frame of the reactor safety project RS1173, sponsored by the German Federal Ministry of Economics and Technology, analyses of international integral and separate effects tests have been performed for the validation of the code system ATHLET/ATHLET-CD. The work mainly comprised post-test calculations of selected experiments and the contributions to the working groups accompanying the experimental programs. For the assessment of the thermal-hydraulic models in ATHLET 8 integral tests and 4 separate effect tests have been considered. Together with the corroboration of the existing models, the validation analyses were mainly dedicated to the assessment of the modelling of non-condensable gases and their influence on two-phase natural circulation and on the primary heat removal through steam generators, as well as of the simulation of multi-dimensional flow processes. The validation calculations with respect to the simulation of multi-dimensional one- and two-phase flows aimed to investigate the range of applicability and limitations of the method of parallel channels in connection with the separate momentum equations for water and steam current used in ATHLET as well as to assess the status of the coupled version ATHLET/FLUBOX-3D. The ATHLET-CD validation analyses included the post-test calculations of 9 bundle tests, and was mainly focussed on the assessment of the improved and new models for core degradation, including the models for oxidation, melt formation and relocation for BWR components, as well as of the modelling of fission products and aerosol transport within the primary circuit taking into account chemical reactions within the module SOPHAEROS. As an additional contribution to code validation, the GRS methodology of uncertainty and sensitivity analysis was applied exemplarily to two validation calculations, one with ATHLET and one with ATHLET-CD. The results of these uncertainty analyses endorse the capability of the code system to reproduce

Assessing Knowledge Sharing Among Academics: A Validation of the Knowledge Sharing Behavior Scale (KSBS).

Science.gov (United States)

Ramayah, T; Yeap, Jasmine A L; Ignatius, Joshua

2014-04-01

There is a belief that academics tend to hold on tightly to their knowledge and intellectual resources. However, not much effort has been put into the creation of a valid and reliable instrument to measure knowledge sharing behavior among the academics. To apply and validate the Knowledge Sharing Behavior Scale (KSBS) as a measure of knowledge sharing behavior within the academic community. Respondents (N = 447) were academics from arts and science streams in 10 local, public universities in Malaysia. Data were collected using the 28-item KSBS that assessed four dimensions of knowledge sharing behavior namely written contributions, organizational communications, personal interactions, and communities of practice. The exploratory factor analysis showed that the items loaded on the dimension constructs that they were supposed to represent, thus proving construct validity. A within-factor analysis revealed that each set of items representing their intended dimension loaded on only one construct, therefore establishing convergent validity. All four dimensions were not perfectly correlated with each other or organizational citizenship behavior, thereby proving discriminant validity. However, all four dimensions correlated with organizational commitment, thus confirming predictive validity. Furthermore, all four factors correlated with both tacit and explicit sharing, which confirmed their concurrent validity. All measures also possessed sufficient reliability (α > .70). The KSBS is a valid and reliable instrument that can be used to formally assess the types of knowledge artifacts residing among academics and the degree of knowledge sharing in relation to those artifacts. © The Author(s) 2014.

Validation of the ASSERT subchannel code for prediction of CHF in standard and non-standard CANDU bundle geometries

International Nuclear Information System (INIS)

Kiteley, J.C.; Carver, M.B.; Zhou, Q.N.

1993-01-01

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting critical heat flux (CHF) at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is the only tool available to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries. 28 refs., 12 figs

Validation of the ASSERT subchannel code: Prediction of critical heat flux in standard and nonstandard CANDU bundle geometries

International Nuclear Information System (INIS)

Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.

1995-01-01

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized

Three-dimensional all-speed CFD code for safety analysis of nuclear reactor containment: Status of GASFLOW parallelization, model development, validation and application

Energy Technology Data Exchange (ETDEWEB)

Xiao, Jianjun, E-mail: jianjun.xiao@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Travis, John R., E-mail: jack_travis@comcast.com [Engineering and Scientific Software Inc., 3010 Old Pecos Trail, Santa Fe, NM 87505 (United States); Royl, Peter, E-mail: peter.royl@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Necker, Gottfried, E-mail: gottfried.necker@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Svishchev, Anatoly, E-mail: anatoly.svishchev@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Jordan, Thomas, E-mail: thomas.jordan@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

2016-05-15

Highlights: • 3-D scalable semi-implicit pressure-based CFD code for containment safety analysis. • Robust solution algorithm valid for all-speed flows. • Well validated and widely used CFD code for hydrogen safety analysis. • Code applied in various types of nuclear reactor containments. • Parallelization enables high-fidelity models in large scale containment simulations. - Abstract: GASFLOW is a three dimensional semi-implicit all-speed CFD code which can be used to predict fluid dynamics, chemical kinetics, heat and mass transfer, aerosol transportation and other related phenomena involved in postulated accidents in nuclear reactor containments. The main purpose of the paper is to give a brief review on recent GASFLOW code development, validations and applications in the field of nuclear safety. GASFLOW code has been well validated by international experimental benchmarks, and has been widely applied to hydrogen safety analysis in various types of nuclear power plants in European and Asian countries, which have been summarized in this paper. Furthermore, four benchmark tests of a lid-driven cavity flow, low Mach number jet flow, 1-D shock tube and supersonic flow over a forward-facing step are presented in order to demonstrate the accuracy and wide-ranging capability of ICE’d ALE solution algorithm for all-speed flows. GASFLOW has been successfully parallelized using the paradigms of Message Passing Interface (MPI) and domain decomposition. The parallel version, GASFLOW-MPI, adds great value to large scale containment simulations by enabling high-fidelity models, including more geometric details and more complex physics. It will be helpful for the nuclear safety engineers to better understand the hydrogen safety related physical phenomena during the severe accident, to optimize the design of the hydrogen risk mitigation systems and to fulfill the licensing requirements by the nuclear regulatory authorities. GASFLOW-MPI is targeting a high

Validation of ICD-9-CM coding algorithm for improved identification of hypoglycemia visits

Directory of Open Access Journals (Sweden)

Lieberman Rebecca M

2008-04-01

Full Text Available Abstract Background Accurate identification of hypoglycemia cases by International Classification of Diseases, Ninth Revision, Clinical Modification (ICD-9-CM codes will help to describe epidemiology, monitor trends, and propose interventions for this important complication in patients with diabetes. Prior hypoglycemia studies utilized incomplete search strategies and may be methodologically flawed. We sought to validate a new ICD-9-CM coding algorithm for accurate identification of hypoglycemia visits. Methods This was a multicenter, retrospective cohort study using a structured medical record review at three academic emergency departments from July 1, 2005 to June 30, 2006. We prospectively derived a coding algorithm to identify hypoglycemia visits using ICD-9-CM codes (250.3, 250.8, 251.0, 251.1, 251.2, 270.3, 775.0, 775.6, and 962.3. We confirmed hypoglycemia cases by chart review identified by candidate ICD-9-CM codes during the study period. The case definition for hypoglycemia was documented blood glucose 3.9 mmol/l or emergency physician charted diagnosis of hypoglycemia. We evaluated individual components and calculated the positive predictive value. Results We reviewed 636 charts identified by the candidate ICD-9-CM codes and confirmed 436 (64% cases of hypoglycemia by chart review. Diabetes with other specified manifestations (250.8, often excluded in prior hypoglycemia analyses, identified 83% of hypoglycemia visits, and unspecified hypoglycemia (251.2 identified 13% of hypoglycemia visits. The absence of any predetermined co-diagnosis codes improved the positive predictive value of code 250.8 from 62% to 92%, while excluding only 10 (2% true hypoglycemia visits. Although prior analyses included only the first-listed ICD-9 code, more than one-quarter of identified hypoglycemia visits were outside this primary diagnosis field. Overall, the proposed algorithm had 89% positive predictive value (95% confidence interval, 86–92 for

GOTHIC-IST 6.1b code validation exercises relating to heat removal by dousing and air coolers in CANDU containment

International Nuclear Information System (INIS)

Ramachandran, S.; Krause, M.; Nguyen, T.

2003-01-01

This paper presents validation results relating to the use of the GOTHIC containment analysis code for CANDU safety analysis. The validation results indicate that GOTHIC predicts heat removal by dousing and air cooler heat transfer with reasonable accuracy. (author)

Verification and Validation of Monte Carlo n-Particle Code 6 (MCNP6) with Neutron Protection Factor Measurements of an Iron Box

Science.gov (United States)

2014-03-27

Vehicle Code System (VCS), the Monte Carlo Adjoint SHielding (MASH), and the Monte Carlo n- Particle ( MCNP ) code. Of the three, the oldest and still most...widely utilized radiation transport code is MCNP . First created at Los Alamos National Laboratory (LANL) in 1957, the code simulated neutral...particle types, and previous versions of MCNP were repeatedly validated using both simple and complex 10 geometries [12, 13]. Much greater discussion and

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

International Nuclear Information System (INIS)

Bardet, Philippe; Ricciardi, Guillaume

2016-01-01

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging task in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

Energy Technology Data Exchange (ETDEWEB)

Bardet, Philippe [George Washington Univ., Washington, DC (United States); Ricciardi, Guillaume [Atomic Energy Commission (CEA) (France)

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

The Development and Preliminary Validation of the Behavior, Environment, and Changeability Survey (BECS)

Science.gov (United States)

Walsh, Jennifer R.; Hebert, Angel; Byrd-Bredbenner, Carol; Carey, Gale; Colby, Sarah; Brown-Esters, Onikia N.; Greene, Geoffrey; Hoerr, Sharon; Horacek, Tanya; Kattelmann, Kendra; Kidd, Tandalayo; Koenings, Mallory; Phillips, Beatrice; Shelnutt, Karla P.; White, Adrienne A.

2012-01-01

Objective: To develop and test the validity of the Behavior, Environment, and Changeability Survey (BECS) for identifying the importance and changeability of nutrition, exercise, and stress management behavior and related aspects of the environment. Design: A cross-sectional, online survey of the BECS and selected validated instruments. Setting:…

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

International Nuclear Information System (INIS)

Fermandjian, J.

1984-09-01

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

Indications for spine surgery: validation of an administrative coding algorithm to classify degenerative diagnoses

Science.gov (United States)

Lurie, Jon D.; Tosteson, Anna N.A.; Deyo, Richard A.; Tosteson, Tor; Weinstein, James; Mirza, Sohail K.

2014-01-01

Study Design Retrospective analysis of Medicare claims linked to a multi-center clinical trial. Objective The Spine Patient Outcomes Research Trial (SPORT) provided a unique opportunity to examine the validity of a claims-based algorithm for grouping patients by surgical indication. SPORT enrolled patients for lumbar disc herniation, spinal stenosis, and degenerative spondylolisthesis. We compared the surgical indication derived from Medicare claims to that provided by SPORT surgeons, the “gold standard”. Summary of Background Data Administrative data are frequently used to report procedure rates, surgical safety outcomes, and costs in the management of spinal surgery. However, the accuracy of using diagnosis codes to classify patients by surgical indication has not been examined. Methods Medicare claims were link to beneficiaries enrolled in SPORT. The sensitivity and specificity of three claims-based approaches to group patients based on surgical indications were examined: 1) using the first listed diagnosis; 2) using all diagnoses independently; and 3) using a diagnosis hierarchy based on the support for fusion surgery. Results Medicare claims were obtained from 376 SPORT participants, including 21 with disc herniation, 183 with spinal stenosis, and 172 with degenerative spondylolisthesis. The hierarchical coding algorithm was the most accurate approach for classifying patients by surgical indication, with sensitivities of 76.2%, 88.1%, and 84.3% for disc herniation, spinal stenosis, and degenerative spondylolisthesis cohorts, respectively. The specificity was 98.3% for disc herniation, 83.2% for spinal stenosis, and 90.7% for degenerative spondylolisthesis. Misclassifications were primarily due to codes attributing more complex pathology to the case. Conclusion Standardized approaches for using claims data to accurately group patients by surgical indications has widespread interest. We found that a hierarchical coding approach correctly classified over 90

Modification V to the computer code, STRETCH, for predicting coated-particle behavior

International Nuclear Information System (INIS)

Valentine, K.H.

1975-04-01

Several modifications have been made to the stress analysis code, STRETCH, in an attempt to improve agreement between the calculated and observed behavior of pyrocarbon-coated fuel particles during irradiation in a reactor environment. Specific areas of the code that have been modified are the neutron-induced densification model and the neutron-induced creep calculation. Also, the capability for modeling surface temperature variations has been added. HFIR Target experiments HT-12 through HT-15 have been simulated with the modified code, and the neutron-fluence vs particle-failure predictions compare favorably with the experimental results. Listings of the modified FORTRAN IV main source program and additional FORTRAN IV functions are provided along with instructions for supplying the additional input data. (U.S.)

Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

Energy Technology Data Exchange (ETDEWEB)

Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

2010-05-15

Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter and the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 tim es the hydraulic pipe diameter. (orig.)

Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

Energy Technology Data Exchange (ETDEWEB)

Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

2010-06-15

Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter und the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 times the hydraulic pipe diameter. (orig.)

CSNI Integral test facility validation matrix for the assessment of thermal-hydraulic codes for LWR LOCA and transients

International Nuclear Information System (INIS)

1996-07-01

This report deals with an internationally agreed integral test facility (ITF) matrix for the validation of best estimate thermal-hydraulic computer codes. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a life of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. The construction of such a matrix is an attempt to collect together in a systematic way the best sets of openly available test data for code validation, assessment and improvement, including quantitative assessment of uncertainties in the modelling of phenomena by the codes. In addition to this objective, it is an attempt to record information which has been generated around the world over the last 20 years so that it is more accessible to present and future workers in that field than would otherwise be the case

Validation of DRAGON code in connection with WIMS-AECL/RFSP code system based on ENDF/B-VI library and two group model

International Nuclear Information System (INIS)

Hong, In Seob; Suk, Ho Chun; Kim, Soon Young; Jo, Chang Keun

2002-06-01

The major objective of this research is to validate the incremental cross section property of DRAGON code in connection with WIMS-AECL/DRAGON/RFSP code system with ENDF/B-VI library and full 2G calculation model. The direct comparison between the incremental cross section results calculated by DRAGON with ENDF/B-VI and ENDF/B-V and MULTICELL with ENDF/B-V indicate that there are not much differences between the incremental cross sections of DRAGON with ENDF/B-V and ENDF/B-VI, but there exists large discrepancies between the results of DRAGON and those of MULTICELL. In the analysis of the difference between calculated and measured reactivity worths of various types of control devices during Phase-B Post-Simulation of Wolsong Units 2, 3 and 4, WIMS-AECL/DRAGON/RFSP analysis well agrees with those of previous WIMS-AECL /MULTICELL/RFSP analysis within very small differences. From those results, we can conclude that DRAGON code can be used as a general purpose incremental cross section generation tool for not only the natural uranium fuel but also slightly enriched fuel such as RU or SEU, to cover the shortcomings of natural uranium based MULTICELL code

The Validity of the Interpersonal Behaviors Questionnaire (IBQ) in Sport

Science.gov (United States)

Rocchi, Meredith; Pelletier, Luc; Desmarais, Philippe

2017-01-01

According to Self-Determination Theory (SDT), basic psychological needs will be influenced by other individuals' interpersonal behaviors. The objective of the present research is to extend the validity of the Interpersonal Behaviors Questionnaire (IBQ and IBQ-Self) to the sport context. The measure was designed to assess perceptions of…

Identifying Psoriasis and Psoriatic Arthritis Patients in Retrospective Databases When Diagnosis Codes Are Not Available: A Validation Study Comparing Medication/Prescriber Visit-Based Algorithms with Diagnosis Codes.

Science.gov (United States)

Dobson-Belaire, Wendy; Goodfield, Jason; Borrelli, Richard; Liu, Fei Fei; Khan, Zeba M

2018-01-01

Using diagnosis code-based algorithms is the primary method of identifying patient cohorts for retrospective studies; nevertheless, many databases lack reliable diagnosis code information. To develop precise algorithms based on medication claims/prescriber visits (MCs/PVs) to identify psoriasis (PsO) patients and psoriatic patients with arthritic conditions (PsO-AC), a proxy for psoriatic arthritis, in Canadian databases lacking diagnosis codes. Algorithms were developed using medications with narrow indication profiles in combination with prescriber specialty to define PsO and PsO-AC. For a 3-year study period from July 1, 2009, algorithms were validated using the PharMetrics Plus database, which contains both adjudicated medication claims and diagnosis codes. Positive predictive value (PPV), negative predictive value (NPV), sensitivity, and specificity of the developed algorithms were assessed using diagnosis code as the reference standard. Chosen algorithms were then applied to Canadian drug databases to profile the algorithm-identified PsO and PsO-AC cohorts. In the selected database, 183,328 patients were identified for validation. The highest PPVs for PsO (85%) and PsO-AC (65%) occurred when a predictive algorithm of two or more MCs/PVs was compared with the reference standard of one or more diagnosis codes. NPV and specificity were high (99%-100%), whereas sensitivity was low (≤30%). Reducing the number of MCs/PVs or increasing diagnosis claims decreased the algorithms' PPVs. We have developed an MC/PV-based algorithm to identify PsO patients with a high degree of accuracy, but accuracy for PsO-AC requires further investigation. Such methods allow researchers to conduct retrospective studies in databases in which diagnosis codes are absent. Copyright © 2018 International Society for Pharmacoeconomics and Outcomes Research (ISPOR). Published by Elsevier Inc. All rights reserved.

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)

Studies on the Corrosion Behavior of TiCode-12 with the Variation of Environmental Factors and Heat Treatment Conditions

International Nuclear Information System (INIS)

Yoon, S. R.; Kim, T. Y.; Lee, K. H.

1989-01-01

Corrosion behavior of TiCode-12 (Ti-0.8Ni-0.3Mo) has been studied by means of electrochemical polarization measurements and corrosion morphology examinations in various corrosive environments and different heat treatment conditions of the alloy. 1N H 2 SO 4 at 45 .deg. C was taken as a standard corrosive solution in which Cl - , Fe 3+ and Br - ion were added to investigate their effects. Acid concentration and temperature were also varied. Polarization behaviors of pure Ti, Ni, Mo and Ti 2 Ni were compared with those of heat-treated TiCode-12 specimens to find out how the constituent elements and the intermetallic compound formed during heat treatment of TiCode-12 affect the corrosion of the alloy. Mill-annealed TiCode-12 showed primary and secondary active-passive transition behavior in all the tested H 2 SO 4 solutions. The former behavior was confirmed to be due to Ti and the latter due to Ni and Mo. The passive current densities increased with increased Cl - ion concentration but decreased reversely beyond certain concentration. Fe 3+ ion raised the corrosion potential of TiCode-12 to the passive region, thus reducing the corrosion rate. Br - ion was turned out to be a critical species to induce the pitting of TiCode-12 by some unknown reason. Cathodic polarization behavior of pure Ni and Ti 2 Ni revealed that hydrogen evolution reaction was promoted on these electrodes in acid media. This was ascribed to the cause for sensitization phenomena of TiCode-12 heat-treated in the temperature range in which the eutectoid reaction β→α + Ti 2 Ni occurs. Near pits, observed on a sensitized TiCode-12 specimen immersed in H 2 SO 4 , always found were β crystals in which Ni peak was detected by EDS

Relative validity of the pre-coded food diary used in the Danish National Survey of Diet and Physical Activity

DEFF Research Database (Denmark)

Knudsen, Vibeke Kildegaard; Gille, Maj-Britt; Nielsen, Trine Holmgaard

2011-01-01

Objective: To determine the relative validity of the pre-coded food diary applied in the Danish National Survey of Dietary Habits and Physical Activity. Design: A cross-over study among seventy-two adults (aged 20 to 69 years) recording diet by means of a pre-coded food diary over 4 d and a 4 d...

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

Spent reactor fuel benchmark composition data for code validation

International Nuclear Information System (INIS)

Bierman, S.R.

1991-09-01

To establish criticality safety margins utilizing burnup credit in the storage and transport of spent reactor fuels requires a knowledge of the uncertainty in the calculated fuel composition used in making the reactivity assessment. To provide data for validating such calculated burnup fuel compositions, radiochemical assays are being obtained as part of the United States Department of Energy From-Reactor Cask Development Program. Destructive assay data are being obtained from representative reactor fuels having experienced irradiation exposures up to about 55 GWD/MTM. Assay results and associated operating histories on the initial three samples analyzed in this effort are presented. The three samples were taken from different axial regions of the same fuel rod and represent radiation exposures of about 27, 37, and 44 GWD/MTM. The data are presented in a benchmark type format to facilitate identification/referencing and computer code input

Development of assessment technology for hydrogen burn and fission product behavior in containment

International Nuclear Information System (INIS)

Kim, S. B.; Kim, J. T.; Ha, K. S.; Hong, S. W.; Song, Y. M.; Park, J. H.; Cho, Y. R.; Kang, H. S.

2012-04-01

Analysis tools for hydrogen burn was established to resolve the hydrogen issues in containment. To validate CFX commercial CFD(computational fluid dynamics) code, the hydrogen combustion experiments such as FLAME and ENACEFF for reactor containment were analyzed. And OpenFOAM hydrogen combustion code was developed and validated. Experiments for the flame propagation characteristics in IRWST and the run-up-distance for DDT(Deflagration to detonation transition) were performed and analytical model was evaluated to evaluation of the performance of hydrogen mitigation system, that is, PAR(Passive auto-catalistic re-combiner) To improvement of the fission product modelling in containment, separate analysis module for Iodine behavior and its application tool of K-IODIP (Korea IODIne Package) were developed. PHEBUS FPT-3 analysis was performed to validate MELCOR code. And also the characteristics of fission product behaviors in Future Reactors(GEN-IV) were compared

Social Validity of a Positive Behavior Interventions and Support Model

Science.gov (United States)

Miramontes, Nancy Y.; Marchant, Michelle; Heath, Melissa Allen; Fischer, Lane

2011-01-01

As more schools turn to positive behavior interventions and support (PBIS) to address students' academic and behavioral problems, there is an increased need to adequately evaluate these programs for social relevance. The present study used social validation measures to evaluate a statewide PBIS initiative. Active consumers of the program were…

Regulating professional behavior: codes of ethics or law? Suggested criteria.

Science.gov (United States)

Libman, Liron A

2013-09-01

This paper suggests considering a few parameters when making policy decisions as to the proper "tool" to regulate professional behavior: law or professional ethics. This is done on the background of understanding the place of codes of professional ethics between "pure" ethics and law. Suggested criteria are then illustrated using a few examples. Further discourse may reveal additional factors to support a more rational process of decision-making in this field.

Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

Energy Technology Data Exchange (ETDEWEB)

Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

2012-07-01

The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

First validation of the new continuous energy version of the MORET5 Monte Carlo code

International Nuclear Information System (INIS)

Miss, Joachim; Bernard, Franck; Forestier, Benoit; Haeck, Wim; Richet, Yann; Jacquet, Olivier

2008-01-01

The 5.A.1 version is the next release of the MORET Monte Carlo code dedicated to criticality and reactor calculations. This new version combines all the capabilities that are already available in the multigroup version with many new and enhanced features. The main capabilities of the previous version are the powerful association of a deterministic and Monte Carlo approach (like for instance APOLLO-MORET), the modular geometry, five source sampling techniques and two simulation strategies. The major advance in MORET5 is the ability to perform calculations either a multigroup or a continuous energy simulation. Thanks to these new developments, we now have better control over the whole process of criticality calculations, from reading the basic nuclear data to the Monte Carlo simulation itself. Moreover, this new capability enables us to better validate the deterministic-Monte Carlo multigroup calculations by performing continuous energy calculations with the same code, using the same geometry and tracking algorithms. The aim of this paper is to describe the main options available in this new release, and to present the first results. Comparisons of the MORET5 continuous-energy results with experimental measurements and against another continuous-energy Monte Carlo code are provided in terms of validation and time performance. Finally, an analysis of the interest of using a unified energy grid for continuous energy Monte Carlo calculations is presented. (authors)

VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

Directory of Open Access Journals (Sweden)

Muhammad Darwis Isnaini

2016-03-01

Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

Validation of CBZ code system for post-irradiation examination analysis and sensitivity analysis of (n,γ) branching ratio

International Nuclear Information System (INIS)

Kawamoto, Yosuke; Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

2013-01-01

A code system CBZ is being developed in Hokkaido University. In order to validate it, PIE data, which are nuclide composition data of a spent fuel, have been analyzed with CBZ. The validity is evaluated as ratios of the calculation values to the experimental ones (C/E ratios). Differences between experimental values and calculation ones are smaller than 20% except some nuclides. Thus this code system is validated. Additionally, we evaluate influence of change of (n,γ) branching ratio on inventories of fission products and actinides. As a result, branching ratios of Sb-121, Pm-147, and Am-241 influence inventories of several nuclides. We perform PIE analysis using different (n,γ) branching ratio data from the ORIGEN-2 library, JNDC-Ver.2, and JEFF-3.1A, and find that differences in (n,γ) branching ratios between different nuclear libraries have a non-negligible influence on inventories of several nuclides. (author)

Status of the MELTSPREAD-1 computer code for the analysis of transient spreading of core debris melts

International Nuclear Information System (INIS)

Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.; Chu, C.C.

1992-01-01

A transient, one dimensional, finite difference computer code (MELTSPREAD-1) has been developed to predict spreading behavior of high temperature melts flowing over concrete and/or steel surfaces submerged in water, or without the effects of water if the surface is initially dry. This paper provides a summary overview of models and correlations currently implemented in the code, code validation activities completed thus far, LWR spreading-related safety issues for which the code has been applied, and the status of documentation for the code

Validation of the CATHARE2 code against experimental data from Brayton-cycle plants

International Nuclear Information System (INIS)

Bentivoglio, Fabrice; Tauveron, Nicolas; Geffraye, Genevieve; Gentner, Herve

2008-01-01

In recent years the Commissariat a l'Energie Atomique (CEA) has commissioned a wide range of feasibility studies of future-advanced nuclear reactors, in particular gas-cooled reactors (GCR). The thermohydraulic behaviour of these systems is a key issue for, among other things, the design of the core, the assessment of thermal stresses, and the design of decay heat removal systems. These studies therefore require efficient and reliable simulation tools capable of modelling the whole reactor, including the core, the core vessel, piping, heat exchangers and turbo-machinery. CATHARE2 is a thermal-hydraulic 1D reference safety code developed and extensively validated for the French pressurized water reactors. It has been recently adapted to deal also with gas-cooled reactor applications. In order to validate CATHARE2 for these new applications, CEA has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops. In the short-term perspective, CATHARE2 is being validated against existing experimental data. And in particular from the German power plants Oberhausen I and II. These facilities have both been operated by the German utility Energie Versorgung Oberhausen (E.V.O.) and their power conversion systems resemble to the high-temperature reactor concepts: Oberhausen I is a 13.75-MWe Brayton-cycle air turbine plant, and Oberhausen II is a 50-MWe Brayton-cycle helium turbine plant. The paper presents these two plants, the adopted CATHARE2 modelling and a comparison between experimental data and code results for both steady state and transient cases

Evidence of Validity of the Job Crafting Behaviors Scale

Directory of Open Access Journals (Sweden)

Renata Silva de Carvalho Chinelato

2015-12-01

Full Text Available AbstractJob crafting behavior refers to the changes made by workers in their job context for adjusting their activities to their preferences. We sought to adapt and collect validity evidences of the Job Crafting Behaviors Scale for the Brazilian context, in a sample of 491 workers, with a mean age of 26.7 years. Factor analysis revealed that the final instrument consisted of three dimensions (increasing structural job resources, increasing social job resources, increasing challenging job demands, which showed good internal consistency indexes. These dimensions showed low or moderate correlations with work engagement, positive psychological capital, positive job affect, and in-role performance. The scale showed evidence of validity, the use of which is recommended for future research on the changes that people make in their jobs.

IAEA programme to support development and validation of advanced design and safety analysis codes

Energy Technology Data Exchange (ETDEWEB)

Choi, J., E-mail: J.H.Choi@iaea.org [International Atomic Energy Agency, Vienna (Austria)

2013-07-01

The International Atomic Energy Agency (IAEA) has been organized many international collaboration programs to support the development and validation of design and safety analysis computer codes for nuclear power plants. These programs are normally implemented with a frame of Coordinated Research Project (CRP) or International Collaborative Standard Problem (ICSP). This paper introduces CRPs and ICSPs currently being organized or recently completed by IAEA for this purpose. (author)

Measuring cervical cancer risk: development and validation of the CARE Risky Sexual Behavior Index.

Science.gov (United States)

Reiter, Paul L; Katz, Mira L; Ferketich, Amy K; Ruffin, Mack T; Paskett, Electra D

2009-12-01

To develop and validate a risky sexual behavior index specific to cervical cancer research. Sexual behavior data on 428 women from the Community Awareness Resources and Education (CARE) study were utilized. A weighting scheme for eight risky sexual behaviors was generated and validated in creating the CARE Risky Sexual Behavior Index. Cutpoints were then identified to classify women as having a low, medium, or high level of risky sexual behavior. Index scores ranged from 0 to 35, with women considered to have a low level of risky sexual behavior if their score was less than six (31.3% of sample), a medium level if their score was 6–10 (30.6%), or a high level if their score was 11 or greater (38.1%). A strong association was observed between the created categories and having a previous abnormal Pap smear test (p Sexual Behavior Index provides a tool for measuring risky sexual behavior level for cervical cancer research. Future studies are needed to validate this index in varied populations and test its use in the clinical setting.

Value-Eroding Teacher Behaviors Scale: A Validity and Reliability Study

Science.gov (United States)

Arseven, Zeynep; Kiliç, Abdurrahman; Sahin, Seyma

2016-01-01

In the present study, it is aimed to develop a valid and reliable scale for determining value-eroding behaviors of teachers, hence their values of judgment. The items of the "Value-eroding Teacher Behaviors Scale" were designed in the form of 5-point likert type rating scale. The exploratory factor analysis (EFA) was conducted to…

Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

International Nuclear Information System (INIS)

Lee, A.G.; Wilkin, G.B.

1995-01-01

This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community

CITADEL: a computer code for the analysis of iodine behavior in steam generator tube rupture accidents

International Nuclear Information System (INIS)

1982-04-01

The computer code CITADEL was written to analyze iodine behavior during steam generator tube rupture accidents. The code models the transport and deposition of iodine from its point of escape at the steam generator primary break until its release to the environment. This report provides a brief description of the code including its input requirements and the nature and form of its output. A user's guide describing the manner in which the input data are required to be set up to run the code is also provided

Validation of fuel performance codes at the NRI Rez plc for Temelin and Dukovany NPPs fuel safety evaluations and operation support

International Nuclear Information System (INIS)

Valach, M.; Hejna, J.; Zymak, J.

2003-05-01

The report summarises the first phase of the FUMEX II related work performed in the period September 2002 - May 2003. An inventory of the PIN and FRAS codes family used and developed during previous years was made in light of their applicability (validity) in the domain of high burn-up and FUMEX II Project Experimental database. KOLA data were chosen as appropriate for the first step of both codes fixing (both tuned for VVER fuel originally). The modern requirements, expressed by adaptation of the UO 2 conductivity degradation from OECD HRP, RIM and FGR (athermal) modelling implementation into the PIN code and a diffusion FGR model development planned for embedding, into this code allow us to reasonably shadow or keep tight contact with top quality models as TRANSURANUS, COPERNIC, CYRANO, FEMAXI, FRAPCON3 or ENIGMA. Testing and validation runs with prepared input KOLA deck were made. FUMEX II exercise propose LOCA and RIA like transients, so we started development of those two codes coupling - denominated as PIN2FRAS code. Principles of the interface were tested, benchmarking on tentative RIA pulses on highly burned KOLA fuel are presented as the first achievement from our work. (author)

Experimental validation for combustion analysis of GOTHIC code in 2-dimensional combustion chamber

International Nuclear Information System (INIS)

Lee, J. W.; Yang, S. Y.; Park, K. C.; Jung, S. H.

2002-01-01

In this study, the prediction capability of GOTHIC code for hydrogen combustion phenomena was validated with the results of two-dimensional premixed hydrogen combustion experiment executed by Seoul National University. The experimental chamber has about 24 liter free volume (1x0.024x1 m 3 ) and 2-dimensional rectangular shape. The test were preformed with 10% hydrogen/air gas mixture and conducted with combination of two igniter positions (top center, top corner) and two boundary conditions (bottom full open, bottom right half open). Using the lumped parameter and mechanistic combustion model in GOTHIC code, the SNU experiments were simulated under the same conditions. The GOTHIC code prediction of the hydrogen combustion phenomena did not compare well with the experimental results. In case of lumped parameter simulation, the combustion time was predicted appropriately. But any other local information related combustion phenomena could not be obtained. In case of mechanistic combustion analysis, the physical combustion phenomena of gas mixture were not matched experimental ones. In boundary open cases, the GOTHIC predicted very long combustion time and the flame front propagation could not simulate appropriately. Though GOTHIC showed flame propagation phenomenon in adiabatic calculation, the induction time of combustion was still very long compare with experimental results. Also, it was found that the combustion model of GOTHIC code had some weak points in low concentration of hydrogen combustion simulation

Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104)

International Nuclear Information System (INIS)

Kress, T.S.

1985-04-01

The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time

Validation of a Behavioral Approach for Measuring Saccades in Parkinson's Disease.

Science.gov (United States)

Turner, Travis H; Renfroe, Jenna B; Duppstadt-Delambo, Amy; Hinson, Vanessa K

2017-01-01

Speed and control of saccades are related to disease progression and cognitive functioning in Parkinson's disease (PD). Traditional eye-tracking complexities encumber application for individual evaluations and clinical trials. The authors examined psychometric properties of standalone tasks for reflexive prosaccade latency, volitional saccade initiation, and saccade inhibition (antisaccade) in a heterogeneous sample of 65 PD patients. Demographics had minimal impact on task performance. Thirty-day test-retest reliability estimates for behavioral tasks were acceptable and similar to traditional eye tracking. Behavioral tasks demonstrated concurrent validity with traditional eye-tracking measures; discriminant validity was less clear. Saccade initiation and inhibition discriminated PD patients with cognitive impairment. The present findings support further development and use of the behavioral tasks for assessing latency and control of saccades in PD.

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

Validation of the TIARA code to tritium inventory data

International Nuclear Information System (INIS)

Billone, M.C.

1994-03-01

The TIARA code has been developed to predict tritium inventory in Li 2 O breeder ceramic and to predict purge exit flow rate and composition. Inventory predictions are based on models for bulk diffusion, surface desorption, solubility and precipitation. Parameters for these models are determined from the results of laboratory annealing studies on unirradiated and irradiated Li 2 O. Inventory data from in-reactor purge flow tests are used for model improvement, fine-tuning of model parameters and validation. In this current work, the inventory measurement near the purge inlet from the BEATRIX-II thin-ring sample is used to fine tune the surface desorption model parameters for T > 470 degrees C, and the inventory measurement near the midplane from VOM-15H is used to fine tune the moisture solubility model parameters. predictions are then validated to the remaining inventory data from EXOTIC-2 (1 point), SIBELIUS (3 axial points), VOM-15H (2 axial points), CRITIC-1 (4 axial points), BEATRIX-II thin ring (3 axial points) and BEATRIX-II thick pellet (5 radial points). Thus. of the 20 data points, two we re used for fine tuning model parameters and 18 were used for validation. The inventory data span the range of 0.05--1.44 wppm with an average of 0.48 wppm. The data pertain to samples whose end-of-life temperatures were in the range of 490--1000 degrees C. On the average, the TIARA predictions agree quite well with the data (< 0.02 wppm difference). However, the root-mean-square deviation is 0.44 wppm, mostly due to over-predictions for the SIBELIUS samples and the higher-temperature radial samples from the BEATRIX-11 thick-pellet

Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

Directory of Open Access Journals (Sweden)

A. Del Nevo

2012-01-01

Full Text Available The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents, a primary-to-secondary leak, and a parametric study (natural circulation test aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

Modeling of the WWER-1000 fuel-rod behavior in steady-state condition with FRAPCONE-3 computer code

International Nuclear Information System (INIS)

Andreeva, Marina; Totev, Totju; Stoyanov, Stoyan

2008-01-01

It is presented within the paper the results of the modeling and the assessment of the integral code predictions of the WWER fuel-rod behavior in steady-state condition. The assessments in this paper have used the MASSIH and ANS 5.4 subroutine in the code. The modeling and calculations have been performed with FRAPCONE-3 computer code in Argonne National Laboratory, USA

Thermal-Hydraulic Analysis of SWAMUP Facility Using ATHLET-SC Code

Energy Technology Data Exchange (ETDEWEB)

Wang, Zidi; Cao, Zhen; Liu, Xiaojing, E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai (China)

2015-03-16

During the loss of coolant accident (LOCA) of supercritical water-cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP) with 2 × 2 rod bundle in Shanghai Jiao Tong University (SJTU) will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior, e.g., system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g., heat transfer coefficient, critical heat flux correlation are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal-hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

Analysis of the technology acceptance model in examining hospital nurses' behavioral intentions toward the use of bar code medication administration.

Science.gov (United States)

Song, Lunar; Park, Byeonghwa; Oh, Kyeung Mi

2015-04-01

Serious medication errors continue to exist in hospitals, even though there is technology that could potentially eliminate them such as bar code medication administration. Little is known about the degree to which the culture of patient safety is associated with behavioral intention to use bar code medication administration. Based on the Technology Acceptance Model, this study evaluated the relationships among patient safety culture and perceived usefulness and perceived ease of use, and behavioral intention to use bar code medication administration technology among nurses in hospitals. Cross-sectional surveys with a convenience sample of 163 nurses using bar code medication administration were conducted. Feedback and communication about errors had a positive impact in predicting perceived usefulness (β=.26, Pmodel predicting for behavioral intention, age had a negative impact (β=-.17, Pmodel explained 24% (Ptechnology.

Enhancing the ABAQUS Thermomechanics Code to Simulate Multidimensional Steady and Transient Fuel Rod Behavior

Energy Technology Data Exchange (ETDEWEB)

Williamson, R.L.; Knoll, D.A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-3855 (United States)

2009-06-15

isotropic and anisotropic elasticity, thermal expansion, plasticity, and thermal creep. Due to its commercial nature and widespread use, ABAQUS includes powerful pre- and post-processing software, facilitating efficient problem setup and rapid interpretation of results. The code undergoes extensive and well-documented verification and validation procedures, reducing the risk of coding errors leading to invalid results. Essential for the intended application, ABAQUS includes a user interface permitting the inclusion of user-developed models to simulate fuel-specific phenomena. This powerful interface makes it possible to define any constitutive model of arbitrary complexity. Focusing initially on UO{sub 2} fuel, user subroutines have been developed and tested to describe temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, and fission gas release. Assuming Zircaloy as the clad material, models have been included for both thermal and irradiation creep. A subroutine has been developed to model the complex heat exchange across the fuel-clad gap. This model accounts for the temperature, pressure, and species-dependent gap gas thermal conductivity, the changing gap width, the roughness of both the fuel and clad, jump distances, contact pressure, and the cladding hardness. Gap heat transfer is tightly coupled to the fission gas model and the basic mechanics. Field variables are defined at each integration point to track fuel specific phenomena such as burnup, fission gas produced, and fission gas released. Fuels modeling capability is demonstrated using a 2D axisymmetric analysis of a rod section including two individual pellets, the associated cladding, the gas-filled gap, and an upper plenum region. The plenum and gap gas are modeled as a 'hydro' cavity in ABAQUS with the cavity volume dictated by the evolving solid mechanics, the gas mass and species mixture, and the gas state computed assuming ideal gas behavior. A simple

Behavioral analysis of malicious code through network traffic and system call monitoring

Science.gov (United States)

Grégio, André R. A.; Fernandes Filho, Dario S.; Afonso, Vitor M.; Santos, Rafael D. C.; Jino, Mario; de Geus, Paulo L.

2011-06-01

Malicious code (malware) that spreads through the Internet-such as viruses, worms and trojans-is a major threat to information security nowadays and a profitable business for criminals. There are several approaches to analyze malware by monitoring its actions while it is running in a controlled environment, which helps to identify malicious behaviors. In this article we propose a tool to analyze malware behavior in a non-intrusive and effective way, extending the analysis possibilities to cover malware samples that bypass current approaches and also fixes some issues with these approaches.

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

International Nuclear Information System (INIS)

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

1997-07-01

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

Chiari malformation Type I surgery in pediatric patients. Part 1: validation of an ICD-9-CM code search algorithm.

Science.gov (United States)

Ladner, Travis R; Greenberg, Jacob K; Guerrero, Nicole; Olsen, Margaret A; Shannon, Chevis N; Yarbrough, Chester K; Piccirillo, Jay F; Anderson, Richard C E; Feldstein, Neil A; Wellons, John C; Smyth, Matthew D; Park, Tae Sung; Limbrick, David D

2016-05-01

OBJECTIVE Administrative billing data may facilitate large-scale assessments of treatment outcomes for pediatric Chiari malformation Type I (CM-I). Validated International Classification of Diseases, Ninth Revision, Clinical Modification (ICD-9-CM) code algorithms for identifying CM-I surgery are critical prerequisites for such studies but are currently only available for adults. The objective of this study was to validate two ICD-9-CM code algorithms using hospital billing data to identify pediatric patients undergoing CM-I decompression surgery. METHODS The authors retrospectively analyzed the validity of two ICD-9-CM code algorithms for identifying pediatric CM-I decompression surgery performed at 3 academic medical centers between 2001 and 2013. Algorithm 1 included any discharge diagnosis code of 348.4 (CM-I), as well as a procedure code of 01.24 (cranial decompression) or 03.09 (spinal decompression or laminectomy). Algorithm 2 restricted this group to the subset of patients with a primary discharge diagnosis of 348.4. The positive predictive value (PPV) and sensitivity of each algorithm were calculated. RESULTS Among 625 first-time admissions identified by Algorithm 1, the overall PPV for CM-I decompression was 92%. Among the 581 admissions identified by Algorithm 2, the PPV was 97%. The PPV for Algorithm 1 was lower in one center (84%) compared with the other centers (93%-94%), whereas the PPV of Algorithm 2 remained high (96%-98%) across all subgroups. The sensitivity of Algorithms 1 (91%) and 2 (89%) was very good and remained so across subgroups (82%-97%). CONCLUSIONS An ICD-9-CM algorithm requiring a primary diagnosis of CM-I has excellent PPV and very good sensitivity for identifying CM-I decompression surgery in pediatric patients. These results establish a basis for utilizing administrative billing data to assess pediatric CM-I treatment outcomes.

A rational method to evaluate tornado-borne missile speed in nuclear power plants. Validation of a numerical code based on Fujita's tornado model

International Nuclear Information System (INIS)

Eguchi, Yuzuru; Sugimoto, Soichiro; Hattori, Yasuo; Hirakuchi, Hiromaru

2015-01-01

Explanation is given about a rational method to evaluate tornado-borne missile speed, flight distance and flight height to be used for safety design of a nuclear power plant. In the method, the authors employed Fujita's DBT-77 model as a tornado wind model to take the near-ground tornado wind profile into account. A liftoff model of an object on the ground was developed by conservatively modeling the lift force due to ground effect. The wind field model and the liftoff model have been compiled together with a conventional flight model into a computer code, named TONBOS. In this study, especially, the code is verified for one- and two-dimensional free-fall problems as well as a case of 1957 Dallas tornado wind field model, whose solutions are theoretically or numerically known. Finally, the code is validated by typical car behaviors characterized by tornado wind speeds of the enhanced Fujita scale, as well as by an actual event where a truck was blown away by a tornado which struck a part of the town of Saroma, Hokkaido in November, 2006. (author)

Concurrent validation of an inertial measurement system to quantify kicking biomechanics in four football codes.

Science.gov (United States)

Blair, Stephanie; Duthie, Grant; Robertson, Sam; Hopkins, William; Ball, Kevin

2018-05-17

Wearable inertial measurement systems (IMS) allow for three-dimensional analysis of human movements in a sport-specific setting. This study examined the concurrent validity of a IMS (Xsens MVN system) for measuring lower extremity and pelvis kinematics in comparison to a Vicon motion analysis system (MAS) during kicking. Thirty footballers from Australian football (n = 10), soccer (n = 10), rugby league and rugby union (n = 10) clubs completed 20 kicks across four conditions. Concurrent validity was assessed using a linear mixed-modelling approach, which allowed the partition of between and within-subject variance from the device measurement error. Results were expressed in raw and standardised units for assessments of differences in means and measurement error, and interpreted via non-clinical magnitude-based inferences. Trivial to small differences were found in linear velocities (foot and pelvis), angular velocities (knee, shank and thigh), sagittal joint (knee and hip) and segment angle (shank and pelvis) means (mean difference: 0.2-5.8%) between the IMS and MAS in Australian football, soccer and the rugby codes. Trivial to small measurement errors (from 0.1 to 5.8%) were found between the IMS and MAS in all kinematic parameters. The IMS demonstrated acceptable levels of concurrent validity compared to a MAS when measuring kicking biomechanics across the four football codes. Wearable IMS offers various benefits over MAS, such as, out-of-laboratory testing, larger measurement range and quick data output, to help improve the ecological validity of biomechanical testing and the timing of feedback. The results advocate the use of IMS to quantify biomechanics of high-velocity movements in sport-specific settings. Copyright © 2018 Elsevier Ltd. All rights reserved.

Optimization and Validation of the Developed Uranium Isotopic Analysis Code

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Enhancing the ABAQUS Thermomechanics Code to Simulate Steady and Transient Fuel Rod Behavior

International Nuclear Information System (INIS)

Williamson, R.L.; Knoll, D.A.

2009-01-01

A powerful multidimensional fuels performance capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. The various modeling capabilities are demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multi-pellet fuel rod, during both steady and transient operation. Computational results demonstrate the importance of a multidimensional fully-coupled thermomechanics treatment. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermo-mechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

International Nuclear Information System (INIS)

Lee, A.G.; Wilkin, G.B.

1996-03-01

During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods various organizations use to verify and validate their codes and libraries. Five organizations, Atomic Energy of Canada Limited (AECL, Canada), China Institute of Atomic Energy (CIAE, People's Republic of China), Japan Atomic Energy Research Institute (JAERI, Japan), Oak Ridge National Laboratories (ORNL, USA), and Siemens (Germany) responded to the survey. The results of the survey are compiled in this report. (author) 36 refs., 3 tabs

Summary of aerosol code-comparison results for LWR aerosol containment tests LA1, LA2, and LA3

International Nuclear Information System (INIS)

Wright, A.L.; Wilson, J.H.; Arwood, P.C.

1987-01-01

The light-water reactor (LWR) aerosol containment experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities for the LACE tests are being coordinated at the Oak Ridge National Laboratory. For each of the six experiments, pretest calculations (for code-to-code comparisons) and blind post-test calculations (for code-to-test data comparisons) are being performed. This paper presents a summary of the pretest aerosol-code results for tests LA1, LA2, and LA3

First experimental validation on the core equilibrium code: HARMONIE

International Nuclear Information System (INIS)

Van Dorsselaere, J.; Cozzani, M.; Gnuffi, M.

1981-08-01

The code HARMONIE calculates the mechanical equilibrium of a fast reactor. An experimental program of deformation, in air, of groups of subassemblies, was performed on a mock-up, in the Super Phenix 1- geometry. This program included three kinds of tests, all performed without and then with grease: on groups of 2 or 3 rings of subassemblies, subjected to a force acting upon flats or angles; on groups of 35 and 41 subassemblies, subjected to a force acting on the first row, then with 1 or 2 empty cells; and on groups with 1 or 2 bowed subassemblies or 1 enlarged one over flats. A preliminary test on the friction coefficient in air between two pads showed some dependance upon the pad surface condition with a scattering factor of 8. Two basic code hypotheses were validated: the rotation of the subassemblies around their axis was negligible after deformation of the group, and the choice of a mean Maxwell coefficient, between those of 1st and 2nd slope, led to very similar results to experimental. The agreement between tests and HARMONIE calculations was suitable, qualitatively for all the groups and quantitatively for regular groups of 3 rings at most. But the difference increased for larger groups of 35 or 41 subassemblies: friction between pads, neglected by HARMONIE, seems to be the main reason. Other reasons for these differences are: the influence of the loading order on the mock-up, and the initial contacts issued from the gap between foot and diagrid-insert, and from manufacture bowings

TOUGH+HYDRATE v1.2 User's Manual: A Code for the Simulation of System Behavior in Hydrate-Bearing Geologic Media

Energy Technology Data Exchange (ETDEWEB)

Moridis, George J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Kowalsky, Michael B. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Pruess, Karsten [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2012-08-01

TOUGH+HYDRATE v1.2 is a code for the simulation of the behavior of hydratebearing geologic systems, and represents the second update of the code since its first release [Moridis et al., 2008]. By solving the coupled equations of mass and heat balance, TOUGH+HYDRATE can model the non-isothermal gas release, phase behavior and flow of fluids and heat under conditions typical of common natural CH4-hydrate deposits (i.e., in the permafrost and in deep ocean sediments) in complex geological media at any scale (from laboratory to reservoir) at which Darcy’s law is valid. TOUGH+HYDRATE v1.2 includes both an equilibrium and a kinetic model of hydrate formation and dissociation. The model accounts for heat and up to four mass components, i.e., water, CH4, hydrate, and water-soluble inhibitors such as salts or alcohols. These are partitioned among four possible phases (gas phase, liquid phase, ice phase and hydrate phase). Hydrate dissociation or formation, phase changes and the corresponding thermal effects are fully described, as are the effects of inhibitors. The model can describe all possible hydrate dissociation mechanisms, i.e., depressurization, thermal stimulation, salting-out effects and inhibitor-induced effects. TOUGH+HYDRATE is a member of TOUGH+, the successor to the TOUGH2 [Pruess et al., 1991] family of codes for multi-component, multiphase fluid and heat flow developed at the Lawrence Berkeley National Laboratory. It is written in standard FORTRAN 95/2003, and can be run on any computational platform (workstation, PC, Macintosh) for which such compilers are available.

Validity of administrative database code algorithms to identify vascular access placement, surgical revisions, and secondary patency.

Science.gov (United States)

Al-Jaishi, Ahmed A; Moist, Louise M; Oliver, Matthew J; Nash, Danielle M; Fleet, Jamie L; Garg, Amit X; Lok, Charmaine E

2018-03-01

We assessed the validity of physician billing codes and hospital admission using International Classification of Diseases 10th revision codes to identify vascular access placement, secondary patency, and surgical revisions in administrative data. We included adults (≥18 years) with a vascular access placed between 1 April 2004 and 31 March 2013 at the University Health Network, Toronto. Our reference standard was a prospective vascular access database (VASPRO) that contains information on vascular access type and dates of placement, dates for failure, and any revisions. We used VASPRO to assess the validity of different administrative coding algorithms by calculating the sensitivity, specificity, and positive predictive values of vascular access events. The sensitivity (95% confidence interval) of the best performing algorithm to identify arteriovenous access placement was 86% (83%, 89%) and specificity was 92% (89%, 93%). The corresponding numbers to identify catheter insertion were 84% (82%, 86%) and 84% (80%, 87%), respectively. The sensitivity of the best performing coding algorithm to identify arteriovenous access surgical revisions was 81% (67%, 90%) and specificity was 89% (87%, 90%). The algorithm capturing arteriovenous access placement and catheter insertion had a positive predictive value greater than 90% and arteriovenous access surgical revisions had a positive predictive value of 20%. The duration of arteriovenous access secondary patency was on average 578 (553, 603) days in VASPRO and 555 (530, 580) days in administrative databases. Administrative data algorithms have fair to good operating characteristics to identify vascular access placement and arteriovenous access secondary patency. Low positive predictive values for surgical revisions algorithm suggest that administrative data should only be used to rule out the occurrence of an event.

Systematic review of validated case definitions for diabetes in ICD-9-coded and ICD-10-coded data in adult populations.

Science.gov (United States)

Khokhar, Bushra; Jette, Nathalie; Metcalfe, Amy; Cunningham, Ceara Tess; Quan, Hude; Kaplan, Gilaad G; Butalia, Sonia; Rabi, Doreen

2016-08-05

With steady increases in 'big data' and data analytics over the past two decades, administrative health databases have become more accessible and are now used regularly for diabetes surveillance. The objective of this study is to systematically review validated International Classification of Diseases (ICD)-based case definitions for diabetes in the adult population. Electronic databases, MEDLINE and Embase, were searched for validation studies where an administrative case definition (using ICD codes) for diabetes in adults was validated against a reference and statistical measures of the performance reported. The search yielded 2895 abstracts, and of the 193 potentially relevant studies, 16 met criteria. Diabetes definition for adults varied by data source, including physician claims (sensitivity ranged from 26.9% to 97%, specificity ranged from 94.3% to 99.4%, positive predictive value (PPV) ranged from 71.4% to 96.2%, negative predictive value (NPV) ranged from 95% to 99.6% and κ ranged from 0.8 to 0.9), hospital discharge data (sensitivity ranged from 59.1% to 92.6%, specificity ranged from 95.5% to 99%, PPV ranged from 62.5% to 96%, NPV ranged from 90.8% to 99% and κ ranged from 0.6 to 0.9) and a combination of both (sensitivity ranged from 57% to 95.6%, specificity ranged from 88% to 98.5%, PPV ranged from 54% to 80%, NPV ranged from 98% to 99.6% and κ ranged from 0.7 to 0.8). Overall, administrative health databases are useful for undertaking diabetes surveillance, but an awareness of the variation in performance being affected by case definition is essential. The performance characteristics of these case definitions depend on the variations in the definition of primary diagnosis in ICD-coded discharge data and/or the methodology adopted by the healthcare facility to extract information from patient records. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Development of LEAP-JET code for sodium-water reaction analysis. Validation by sodium-water reaction tests (SWAT-1R)

International Nuclear Information System (INIS)

Seino, Hiroshi; Hamada, Hirotsugu

2004-03-01

The sodium-water reaction event in an FBR steam generator (SG) has influence on the safety, economical efficiency, etc. of the plant, so that the selection of design base leak (DBL) of the SG is considered as one of the important matters. The clarification of the sodium-water reaction phenomenon and the development of an analysis model are necessary to estimate the sodium-water reaction event with high accuracy and rationality in selecting the DBL. The reaction jet model is pointed out as a part of the necessary improvements to evaluate the overheating tube rupture of large SGs, since the behavior of overheating tube rupture is largely affected by the reaction jet conditions outside the tube. Therefore, LEAP-JET has been developed as an analysis code for the simulation of sodium-water reactions. This document shows the validation of the LEAP-JET code by the Sodium-Water Reaction Test (SWAT-1R). The following results have been obtained: (1) The reaction rate constant, K, is estimated at between 0.001≤K≤0.1 from the LEAP-JET analysis of the SWAT-1R data. (2) The analytical results on the high-temperature region and the behaviors of reaction consumption (Na, H 2 O) and products (H 2 , NaOH, Na 2 O) are considered to be physically reasonable. (3) The LEAP-JET analysis shows the tendency of overestimation in the maximum temperature and temperature distribution of the reaction jet. (4) In the LEAP-JET analysis, the numerical calculation becomes unstably, especially in the mesh containing quite small sodium mass. Therefore, it is necessary to modify the computational algorism to stabilize it and obtain the optimum value of K in sodium-water reactions. (author)

Validation of a Computational Fluid Dynamics (CFD) Code for Supersonic Axisymmetric Base Flow

Science.gov (United States)

Tucker, P. Kevin

1993-01-01

The ability to accurately and efficiently calculate the flow structure in the base region of bodies of revolution in supersonic flight is a significant step in CFD code validation for applications ranging from base heating for rockets to drag for protectives. The FDNS code is used to compute such a flow and the results are compared to benchmark quality experimental data. Flowfield calculations are presented for a cylindrical afterbody at M = 2.46 and angle of attack a = O. Grid independent solutions are compared to mean velocity profiles in the separated wake area and downstream of the reattachment point. Additionally, quantities such as turbulent kinetic energy and shear layer growth rates are compared to the data. Finally, the computed base pressures are compared to the measured values. An effort is made to elucidate the role of turbulence models in the flowfield predictions. The level of turbulent eddy viscosity, and its origin, are used to contrast the various turbulence models and compare the results to the experimental data.

The reactor kinetics code tank: a validation against selected SPERT-1b experiments

International Nuclear Information System (INIS)

Ellis, R.J.

1990-01-01

The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

STAR-CCM+ Verification and Validation Plan

Energy Technology Data Exchange (ETDEWEB)

Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-30

The commercial Computational Fluid Dynamics (CFD) code STAR-CCM+ provides general purpose finite volume method solutions for fluid dynamics and energy transport. This document defines plans for verification and validation (V&V) of the base code and models implemented within the code by the Consortium for Advanced Simulation of Light water reactors (CASL). The software quality assurance activities described herein are port of the overall software life cycle defined in the CASL Software Quality Assurance (SQA) Plan [Sieger, 2015]. STAR-CCM+ serves as the principal foundation for development of an advanced predictive multi-phase boiling simulation capability within CASL. The CASL Thermal Hydraulics Methods (THM) team develops advanced closure models required to describe the subgrid-resolution behavior of secondary fluids or fluid phases in multiphase boiling flows within the Eulerian-Eulerian framework of the code. These include wall heat partitioning models that describe the formation of vapor on the surface and the forces the define bubble/droplet dynamic motion. The CASL models are implemented as user coding or field functions within the general framework of the code. This report defines procedures and requirements for V&V of the multi-phase CFD capability developed by CASL THM. Results of V&V evaluations will be documented in a separate STAR-CCM+ V&V assessment report. This report is expected to be a living document and will be updated as additional validation cases are identified and adopted as part of the CASL THM V&V suite.

Flight code validation simulator

Science.gov (United States)

Sims, Brent A.

1996-05-01

An End-To-End Simulation capability for software development and validation of missile flight software on the actual embedded computer has been developed utilizing a 486 PC, i860 DSP coprocessor, embedded flight computer and custom dual port memory interface hardware. This system allows real-time interrupt driven embedded flight software development and checkout. The flight software runs in a Sandia Digital Airborne Computer and reads and writes actual hardware sensor locations in which Inertial Measurement Unit data resides. The simulator provides six degree of freedom real-time dynamic simulation, accurate real-time discrete sensor data and acts on commands and discretes from the flight computer. This system was utilized in the development and validation of the successful premier flight of the Digital Miniature Attitude Reference System in January of 1995 at the White Sands Missile Range on a two stage attitude controlled sounding rocket.

Organizational citizenship behavior in schools: validation of a questionnaire.

Science.gov (United States)

Neves, Paula C; Paixão, Rui; Alarcão, Madalena; Gomes, A Duarte

2014-01-01

The present study examines the psychometric properties (including factorial validity) of an organizational citizenship behavior (OCB) scale in a school context. A total of 321 middle and high school teachers from 59 schools in urban and rural areas of central Portugal completed the OCB scale at their schools. The confirmatory factor analysis validated a hierarchical model with four latent factors on the first level (altruism, conscientiousness, civic participation and courtesy) and a second order factor (OCB). The revised model fit with the data, χ 2 /gl = 1.97; CFI = .962; GFI = .952, RMSEA = .05. The proposed scale (comportamentos de cidadania organizacional em escolas- Revista CCOE-R)- is a valid instrument to assess teacher's perceptions of OCB in their schools, allowing investigation at the organizational level of analysis.

Approaches to Validation of Models for Low Gravity Fluid Behavior

Science.gov (United States)

Chato, David J.; Marchetta, Jeffery; Hochstein, John I.; Kassemi, Mohammad

2005-01-01

This paper details the author experiences with the validation of computer models to predict low gravity fluid behavior. It reviews the literature of low gravity fluid behavior as a starting point for developing a baseline set of test cases. It examines authors attempts to validate their models against these cases and the issues they encountered. The main issues seem to be that: Most of the data is described by empirical correlation rather than fundamental relation; Detailed measurements of the flow field have not been made; Free surface shapes are observed but through thick plastic cylinders, and therefore subject to a great deal of optical distortion; and Heat transfer process time constants are on the order of minutes to days but the zero-gravity time available has been only seconds.

FACTAR validation

International Nuclear Information System (INIS)

Middleton, P.B.; Wadsworth, S.L.; Rock, R.C.; Sills, H.E.; Langman, V.J.

1995-01-01

A detailed strategy to validate fuel channel thermal mechanical behaviour codes for use of current power reactor safety analysis is presented. The strategy is derived from a validation process that has been recently adopted industry wide. Focus of the discussion is on the validation plan for a code, FACTAR, for application in assessing fuel channel integrity safety concerns during a large break loss of coolant accident (LOCA). (author)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

International Nuclear Information System (INIS)

Hussein, M.S; Lewis, B.J.; Bonin, H.W.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

Development of an automatic validation system for simulation codes of the fusion research; Entwicklung eines automatischen Validierungssystems fuer Simulationscodes der Fusionsforschung

Energy Technology Data Exchange (ETDEWEB)

Galonska, Andreas

2010-03-15

In the present master thesis the development oa an automatic validation system for the simulation code ERO is documented. This 3D Monte-carlo code models the transport of impurities as well as plasma-wall interaction processes and has great importance for the fusion research. The validation system is based on JuBE (Julich Benchmarking Environment), the flexibility of which allows a slight extension of the system to other codes, for instance such, which are operated in the framework of the EU Task Force ITM (Integrated Tokamak Modelling). The chosen solution - JuBE and a special program for the ''intellectual'' comparison of actual and reference-edition data of ERO is described and founded. The use of this program and the configuration of JuBE are detailedly described. Simulations to different plasma experiments, which serve as reference cases for the automatic validation, are explained. The working of the system is illustrated by the description of a test case. This treats the failure localization and improvement in the parallelization of an important ERO module (tracking of physically eroded particle). It is demonstrated, how the system reacts in an erroneous validation and the subsequently performed error correction leads to a positive result. Finally a speed-up curve of the parallelization is established by means of the output data of JuBE.

[Reliability and Validity of the Behavioral Check List for Preschool Children to Measure Attention Deficit Hyperactivity Behaviors].

Science.gov (United States)

Tsuno, Kanami; Yoshimasu, Kouichi; Hayashi, Takashi; Tatsuta, Nozomi; Ito, Yuki; Kamijima, Michihiro; Nakai, Kunihiko

2018-01-01

Nowadays, attention deficit hyperactivity (ADH) problems are observed commonly among school-age children. However, questionnaires specific to ADH behaviors among preschool children are very few. The aim of this study was to investigate the reliability and validity of the 25-item Behavioral Check List (BCL), which was developed from interviews of parents with children who were diagnosed as having Attention-deficit/hyperactivity disorder (ADHD) and measures ADH behaviors in preschool age. We recruited 22 teachers from 10 nurseries/kindergartens in Miyagi Prefecture, Japan. A total of 138 preschool children were assessed using the BCL. To investigate inter-rater reliability, two teachers from each facility assess seven to twenty children in their class, and intraclass correlation coefficients (ICCs) were calculated. The teachers additionally answered questions in the 1/5-5 Caregiver-Teacher Report Form (C-TRF) to investigate the criterion validity of the BCL. To investigate structural validity, exploratory factor analysis with promax rotation and confirmatory factor analysis were performed. The internal consistency reliability of the BCL was good (α = 0.92) and correlation analyses also confirmed its excellent criterion validity. Although exploratory factor analysis for the BCL yielded a five-factor model that consisted of a factor structure different from that of the original one, the results were similar to the original six factors. The ICCs of the BCL were 0.38-0.99 and it was not high enough for inter-rater reliability in some facilities. However, there is a possibility to improve it by giving raters adequate explanations when using BCL. The present study showed acceptable levels of reliability and validity of the BCL among Japanese preschool children.

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

Energy Technology Data Exchange (ETDEWEB)

Glenn McCreery; Hugh McIlroy

2007-09-01

Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related

Development and validation of a new self-report measure of pain behaviors.

Science.gov (United States)

Cook, Karon F; Keefe, Francis; Jensen, Mark P; Roddey, Toni S; Callahan, Leigh F; Revicki, Dennis; Bamer, Alyssa M; Kim, Jiseon; Chung, Hyewon; Salem, Rana; Amtmann, Dagmar

2013-12-01

Pain behaviors that are maintained beyond the acute stage after injury can contribute to subsequent psychosocial and physical disability. Critical to the study of pain behaviors is the availability of psychometrically sound pain behavior measures. In this study we developed a self-report measure of pain behaviors, the Pain Behaviors Self Report (PaB-SR). PaB-SR scores were developed using item response theory and evaluated using a rigorous, multiple-witness approach to validity testing. Participants included 661 survey participants with chronic pain and with multiple sclerosis, back pain, or arthritis; 618 survey participants who were significant others of a chronic pain participant; and 86 participants in a videotaped pain behavior observation protocol. Scores on the PaB-SR were found to be measurement invariant with respect to clinical condition. PaB-SR scores, observer reports, and the videotaped protocol yielded distinct, but convergent views of pain behavior, supporting the validity of the new measure. The PaB-SR is expected to be of substantial utility to researchers wishing to explore the relationship between pain behaviors and constructs such as pain intensity, pain interference, and disability. Copyright © 2013 International Association for the Study of Pain. Published by Elsevier B.V. All rights reserved.

Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

International Nuclear Information System (INIS)

Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

2015-01-01

Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

Reliability and Validity of the Dyadic Observed Communication Scale (DOCS).

Science.gov (United States)

Hadley, Wendy; Stewart, Angela; Hunter, Heather L; Affleck, Katelyn; Donenberg, Geri; Diclemente, Ralph; Brown, Larry K

2013-02-01

We evaluated the reliability and validity of the Dyadic Observed Communication Scale (DOCS) coding scheme, which was developed to capture a range of communication components between parents and adolescents. Adolescents and their caregivers were recruited from mental health facilities for participation in a large, multi-site family-based HIV prevention intervention study. Seventy-one dyads were randomly selected from the larger study sample and coded using the DOCS at baseline. Preliminary validity and reliability of the DOCS was examined using various methods, such as comparing results to self-report measures and examining interrater reliability. Results suggest that the DOCS is a reliable and valid measure of observed communication among parent-adolescent dyads that captures both verbal and nonverbal communication behaviors that are typical intervention targets. The DOCS is a viable coding scheme for use by researchers and clinicians examining parent-adolescent communication. Coders can be trained to reliably capture individual and dyadic components of communication for parents and adolescents and this complex information can be obtained relatively quickly.

Cross-cultural Exploration of Consumers’ Beliefs and Behavioral Intentions towards QR Codes in Marketing: An Experimental Study in India and USA

Directory of Open Access Journals (Sweden)

Hemant Bamoriya

2014-08-01

Full Text Available Current study focuses on examining consumers‟ beliefs and behavioral intentions towards QR Codes in marketing across different cultures such as India & USA. Study also examines select variables that moderate the relationship between beliefs and behavioral intentions. For this a scenario based experimental design was used. Findings suggest positive relationship between beliefs & behavioral intention and between culture & beliefs. Based on the empirical findings, study make important implications for the marketers so as to bring effectiveness in QR Code based marketing campaigns. Use of QR Codes in marketing is prevailing rapidly, however its effective integration in marketing mix remains mysterious as very little is known about consumers‟ beliefs and behavioral intentions towards it. Reason is that such academic research is practically non-existent, thus current study is of particular value.

International integral experiments databases in support of nuclear data and code validation

International Nuclear Information System (INIS)

Briggs, J. Blair; Gado, Janos; Hunter, Hamilton; Kodeli, Ivan; Salvatores, Massimo; Sartori, Enrico

2002-01-01

The OECD/NEA Nuclear Science Committee (NSC) has identified the need to establish international databases containing all the important experiments that are available for sharing among the specialists. The NSC has set up or sponsored specific activities to achieve this. The aim is to preserve them in an agreed standard format in computer accessible form, to use them for international activities involving validation of current and new calculational schemes including computer codes and nuclear data libraries, for assessing uncertainties, confidence bounds and safety margins, and to record measurement methods and techniques. The databases so far established or in preparation related to nuclear data validation cover the following areas: SINBAD - A Radiation Shielding Experiments database encompassing reactor shielding, fusion blanket neutronics, and accelerator shielding. ICSBEP - International Criticality Safety Benchmark Experiments Project Handbook, with more than 2500 critical configurations with different combination of materials and spectral indices. IRPhEP - International Reactor Physics Experimental Benchmarks Evaluation Project. The different projects are described in the following including results achieved, work in progress and planned. (author)

Code Verification Capabilities and Assessments in Support of ASC V&V Level 2 Milestone #6035

Energy Technology Data Exchange (ETDEWEB)

Doebling, Scott William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Budzien, Joanne Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Ferguson, Jim Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Harwell, Megan Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hickmann, Kyle Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Israel, Daniel M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Magrogan, William Richard III [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Singleton, Jr., Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Srinivasan, Gowri [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Walter, Jr, John William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Woods, Charles Nathan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-09-26

This document provides a summary of the code verification activities supporting the FY17 Level 2 V&V milestone entitled “Deliver a Capability for V&V Assessments of Code Implementations of Physics Models and Numerical Algorithms in Support of Future Predictive Capability Framework Pegposts.” The physics validation activities supporting this milestone are documented separately. The objectives of this portion of the milestone are: 1) Develop software tools to support code verification analysis; 2) Document standard definitions of code verification test problems; and 3) Perform code verification assessments (focusing on error behavior of algorithms). This report and a set of additional standalone documents serve as the compilation of results demonstrating accomplishment of these objectives.

Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

Energy Technology Data Exchange (ETDEWEB)

Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Folsom, Charles Pearson [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-05-01

One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

International Nuclear Information System (INIS)

Williamson, R.L.

2011-01-01

Highlights: → The ABAQUS thermomechanics code is enhanced to enable simulation of nuclear fuel behavior. → Comparisons are made between discrete and smeared fuel pellet analysis. → Multidimensional and multipellet analysis is important for accurate prediction of PCMI. → Fully coupled thermomechanics results in very smooth prediction of fuel-clad gap closure. → A smeared-pellet approximation results in significant underprediction of clad radial displacements and plastic strain. - Abstract: A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO 2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

Understanding Mixed Code and Classroom Code-Switching: Myths and Realities

Science.gov (United States)

Li, David C. S.

2008-01-01

Background: Cantonese-English mixed code is ubiquitous in Hong Kong society, and yet using mixed code is widely perceived as improper. This paper presents evidence of mixed code being socially constructed as bad language behavior. In the education domain, an EDB guideline bans mixed code in the classroom. Teachers are encouraged to stick to…

What do conscientious people do? Development and validation of the Behavioral Indicators of Conscientiousness (BIC).

Science.gov (United States)

Jackson, Joshua J; Wood, Dustin; Bogg, Tim; Walton, Kate E; Harms, Peter D; Roberts, Brent W

2010-08-01

Typical assessments of personality traits collapse behaviors, thoughts, and feelings into a single measure without distinguishing between these different manifestations. To address this lack of specification, the current study develops and validates a measure that assesses a number of broad behaviors associated with the personality trait of conscientiousness (the Behavioral Indicators of Conscientiousness; BIC). Findings suggest that the lower-order structure of conscientious behaviors is mostly similar to the lower-order structure in extant trait measures. Furthermore, a daily diary method was used to validate the BIC against frequency counts of conscientious behavior. Overall, the results identify specific behaviors that conscientious individuals tend to perform and highlight possible advantages of this approach over broad trait assessment.

Thyc, a 3D thermal-hydraulic code for rod bundles. Recent developments and validation tests

International Nuclear Information System (INIS)

Caremoli, C.; Rascle, P.; Aubry, S.; Olive, J.

1993-09-01

PWR or LMFBR cores or fuel assemblies, PWR steam generators, condensers, tubular heat exchangers, are basic components of a nuclear power plant involving two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate DNB margins in reactor cores, singularity effects (grids, wire spacers, support plates, baffles), corrosion on steam generator tube sheet, bypass effects and vibration risks. For that purpose, Electricite de France has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two phase flows in rod or tube bundles (pressurized water reactor cores, steam generators, condensers, heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum and energy) of each phase over control volumes including fluid and solids. This paper briefly presents the physical model and the numerical method used in THYC. Then, validation tests (comparison with experiments) and applications (coupling with three-dimensional neutronics code and DNB predictions) are presented. They emphasize the last developments and new capabilities of the code. (authors). 10 figs., 3 tabs., 21 refs

Qualitative Content Analysis of Coworkers' Safety Reports of Unprofessional Behavior by Physicians and Advanced Practice Professionals.

Science.gov (United States)

Martinez, William; Pichert, James W; Hickson, Gerald B; Braddy, Casey H; Brown, Amy J; Catron, Thomas F; Moore, Ilene N; Stampfle, Morgan R; Webb, Lynn E; Cooper, William O

2018-03-15

The aims of the study were to develop a valid and reliable taxonomy of coworker reports of alleged unprofessional behavior by physicians and advanced practice professionals and determine the prevalence of reports describing particular types of unprofessional conduct. We conducted qualitative content analysis of coworker reports of alleged unprofessional behavior by physicians and advanced practice professionals to create a standardized taxonomy. We conducted a focus group of experts in medical professionalism to assess the taxonomy's face validity. We randomly selected 120 reports (20%) of the 590 total reports submitted through the medical center's safety event reporting system between June 2015 and September 2016 to measure interrater reliability of taxonomy codes and estimate the prevalence of reports describing particular types of conduct. The initial taxonomy contained 22 codes organized into the following four domains: competent medical care, clear and respectful communication, integrity, and responsibility. All 10 experts agreed that the four domains reflected essential elements of medical professionalism. Interrater reliabilities for all codes and domains had a κ value greater than the 0.60 threshold for good reliability. Most reports (60%, 95% confidence interval = 51%-69%) described disrespectful or offensive communication. Nine codes had a prevalence of less than 1% and were folded into their respective domains resulting in a final taxonomy composed of 13 codes. The final taxonomy represents a useful tool with demonstrated validity and reliability, opening the door for reliable analysis and systems to promote accountability and behavior change. Given the safety implications of unprofessional behavior, understanding the typology of coworker observations of unprofessional behavior may inform organization strategies to address this threat to patient safety.

SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

International Nuclear Information System (INIS)

Dias, A.F.V.

1980-10-01

Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt

Validation of computer code TRAFIC used for estimation of charcoal heatup in containment ventilation systems

International Nuclear Information System (INIS)

Yadav, D.H.; Datta, D.; Malhotra, P.K.; Ghadge, S.G.; Bajaj, S.S.

2005-01-01

Full text of publication follows: Standard Indian PHWRs are provided with a Primary Containment Filtration and Pump-Back System (PCFPB) incorporating charcoal filters in the ventilation circuit to remove radioactive iodine that may be released from reactor core into the containment during LOCA+ECCS failure which is a Design Basis Accident for containment of radioactive release. This system is provided with two identical air circulation loops, each having 2 full capacity fans (1 operating and 1 standby) for a bank of four combined charcoal and High Efficiency Particulate Activity (HEPA) filters, in addition to other filters. While the filtration circuit is designed to operate under forced flow conditions, it is of interest to understand the performance of the charcoal filters, in the event of failure of the fans after operating for some time, i.e., when radio-iodine inventory is at its peak value. It is of interest to check whether the buoyancy driven natural circulation occurring in the filtration circuit is sufficient enough to keep the temperature in the charcoal under safe limits. A computer code TRAFIC (Transient Analysis of Filters in Containment) was developed using conservative one dimensional model to analyze the system. Suitable parametric studies were carried out to understand the problem and to identify the safety of existing system. TRAFIC Code has two important components. The first one estimates the heat generation in charcoal filter based on 'Source Term'; while the other one performs thermal-hydraulic computations. In an attempt validate the Code, experimental studies have been carried out. For this purpose, an experimental set up comprising of scaled down model of filtration circuit with heating coils embedded in charcoal for simulating the heating effect due to radio iodine has been constructed. The present work of validation consists of utilizing the results obtained from experiments conducted for different heat loads, elevations and adsorbent

Criterion Validity of the Child's Challenging Behavior Scale, Version 2 (CCBS-2).

Science.gov (United States)

Bourke-Taylor, Helen M; Cordier, Reinie; Pallant, Julie F

The Child's Challenging Behavior Scale, Version 2 (CCBS-2), measures maternal rating of a child's challenging behaviors that compromise maternal mental health. The CCBS-2, the Child Behavior Checklist (CBCL), and the Strengths and Difficulties Questionnaire (SDQ) were compared in a sample of typically developing young Australian children. Criterion validity was investigated by correlating the CCBS-2 with "gold standard" measures (CBCL and SDQ subscales). Data were collected in a cross-sectional survey of mothers (N = 336) of children ages 3-9 yr. Correlations with the CBCL externalizing subscales demonstrated moderate (ρ = .46) to strong (ρ = .66) correlations. Correlations with the SDQ externalizing behaviors subscales were moderate (ρ = .35) to strong (ρ = .60). The criterion validity established in this study strengthens the psychometric properties that support ongoing development of the CCBS-2 as an efficient tool that may identify children in need of further evaluation. Copyright © 2018 by the American Occupational Therapy Association, Inc.

Computer code validation study of PWR core design system, CASMO-3/MASTER-α

International Nuclear Information System (INIS)

Lee, K. H.; Kim, M. H.; Woo, S. W.

1999-01-01

In this paper, the feasibility of CASMO-3/MASTER-α nuclear design system was investigated for commercial PWR core. Validation calculation was performed as follows. Firstly, the accuracy of cross section generation from table set using linear feedback model was estimated. Secondly, the results of CASMO-3/MASTER-α was compared with CASMO-3/NESTLE 5.02 for a few benchmark problems. Microscopic cross sections computed from table set were almost the same with those from CASMO-3. There were small differences between calculated results of two code systems. Thirdly, the repetition of CASMO-3/MASTER-α calculation for Younggwang Unit-3, Cycle-1 core was done and their results were compared with nuclear design report(NDR) and uncertainty analysis results of KAERI. It was found that uncertainty analysis results were reliable enough because results were agreed each other. It was concluded that the use of nuclear design system CASMO-3/MASTER-α was validated for commercial PWR core

Pellet injection and plasma behavior simulation code PEPSI

International Nuclear Information System (INIS)

Takase, Haruhiko; Tobita, Kenji; Nishio, Satoshi

2003-08-01

Fueling is one of the major issues on design of nuclear fusion reactor and the injection of solid hydrogen pellet to the core plasma is a useful method. On the design of a nuclear fusion reactor, it is necessary to determine requirements on the pellet size, the number of pellets, the injection speed and the injection cycle. PEllet injection and Plasma behavior SImulation code PEPSI has been developed to assess these parameters. PEPSI has two special features: 1) Adopting two numerical pellet models, Parks model and Strauss model, 2) Calculating fusion power and other plasma parameters in combination with a time-dependent one-dimensional transport model. This report describes the numerical models, numerical scheme, sequence of calculation, list of subroutines, list of variables and an example of calculation. (author)

Living Up to the Code's Exhortations? Social Workers' Political Knowledge Sources, Expectations, and Behaviors.

Science.gov (United States)

Felderhoff, Brandi Jean; Hoefer, Richard; Watson, Larry Dan

2016-01-01

The National Association of Social Workers' (NASW's) Code of Ethics urges social workers to engage in political action. However, little recent research has been conducted to examine whether social workers support this admonition and the extent to which they actually engage in politics. The authors gathered data from a survey of social workers in Austin, Texas, to address three questions. First, because keeping informed about government and political news is an important basis for action, the authors asked what sources of knowledge social workers use. Second, they asked what the respondents believe are appropriate political behaviors for other social workers and NASW. Third, they asked for self-reports regarding respondents' own political behaviors. Results indicate that social workers use the Internet and traditional media services to stay informed; expect other social workers and NASW to be active; and are, overall, more active than the general public in many types of political activities. The comparisons made between expectations for others and their own behaviors are interesting in their complex outcomes. Social workers should strive for higher levels of adherence to the code's urgings on political activity. Implications for future work are discussed.

Validation of TEMP: A finite line heat transfer code for geologic repositories for nuclear waste

International Nuclear Information System (INIS)

Atterbury, W.G.; Hetteburg, J.R.; Wurm, K.J.

1987-09-01

TEMP is a FORTRAN computer code for calculating temperatures in a geologic repository for nuclear waste. A previous report discusses the structure, usage, verification, and benchmarking of TEMP V1.0 (Wurm et al., 1987). This report discusses modifications to the program in the development of TEMP V1.1 and documents the validation of TEMP. The development of TEMP V1.1 from TEMP V1.0 consisted of two major efforts. The first was to recode several of the subroutines to improve logic flow and to allow for geometry-independent temperature calculation routines which, in turn, allowed for the addition of the geometry-independent validation option. The validation option provides TEMP with the ability to model any geometry of temperature sources with any step-wise heat release rate. This capability allows TEMP to model the geometry and heat release characteristics of the validation problems. The validation of TEMP V1.1 consists of the comparison of TEMP to three in-ground heater tests. The three tests chosen were Avery Island, Louisiana, Site A; Avery Island, Louisiana, Site C; and Asse Mine, Federal Republic of Germany, Site 2. TEMP shows marginal comparison with the two Avery Island sites and good comparison with the Asse Mine Site. 8 refs., 25 figs., 14 tabs

Validity and Reliability of Knowledge, Attitude and Behavior Assessment Tool Among Vulnerable Women Concerning Sexually Transmitted Diseases

Directory of Open Access Journals (Sweden)

Zahra Boroumandfar

2016-05-01

Full Text Available Objective: The study aimed to design and evaluate the content and face validity, and reliability of knowledge, attitude, and behavior questionnaire on preventive behaviors among vulnerable women concerning sexually transmitted diseases (STDs.Materials and methods: This cross-sectional study was carried out in two phases of an action research. In the first phase, to explain STDs preventive domains, 20 semi- structured interviews were conducted with the vulnerable women, residing at women prison and women referred to counseling centers. After analyzing content of interviews, three domains were identified: improve their knowledge, modify their attitude and change their behaviors. In the second phase, the questionnaire was designed and tested in a pilot study. Then, its content validity was evaluated. Face validity and reliability of the questionnaire were assessed by test re- test method and Cronbach alpha respectively.Results: Index of content validity in each three domain of the questionnaire (knowledge, attitude and behavior concerning STDs was obtained over 0.6. Overall content validity index was 0.86 in all three domains of the questionnaire. The Cronbach’s alpha as reliability of questionnaire was 0.80 for knowledge, 0.79 for attitude and 0.85 for behavior.Conclusion: The results showed that the designed questionnaire was a valid and reliable tool to measure knowledge, attitude and behavior of vulnerable women, predisposed to risk of STDs.

WIMS-AECL/RFSP code validation of reactivity calculations following a long shutdown using the simple-cell history-based method

International Nuclear Information System (INIS)

Ardeshiri, F.; Donnelly, J.V.; Arsenault, B.

1998-01-01

The purpose of this analysis is to validate the Reactor Fuelling Simulation Program (RFSP) using the simple-cell model (SCM) history-based method in a startup simulation following a reactor shutdown period. This study is part of the validation work for history-based calculations, using the WIMS-AECL code with the ENDF/B-V library, and the SCM linked to the RFSP code. In this work, the RFSP code with the SCM history-based method was used to track a 1-year period of the Point Lepreau reactor operating history, that included a 12-day reactor shutdown and subsequent startup. Measured boron and gadolinium concentrations were used in the RFSP simulations, and the predicted values of core reactivity were compared to the reference (pre-shutdown) value. The discrepancies in core reactivity are shown to be better than ±2 milli-k at any time, and better than about ±0.5 milli-k towards the end of the startup transient. The results of this analysis also show that the calculated maximum channel and bundle powers are within an acceptable range during both the core-follow and the reactor startup simulations. (author)

Children's Assertive Behavior: The Reliability and Validity of Three Self-Report Measures.

Science.gov (United States)

Scanlon, Elizabeth M.; Ollendick, Thomas H.

1985-01-01

The internal consistency and validity of three new scales for measuring assertiveness in children were tested. Two of the scales were able to "unbind" aggressive from assertive behavior, while the third was able to "unbind" submissive from assertive behavior. At present, a combination of the three scales is recommended. (KH)

MARS Validation Plan and Status

International Nuclear Information System (INIS)

Ahn, Seung-hoon; Cho, Yong-jin

2008-01-01

The KINS Reactor Thermal-hydraulic Analysis System (KINS-RETAS) under development is directed toward a realistic analysis approach of best-estimate (BE) codes and realistic assumptions. In this system, MARS is pivoted to provide the BE Thermal-Hydraulic (T-H) response in core and reactor coolant system to various operational transients and accidental conditions. As required for other BE codes, the qualification is essential to ensure reliable and reasonable accuracy for a targeted MARS application. Validation is a key element of the code qualification, and determines the capability of a computer code in predicting the major phenomena expected to occur. The MARS validation was made by its developer KAERI, on basic premise that its backbone code RELAP5/MOD3.2 is well qualified against analytical solutions, test or operational data. A screening was made to select the test data for MARS validation; some models transplanted from RELAP5, if already validated and found to be acceptable, were screened out from assessment. It seems to be reasonable, but does not demonstrate whether code adequacy complies with the software QA guidelines. Especially there may be much difficulty in validating the life-cycle products such as code updates or modifications. This paper presents the plan for MARS validation, and the current implementation status

Quasi-3d aerodynamic code for analyzing dynamic flap response

DEFF Research Database (Denmark)

Ramos García, Néstor

A computational model for predicting the aerodynamic behavior of wind turbine airfoil profiles subjected to steady and unsteady motions has been developed. The model is based on a viscous-inviscid interaction technique using strong coupling between the viscous and inviscid parts. The inviscid part...... transition model. Validation of the steady two dimensional version of the code has been carried out against experiments for different airfoil geometries and Reynolds numbers. The unsteady version of the code has been benchmarked against experiments for different airfoil geometries at various reduced...... frequencies and oscillation amplitudes, and generally a good agreement is obtained. The capability of the code to simulate a trailing edge flap under steady or unsteady flow conditions has been proven. A parametric study on rotational effects induced by Coriolis and centrifugal forces in the boundary layer...

Short-Term Predictive Validity of Cluster Analytic and Dimensional Classification of Child Behavioral Adjustment in School

Science.gov (United States)

Kim, Sangwon; Kamphaus, Randy W.; Baker, Jean A.

2006-01-01

A constructive debate over the classification of child psychopathology can be stimulated by investigating the validity of different classification approaches. We examined and compared the short-term predictive validity of cluster analytic and dimensional classifications of child behavioral adjustment in school using the Behavior Assessment System…

The RAGE radiation-hydrodynamic code

International Nuclear Information System (INIS)

Gittings, Michael; Clover, Michael; Betlach, Thomas; Byrne, Nelson; Ranta, Dale; Weaver, Robert; Coker, Robert; Dendy, Edward; Hueckstaedt, Robert; New, Kim; Oakes, W Rob; Stefan, Ryan

2008-01-01

We describe RAGE, the 'radiation adaptive grid Eulerian' radiation-hydrodynamics code, including its data structures, its parallelization strategy and performance, its hydrodynamic algorithm(s), its (gray) radiation diffusion algorithm, and some of the considerable amount of verification and validation efforts. The hydrodynamics is a basic Godunov solver, to which we have made significant improvements to increase the advection algorithm's robustness and to converge stiffnesses in the equation of state. Similarly, the radiation transport is a basic gray diffusion, but our treatment of the radiation-material coupling, wherein we converge nonlinearities in a novel manner to allow larger timesteps and more robust behavior, can be applied to any multi-group transport algorithm

A Mode Propagation Database Suitable for Code Validation Utilizing the NASA Glenn Advanced Noise Control Fan and Artificial Sources

Science.gov (United States)

Sutliff, Daniel L.

2014-01-01

The NASA Glenn Research Center's Advanced Noise Control Fan (ANCF) was developed in the early 1990s to provide a convenient test bed to measure and understand fan-generated acoustics, duct propagation, and radiation to the farfield. A series of tests were performed primarily for the use of code validation and tool validation. Rotating Rake mode measurements were acquired for parametric sets of: (i) mode blockage, (ii) liner insertion loss, (iii) short ducts, and (iv) mode reflection.

The Physician Recommendation Coding System (PhyReCS): A Reliable and Valid Method to Quantify the Strength of Physician Recommendations During Clinical Encounters.

Science.gov (United States)

Scherr, Karen A; Fagerlin, Angela; Williamson, Lillie D; Davis, J Kelly; Fridman, Ilona; Atyeo, Natalie; Ubel, Peter A

2017-01-01

Physicians' recommendations affect patients' treatment choices. However, most research relies on physicians' or patients' retrospective reports of recommendations, which offer a limited perspective and have limitations such as recall bias. To develop a reliable and valid method to measure the strength of physician recommendations using direct observation of clinical encounters. Clinical encounters (n = 257) were recorded as part of a larger study of prostate cancer decision making. We used an iterative process to create the 5-point Physician Recommendation Coding System (PhyReCS). To determine reliability, research assistants double-coded 50 transcripts. To establish content validity, we used 1-way analyses of variance to determine whether relative treatment recommendation scores differed as a function of which treatment patients received. To establish concurrent validity, we examined whether patients' perceived treatment recommendations matched our coded recommendations. The PhyReCS was highly reliable (Krippendorf's alpha = 0.89, 95% CI [0.86, 0.91]). The average relative treatment recommendation score for each treatment was higher for individuals who received that particular treatment. For example, the average relative surgery recommendation score was higher for individuals who received surgery versus radiation (mean difference = 0.98, SE = 0.18, P recommendations matched coded recommendations 81% of the time. The PhyReCS is a reliable and valid way to capture the strength of physician recommendations. We believe that the PhyReCS would be helpful for other researchers who wish to study physician recommendations, an important part of patient decision making. © The Author(s) 2016.

Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

Energy Technology Data Exchange (ETDEWEB)

Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Laboratorio de Analise, Avaliacao e Gerenciamento de Risco (LabRisco/POLI/USP), São Paulo, SP (Brazil); Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: ayabe@ipen.br, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

International Nuclear Information System (INIS)

Giovedi, Claudia; Martins, Marcelo Ramos; Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e

2017-01-01

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Development and validation of a criticality calculation scheme based on French deterministic transport codes

International Nuclear Information System (INIS)

Santamarina, A.

1991-01-01

A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)

Assessment of MARMOT. A Mesoscale Fuel Performance Code

Energy Technology Data Exchange (ETDEWEB)

Tonks, M. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, X. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fromm, B. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yu, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Teague, M. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-04-01

MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. In this report, we provide a high level summary of MARMOT, its capabilities, and its current state of validation. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding. It accomplished this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules. However, additional capability specific to fuel and cladding is available in MARMOT. While some validation of MARMOT has been completed in the areas of fission gas behavior and grain growth, much more validation needs to be conducted. However, new mesoscale data needs to be obtained in order to complete this validation.

LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

International Nuclear Information System (INIS)

Bordner, G.L.

1979-10-01

The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

Code-modulated interferometric imaging system using phased arrays

Science.gov (United States)

Chauhan, Vikas; Greene, Kevin; Floyd, Brian

2016-05-01

Millimeter-wave (mm-wave) imaging provides compelling capabilities for security screening, navigation, and bio- medical applications. Traditional scanned or focal-plane mm-wave imagers are bulky and costly. In contrast, phased-array hardware developed for mass-market wireless communications and automotive radar promise to be extremely low cost. In this work, we present techniques which can allow low-cost phased-array receivers to be reconfigured or re-purposed as interferometric imagers, removing the need for custom hardware and thereby reducing cost. Since traditional phased arrays power combine incoming signals prior to digitization, orthogonal code-modulation is applied to each incoming signal using phase shifters within each front-end and two-bit codes. These code-modulated signals can then be combined and processed coherently through a shared hardware path. Once digitized, visibility functions can be recovered through squaring and code-demultiplexing operations. Pro- vided that codes are selected such that the product of two orthogonal codes is a third unique and orthogonal code, it is possible to demultiplex complex visibility functions directly. As such, the proposed system modulates incoming signals but demodulates desired correlations. In this work, we present the operation of the system, a validation of its operation using behavioral models of a traditional phased array, and a benchmarking of the code-modulated interferometer against traditional interferometer and focal-plane arrays.

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

International Nuclear Information System (INIS)

Leskovar, M.; Mavko, B.

1998-01-01

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

Fission product behavior in the Peach Bottom and Fort St. Vrain HTGRs

International Nuclear Information System (INIS)

Hanson, D.L.; Baldwin, N.L.; Strong, D.E.

1980-11-01

Actual operating data from Peach Bottom and Fort St. Vrain were compared with code predictions to assess the validity of the methods used to predict the behavior of fission products in the primary coolant circuit. For both reactors the measured circuit activities were significantly below design values, and the observations generally verify the codes used for large HTGR design

Development and validation of a low-frequency modeling code for high-moment transmitter rod antennas

Science.gov (United States)

Jordan, Jared Williams; Sternberg, Ben K.; Dvorak, Steven L.

2009-12-01

The goal of this research is to develop and validate a low-frequency modeling code for high-moment transmitter rod antennas to aid in the design of future low-frequency TX antennas with high magnetic moments. To accomplish this goal, a quasi-static modeling algorithm was developed to simulate finite-length, permeable-core, rod antennas. This quasi-static analysis is applicable for low frequencies where eddy currents are negligible, and it can handle solid or hollow cores with winding insulation thickness between the antenna's windings and its core. The theory was programmed in Matlab, and the modeling code has the ability to predict the TX antenna's gain, maximum magnetic moment, saturation current, series inductance, and core series loss resistance, provided the user enters the corresponding complex permeability for the desired core magnetic flux density. In order to utilize the linear modeling code to model the effects of nonlinear core materials, it is necessary to use the correct complex permeability for a specific core magnetic flux density. In order to test the modeling code, we demonstrated that it can accurately predict changes in the electrical parameters associated with variations in the rod length and the core thickness for antennas made out of low carbon steel wire. These tests demonstrate that the modeling code was successful in predicting the changes in the rod antenna characteristics under high-current nonlinear conditions due to changes in the physical dimensions of the rod provided that the flux density in the core was held constant in order to keep the complex permeability from changing.

Predictive Validity of a Student Self-Report Screener of Behavioral and Emotional Risk in an Urban High School

Science.gov (United States)

Dowdy, Erin; Harrell-Williams, Leigh; Dever, Bridget V.; Furlong, Michael J.; Moore, Stephanie; Raines, Tara; Kamphaus, Randy W.

2016-01-01

Increasingly, schools are implementing school-based screening for risk of behavioral and emotional problems; hence, foundational evidence supporting the predictive validity of screening instruments is important to assess. This study examined the predictive validity of the Behavior Assessment System for Children-2 Behavioral and Emotional Screening…

EVLncRNAs: a manually curated database for long non-coding RNAs validated by low-throughput experiments

Science.gov (United States)

Zhao, Huiying; Yu, Jiafeng; Guo, Chengang; Dou, Xianghua; Song, Feng; Hu, Guodong; Cao, Zanxia; Qu, Yuanxu

2018-01-01

Abstract Long non-coding RNAs (lncRNAs) play important functional roles in various biological processes. Early databases were utilized to deposit all lncRNA candidates produced by high-throughput experimental and/or computational techniques to facilitate classification, assessment and validation. As more lncRNAs are validated by low-throughput experiments, several databases were established for experimentally validated lncRNAs. However, these databases are small in scale (with a few hundreds of lncRNAs only) and specific in their focuses (plants, diseases or interactions). Thus, it is highly desirable to have a comprehensive dataset for experimentally validated lncRNAs as a central repository for all of their structures, functions and phenotypes. Here, we established EVLncRNAs by curating lncRNAs validated by low-throughput experiments (up to 1 May 2016) and integrating specific databases (lncRNAdb, LncRANDisease, Lnc2Cancer and PLNIncRBase) with additional functional and disease-specific information not covered previously. The current version of EVLncRNAs contains 1543 lncRNAs from 77 species that is 2.9 times larger than the current largest database for experimentally validated lncRNAs. Seventy-four percent lncRNA entries are partially or completely new, comparing to all existing experimentally validated databases. The established database allows users to browse, search and download as well as to submit experimentally validated lncRNAs. The database is available at http://biophy.dzu.edu.cn/EVLncRNAs. PMID:28985416

In-vessel core degradation code validation matrix update 1996-1999. Report by an OECD/NEA group of experts

International Nuclear Information System (INIS)

2001-02-01

In 1991 the Committee on the Safety of Nuclear Installations (CSNI) issued a State-of-the-Art Report (SOAR) on In-Vessel Core Degradation in Light Water Reactor (LWR) Severe Accidents. Based on the recommendations of this report a Validation Matrix for severe accident modelling codes was produced. Experiments performed up to the end of 1993 were considered for this validation matrix. To include recent experiments and to enlarge the scope, an update was formally inaugurated in January 1999 by the Task Group on Degraded Core Cooling, a sub-group of Principal Working Group 2 (PWG-2) on Coolant System Behaviour, and a selection of writing group members was commissioned. The present report documents the results of this study. The objective of the Validation Matrix is to define a basic set of experiments, for which comparison of the measured and calculated parameters forms a basis for establishing the accuracy of test predictions, covering the full range of in-vessel core degradation phenomena expected in light water reactor severe accident transients. The emphasis is on integral experiments, where interactions amongst key phenomena as well as the phenomena themselves are explored; however separate-effects experiments are also considered especially where these extend the parameter ranges to cover those expected in postulated LWR severe accident transients. As well as covering PWR and BWR designs of Western origin, the scope of the review has been extended to Eastern European (VVER) types. Similarly, the coverage of phenomena has been extended, starting as before from the initial heat-up but now proceeding through the in-core stage to include introduction of melt into the lower plenum and further to core coolability and retention to the lower plenum, with possible external cooling. Items of a purely thermal hydraulic nature involving no core degradation are excluded, having been covered in other validation matrix studies. Concerning fission product behaviour, the effect

Reactor Fuel Isotopics and Code Validation for Nuclear Applications

Energy Technology Data Exchange (ETDEWEB)

Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-02-01

Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for ¹³³Cs and ¹⁵³Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

Verification and Validation of The Tritium Transport Code TMAP7

International Nuclear Information System (INIS)

Glen R. Longhurst; James Ambrosek

2004-01-01

The TMAP Code was written at the Idaho National Engineering and Environmental Laboratory in the late 1980s as a tool for safety analysis of systems involving tritium. Since then it has been upgraded several times and has been used in numerous applications including experiments supporting fusion safety, predictions for advanced systems such as the International Thermonuclear Experimental Reactor (ITER), and estimates involving tritium production technologies. Its most recent upgrade to TMAP7 was accomplished in response to several needs. Prior versions had the capacity to deal with only a single trap for diffusing gaseous species in solid structures. TMAP7 includes up to three separate traps and up to 10 diffusing species. The original code had difficulty dealing with heteronuclear molecule formation such as HD and DT. That has been removed. Under pre-specified boundary enclosure conditions and solution-law dependent diffusion boundary conditions, such as Sieverts' law, TMAP7 automatically generates heteronuclear molecular partial pressures when solubilities and partial pressures of the homonuclear molecular species are provided for law-dependent diffusion boundary conditions. A further sophistication is the addition of non-diffusing surface species. Atoms such as oxygen or nitrogen or formation of hydroxyl radicals on metal surfaces are sometimes important in molecule formation with diffusing hydrogen isotopes but do not, themselves, diffuse appreciably in the material. TMAP7 will accommodate up to 30 such surface species, allowing the user to specify relationships between those surface concentrations and partial pressures of gaseous species above the surfaces or to form them dynamically by combining diffusion species or other surface species. Additionally, TMAP7 allows the user to include a surface binding energy and an adsorption barrier energy and includes asymmetrical diffusion between the surface sites and regular diffusion sites in the bulk. All of the

Validation of the Open Source Code_Aster Software Used in the Modal Analysis of the Fluid-filled Cylindrical Shell

Directory of Open Access Journals (Sweden)

B D. Kashfutdinov

2017-01-01

Full Text Available The paper deals with a modal analysis of the elastic cylindrical shell with a clamped bottom partially filled with fluid in open source Code_Aster software using the finite element method. Natural frequencies and modes obtained in Code_Aster are compared to experimental and theoretical data. The aim of this paper is to prove that Code_Aster has all necessary tools for solving fluid structure interaction problems. Also, Code_Aster can be used in the industrial projects as an alternative to commercial software. The available free pre- and post-processors with a graphical user interface that is compatible with Code_Aster allow creating complex models and processing the results.The paper presents new validation results of open source Code_Aster software used to calculate small natural modes of the cylindrical shell partially filled with non-viscous compressible barotropic fluid under gravity field.The displacement of the middle surface of thin shell and the displacement of the fluid relative to the equilibrium position are described by coupled hydro-elasticity problem. The fluid flow is considered to be potential. The finite element method (FEM is used. The features of computational model are described. The resolution equation has symmetrical block matrices. To compare the results, is discussed the well-known modal analysis problem of cylindrical shell with flat non-deformable bottom, filled with a compressible fluid. The numerical parameters of the scheme were chosen in accordance with well-known experimental and analytical data. Three cases were taken into account: an empty, a partially filled and a full-filled cylindrical shell.The frequencies of Code_Aster are in good agreement with those, obtained in experiment, analytical solution, as well as with results obtained by FEM in other software. The difference between experiment and analytical solution in software is approximately the same. The obtained results extend a set of validation tests for

Examining the Reliability and Validity of the Effective Behavior Support Self-Assessment Survey

Science.gov (United States)

Solomon, Benjamin G.; Tobin, Kevin G.; Schutte, Gregory M.

2015-01-01

The Effective Behavior Support Self-Assessment Survey (SAS; Sugai, Horner, & Todd, 2003) is designed to measure perceived Positive Behavior Interventions and Supports (PBIS) implementation and identify priorities for improvement. Despite its longevity, little published research exists documenting its reliability or validity for these purposes.…

FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Sakagami, M.

1984-01-01

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Experiences using IAEA Code of practice for radiation sterilization of tissue allografts: Validation and routine control

Energy Technology Data Exchange (ETDEWEB)

Hilmy, N. [Batan Research Tissue Bank (BRTB), Centre for Research and Development of Isotopes and Radiation Technology, P.O. Box 7002, JKSKL, Jakarta 12070 (Indonesia)], E-mail: nazly@batan.go.id; Febrida, A.; Basril, A. [Batan Research Tissue Bank (BRTB), Centre for Research and Development of Isotopes and Radiation Technology, P.O. Box 7002, JKSKL, Jakarta 12070 (Indonesia)

2007-11-15

Problems of tissue allografts in using International Standard (ISO) 11137 for validation of radiation sterilization dose (RSD) are limited and low numbers of uniform samples per production batch, those are products obtained from one donor. Allograft is a graft transplanted between two different individuals of the same species. The minimum number of uniform samples needed for verification dose (VD) experiment at the selected sterility assurance level (SAL) per production batch according to the IAEA Code is 20, i.e., 10 for bio-burden determination and the remaining 10 for sterilization test. Three methods of the IAEA Code have been used for validation of RSD, i.e., method A1 that is a modification of method 1 of ISO 11137:1995, method B (ISO 13409:1996), and method C (AAMI TIR 27:2001). This paper describes VD experiments using uniform products obtained from one cadaver donor, i.e., cancellous bones, demineralized bone powders and amnion grafts from one life donor. Results of the verification dose experiments show that RSD is 15.4 kGy for cancellous and demineralized bone grafts and 19.2 kGy for amnion grafts according to method A1 and 25 kGy according to methods B and C.

Real-time assessment of perioperative behaviors in children and parents: development and validation of the perioperative adult child behavioral interaction scale.

Science.gov (United States)

Sadhasivam, Senthilkumar; Cohen, Lindsey L; Hosu, Liana; Gorman, Kristin L; Wang, Yu; Nick, Todd G; Jou, Jing Fang; Samol, Nancy; Szabova, Alexandra; Hagerman, Nancy; Hein, Elizabeth; Boat, Anne; Varughese, Anna; Kurth, Charles Dean; Willging, J Paul; Gunter, Joel B

2010-04-01

Behavior in response to distressful events during outpatient pediatric surgery can contribute to postoperative maladaptive behaviors, such as temper tantrums, nightmares, bed-wetting, and attention seeking. Currently available perioperative behavioral assessment tools have limited utility in guiding interventions to ameliorate maladaptive behaviors because they cannot be used in real time, are only intended to be used during 1 phase of the experience (e.g., perioperative), or provide only a static assessment of the child (e.g., level of anxiety). A simple, reliable, real-time tool is needed to appropriately identify children and parents whose behaviors in response to distressful events at any point in the perioperative continuum could benefit from timely behavioral intervention. Our specific aims were to (1) refine the Perioperative Adult Child Behavioral Interaction Scale (PACBIS) to improve its reliability in identifying perioperative behaviors and (2) validate the refined PACBIS against several established instruments. The PACBIS was used to assess the perioperative behaviors of 89 children aged 3 to 12 years presenting for adenotonsillectomy and their parents. Assessments using the PACBIS were made during perioperative events likely to prove distressing to children and/or parents (perioperative measurement of blood pressure, induction of anesthesia, and removal of the IV catheter before discharge). Static measurements of perioperative anxiety and behavioral compliance during anesthetic induction were made using the modified Yale Preoperative Anxiety Scale and the Induction Compliance Checklist (ICC). Each event was videotaped for later scoring using the Child-Adult Medical Procedure Interaction Scale-Short Form (CAMPIS-SF) and Observational Scale of Behavioral Distress (OSBD). Interrater reliability using linear weighted kappa (kappa(w)) and multiple validations using Spearman correlation coefficients were analyzed. The PACBIS demonstrated good to excellent

TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

International Nuclear Information System (INIS)

Chalhoub, E.S.

1980-09-01

A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt

Investigation of a two-phase nozzle flow and validation of several computer codes by the experimental data

International Nuclear Information System (INIS)

Kedziur, F.

1980-03-01

Stationary experiments with a convergent nozzle are performed in order to validate advanced two-phase computer codes, which find application in the blowdown-phase of a loss-of-coolant accident (LOCA). The steam/water flow presents a broad variety of initial conditions: The pressure varies between 2 and 13 MPa, the void fraction between 0 (subcooled) and about 80%, a great number of subcritical as well as critical experiments with different flow pattern is investigated. Additional air/water experiments serve for the separation of phase transition effects. The transient acceleration of the fluid in the LOCA-case is simulated by a local acceleration in the experiments. The layout of the nozzle and the applied measurement technique allow for a separate testing of physical models and the determination of empirical model parameters, respectively: In the four codes DUESE, DRIX-20, RELAP4/MOD6 and STRUYA the models - if they exist - for slip between the phases, thermodynamic non-equilibrium, pipe friction and critical mass flow rate are validated and criticised in comparison with the experimental data, and the corresponding model parameters are determined. The parameters essentially are a function of the void fraction. (orig.) [de

CONTRANS 2 code conversion from Apollo to HP

International Nuclear Information System (INIS)

Lee, Hae Cho

1996-01-01

CONTRANS2 computer code is used to calculate transient thermal hydraulic responses of containment building to loss of coolant and main steam line break accident. Mass and energy release to the containment following an accident are code inputs. This report firstly describes detailed work carried out for installation of CONTRANS2 on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also describes in relation to installation of CONTRANS2 on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

Verification and validation benchmarks.

Energy Technology Data Exchange (ETDEWEB)

Oberkampf, William Louis; Trucano, Timothy Guy

2007-02-01

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

Real-time validation of receiver state information in optical space-time block code systems.

Science.gov (United States)

Alamia, John; Kurzweg, Timothy

2014-06-15

Free space optical interconnect (FSOI) systems are a promising solution to interconnect bottlenecks in high-speed systems. To overcome some sources of diminished FSOI performance caused by close proximity of multiple optical channels, multiple-input multiple-output (MIMO) systems implementing encoding schemes such as space-time block coding (STBC) have been developed. These schemes utilize information pertaining to the optical channel to reconstruct transmitted data. The STBC system is dependent on accurate channel state information (CSI) for optimal system performance. As a result of dynamic changes in optical channels, a system in operation will need to have updated CSI. Therefore, validation of the CSI during operation is a necessary tool to ensure FSOI systems operate efficiently. In this Letter, we demonstrate a method of validating CSI, in real time, through the use of moving averages of the maximum likelihood decoder data, and its capacity to predict the bit error rate (BER) of the system.

Validation of the metal fuel version of the SAS4A accident analysis code

International Nuclear Information System (INIS)

Tentner, A.M.

1991-01-01

This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in Liquid Metal-Cooled Reactors (LMR), designed to operate in a fast neutron spectrum. SAS4A was initially developed for the analysis of oxide-fueled liquid metal-cooled reactors and has played an important role in the simulation and assessment of the energetics potential for postulated severe accidents in these reactors. Due to the current interest in the metal-fueled liquid metal-cooled reactors, a metal fuel version of the SAS4A accident analysis code is being developed in the Integral Fast Reactor program at Argonne. During such postulated accident scenarios as the unprotected (i.e. without scram) loss-of-flow and transient overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling, and fuel and cladding melting and relocation. Due to strong neutronic feedbacks these events can significantly influence the reactor power history in the accident progression. The paper presents the results of a recent SAS4A simulation of the M7 TREAT experiment. 6 refs., 5 figs

Validation study of computer code SPHINCS for sodium fire safety evaluation of fast reactor

International Nuclear Information System (INIS)

Yamaguchi, Akira; Tajima, Yuji

2003-01-01

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

Thermomechanical DART code improvements for LEU VHD dispersion and monolithic fuel element analysis

International Nuclear Information System (INIS)

Taboada, H.; Saliba, R.; Moscarda, M.V.; Rest, J.

2005-01-01

A collaboration agreement between ANL/US DOE and CNEA Argentina in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual FASTDART version and also a DART THERMAL version were presented during RERTR 2000, 2002 and RERTR 2003 Meetings. During this past year the following activities were completed: Optimization of DART TM code Al diffusion parameters by testing predictions against reliable data from RERTR experiments. Improvements on the 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic fuel. Concerning the first point, by means of an optimization of parameters of the Al diffusion through the interaction product theoretical expression, a reasonable agreement between DART temperature calculations with reliable RERTR PIE data was reached. The 3-D thermomechanical code complex is based upon a finite element thermal-elastic code named TERMELAS, and irradiation behavior provided by the DART code. An adequate and progressive process of coupling calculations of both codes at each time step is currently developed. Compatible thermal calculation between both codes was reached. This is the first stage to benchmark and validate against RERTR PIE data the coupling process. (author)

Experimental benchmark of non-local-thermodynamic-equilibrium plasma atomic physics codes; Validation experimentale des codes de physique atomique des plasmas hors equilibre thermodynamique local

Energy Technology Data Exchange (ETDEWEB)

Nagels-Silvert, V

2004-09-15

The main purpose of this thesis is to get experimental data for the testing and validation of atomic physics codes dealing with non-local-thermodynamical-equilibrium plasmas. The first part is dedicated to the spectroscopic study of xenon and krypton plasmas that have been produced by a nanosecond laser pulse interacting with a gas jet. A Thomson scattering diagnostic has allowed us to measure independently plasma parameters such as electron temperature, electron density and the average ionisation state. We have obtained time integrated spectra in the range between 5 and 10 angstroms. We have identified about one hundred xenon rays between 8.6 and 9.6 angstroms via the use of the Relac code. We have discovered unknown rays for the krypton between 5.2 and 7.5 angstroms. In a second experiment we have extended the wavelength range to the X UV domain. The Averroes/Transpec code has been tested in the ranges from 9 to 15 angstroms and from 10 to 130 angstroms, the first range has been well reproduced while the second range requires a more complex data analysis. The second part is dedicated to the spectroscopic study of aluminium, selenium and samarium plasmas in femtosecond operating rate. We have designed an interferometry diagnostic in the frequency domain that has allowed us to measure the expanding speed of the target's backside. Via the use of an adequate isothermal model this parameter has led us to know the plasma electron temperature. Spectra and emission times of various rays from the aluminium and selenium plasmas have been computed satisfactorily with the Averroes/Transpec code coupled with Film and Multif hydrodynamical codes. (A.C.)

Experimental benchmark of non-local-thermodynamic-equilibrium plasma atomic physics codes; Validation experimentale des codes de physique atomique des plasmas hors equilibre thermodynamique local

Energy Technology Data Exchange (ETDEWEB)

Nagels-Silvert, V

2004-09-15

The main purpose of this thesis is to get experimental data for the testing and validation of atomic physics codes dealing with non-local-thermodynamical-equilibrium plasmas. The first part is dedicated to the spectroscopic study of xenon and krypton plasmas that have been produced by a nanosecond laser pulse interacting with a gas jet. A Thomson scattering diagnostic has allowed us to measure independently plasma parameters such as electron temperature, electron density and the average ionisation state. We have obtained time integrated spectra in the range between 5 and 10 angstroms. We have identified about one hundred xenon rays between 8.6 and 9.6 angstroms via the use of the Relac code. We have discovered unknown rays for the krypton between 5.2 and 7.5 angstroms. In a second experiment we have extended the wavelength range to the X UV domain. The Averroes/Transpec code has been tested in the ranges from 9 to 15 angstroms and from 10 to 130 angstroms, the first range has been well reproduced while the second range requires a more complex data analysis. The second part is dedicated to the spectroscopic study of aluminium, selenium and samarium plasmas in femtosecond operating rate. We have designed an interferometry diagnostic in the frequency domain that has allowed us to measure the expanding speed of the target's backside. Via the use of an adequate isothermal model this parameter has led us to know the plasma electron temperature. Spectra and emission times of various rays from the aluminium and selenium plasmas have been computed satisfactorily with the Averroes/Transpec code coupled with Film and Multif hydrodynamical codes. (A.C.)

Computer code validation by high temperature chemistry

International Nuclear Information System (INIS)

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

1988-01-01

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

ASTEC V2 severe accident integral code: Fission product modelling and validation

International Nuclear Information System (INIS)

Cantrel, L.; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

2014-01-01

One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-15

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

The RAGE radiation-hydrodynamic code

Energy Technology Data Exchange (ETDEWEB)

Gittings, Michael; Clover, Michael; Betlach, Thomas; Byrne, Nelson; Ranta, Dale [Science Applications International Corp. MS A-1, 10260 Campus Point Drive, San Diego, CA 92121 (United States); Weaver, Robert; Coker, Robert; Dendy, Edward; Hueckstaedt, Robert; New, Kim; Oakes, W Rob [Los Alamos National Laboratory, MS T087, PO Box 1663, Los Alamos, NM 87545 (United States); Stefan, Ryan [TaylorMade-adidas Golf, 5545 Fermi Court, Carlsbad, CA 92008-7324 (United States)], E-mail: michael.r.clover@saic.com

2008-10-01

We describe RAGE, the 'radiation adaptive grid Eulerian' radiation-hydrodynamics code, including its data structures, its parallelization strategy and performance, its hydrodynamic algorithm(s), its (gray) radiation diffusion algorithm, and some of the considerable amount of verification and validation efforts. The hydrodynamics is a basic Godunov solver, to which we have made significant improvements to increase the advection algorithm's robustness and to converge stiffnesses in the equation of state. Similarly, the radiation transport is a basic gray diffusion, but our treatment of the radiation-material coupling, wherein we converge nonlinearities in a novel manner to allow larger timesteps and more robust behavior, can be applied to any multi-group transport algorithm.

Proposal to consistently apply the International Code of Nomenclature of Prokaryotes (ICNP) to names of the oxygenic photosynthetic bacteria (cyanobacteria), including those validly published under the International Code of Botanical Nomenclature (ICBN)/International Code of Nomenclature for algae, fungi and plants (ICN), and proposal to change Principle 2 of the ICNP.

Science.gov (United States)

Pinevich, Alexander V

2015-03-01

This taxonomic note was motivated by the recent proposal [Oren & Garrity (2014) Int J Syst Evol Microbiol 64, 309-310] to exclude the oxygenic photosynthetic bacteria (cyanobacteria) from the wording of General Consideration 5 of the International Code of Nomenclature of Prokaryotes (ICNP), which entails unilateral coverage of these prokaryotes by the International Code of Nomenclature for algae, fungi, and plants (ICN; formerly the International Code of Botanical Nomenclature, ICBN). On the basis of key viewpoints, approaches and rules in the systematics, taxonomy and nomenclature of prokaryotes it is reciprocally proposed to apply the ICNP to names of cyanobacteria including those validly published under the ICBN/ICN. For this purpose, a change to Principle 2 of the ICNP is proposed to enable validation of cyanobacterial names published under the ICBN/ICN rules. © 2015 IUMS.

The database 'EDUD Base' for validation of neutron-physics codes used to analyze the WWER-440 cores

International Nuclear Information System (INIS)

Rocek, J.; Belac, J.; Miasnikov, A.

2003-01-01

The program and data system EDUDBase for validation of reactor computing codes was developed at NRI. It is designed for validation and evaluation of the precision of different computer codes used for WWER core analyses. The main goal of this database is to provide data for comparison with calculation results of tested codes and tools for statistical analysis or differences between the calculation results and the test data. The benchmark data sets are based on in-core measurements performed on WWER-440 reactors of Dukovany NPP. The initial data from NPP are verified, errors and inaccuracies are eliminated and data are transferred to a form, which is suitable for comparison with results of calculations. A special reduced operating history data set is created for each operating cycle ('Benchmark Operation History') to be used as an input data for calculation. It contains values of some integral quantities for each time point: effective time, integral thermal power, boron concentration, position of working group control assemblies (group 6) and inlet coolant temperature. At present, sets are available for all completed cycles up to: (unit/cycle) 1/17, 2/16, 3/15, 4/15. Power distribution is described for approx. 40 time steps during each operating cycle. 2D-power distributions are transferred into 60-degree core symmetry sector of reactor core. At present, such data sets are available only for later cycles starting with: (unit/cycle) 1/7, 2/6, 3/5, 4/5 (in other words last II cycles for each unit) (Authors)

Momentary assessment of adults’ physical activity and sedentary behavior: Feasibility and validity