TRAC-PF1/MOD1 assessment at Los Alamos

International Nuclear Information System (INIS)

Knight, T.D.

1984-01-01

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in pressurized water reactors (PWRs). Over the past several years, four distinct versions of the code have been released; each new version introduced improvements to the existing models and numerics and added new models to extend the applications of the code. The first goal of the code was to analyze large-break loss-of-coolant accidents (LOCAs), and the TRAC-P1A and TRAC-PD2 codes primarily addressed the large-break LOCA. (The TRAC-PD2/MOD1 code is essentially the same as the TRAC-PD2 code but it also includes a released set of error corrections.) The TRAC-PF1 code contained major changes to the models and trips and to the numerical methods. These modifications enhanced the computational speed of the code and improved the application to small-break LOCAs. The TRAC-PF1/MOD1 code, the latest released version, added improved steam-generator modeling, a turbine component, and a control system together with modified constitutive relations to model the balance of plant on the secondary side and to extend the applications to non-LOCA transients. The TRAC-PF1/MOD1 code also contains reasonably general reactor-kinetics modeling to facilitate the simulation of transients with delayed scram or without scram. 13 references, 24 figures

TRAC methods and models

International Nuclear Information System (INIS)

Mahaffy, J.H.; Liles, D.R.; Bott, T.F.

1981-01-01

The numerical methods and physical models used in the Transient Reactor Analysis Code (TRAC) versions PD2 and PF1 are discussed. Particular emphasis is placed on TRAC-PF1, the version specifically designed to analyze small-break loss-of-coolant accidents

TRAC development and assessment status

International Nuclear Information System (INIS)

Vigil, J.C.; Knight, T.D.

1981-01-01

TRAC is being developed at the Los Alamos National Laboratory to provide an advanced systems code for light-water reactor accident analysis. The released TRAC versions (P1, P1A, and PD2) were intended primarily as benchmark codes for large-break loss-of-coolant accidents but PD2 has been applied successfully to TMI-type transients and other small-break transients. A fast-running version, PFl, is currently under development to address more efficiently and accurately these types of transients. All of the released versions have been subjected to testing against separate-effects, system-effects, and integral experiments covering a wide range of scales. Assessment results indicate that PD2 does a credible job overall; needed improvements are being addressed in PFl and in modifications to PD2

LINK codes TRAC-BF1/PARCSv2.7 in LINUX without external communication interface

International Nuclear Information System (INIS)

Barrachina, T.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Verdu, G.; Concejal, A.; Solar, A.

2014-01-01

The TRAC-BF1 code is still widely used by the nuclear industry for safety analysis. The plant models developed using this code are highly validated, so it is advisable to continue improving this code before migrating to another completely different code. The coupling with the NRC neutronic code PARCSv2.7 increases the simulation capabilities in transients in which the power distribution plays an important role. In this paper, the procedure for the coupling of TRAC-BF1 and PARCSv2.7 codes without PVM and in Linux is presented. (Author)

Assessment of TRAC-BF1 1D reflood model with CCTF and SCTF data

International Nuclear Information System (INIS)

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

1993-03-01

Post test calculations for six selected Cylindrical Core Test Facility (CCTF) and Slab Core Test Facility (SCTF) tests were performed to assess the core thermal hydraulic models of the TRAC-BF1 code during the reflood in a PWR LOCA. A special version of the TRAC code was developed at JAERI by implementing the constitutive package of the TRAC-BF1 code into the TRAC-PF1 code for this assessment. The TRAC-BF1 model predicted well the void fraction at either bottom or top part of the core and overpredicted the void fraction at the center part of the core in the CCTF and SCTF tests performed under so-called licensing conditions. The TRAC-BF1 model overpredicted the clad temperatures at the center part of the core. The TRAC-BF1 model predicted a jump of void fraction where the flow pattern transition between the bubbly/slug flow and the annular/dispersed flow regimes occurred. The jump caused the water mass flow rate to be unstable and resulted in the overprediction of the void fraction at the center part of the core. It was also found that the TRAC-BF1 film boiling model underestimated the heat transfer coefficient in the vicinity of the quench front and caused the quench front propagation to be delayed. These assessment results suggest the following areas should be improved in future to apply the TRAC-BF1 code to the reflood in a PWR LOCA: (1) Core hydraulic model where flow pattern transition occurs, (2) Core heat transfer model in the film boiling regime, especially for the dependence on the distance from the quench front. (author)

Review of solution approach, methods, and recent results of the TRAC-PF1 system code

International Nuclear Information System (INIS)

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

1983-01-01

The current version of the Transient Reactor Analysis Code (TRAC-PF1) was created to improve on the capabilities of its predecessor (TRAC-PD2) for analyzing slow reactor transients such as small-break loss-of-coolant accidents. TRAC-PF1 continues to use a semi-implicit finite-difference method for modeling three-dimensional flows in the reactor vessel. However, it contains a new stability-enhancing two-step (SETS) finite-difference tecnique for one-dimensional flow calculations. This method is not restricted by a material Courant stability condition, allowing much larger time-step sizes during slow transients than would a semi-implicit method. These have been successfully applied to the analysis of a variety of experiments and hypothetical plant transients covering a full range of two-phase flow regimes

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

International Nuclear Information System (INIS)

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

1990-08-01

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

LINK codes TRAC-BF1/PARCSv2.7 in LINUX without external communication interface; Acoplamiento de los codigos TRAC-BF1/PARCSv2.7 en Linux sin interfaz externa de comunicacion

Energy Technology Data Exchange (ETDEWEB)

Barrachina, T.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Verdu, G.; Concejal, A.; Solar, A.

2014-07-01

The TRAC-BF1 code is still widely used by the nuclear industry for safety analysis. The plant models developed using this code are highly validated, so it is advisable to continue improving this code before migrating to another completely different code. The coupling with the NRC neutronic code PARCSv2.7 increases the simulation capabilities in transients in which the power distribution plays an important role. In this paper, the procedure for the coupling of TRAC-BF1 and PARCSv2.7 codes without PVM and in Linux is presented. (Author)

MINI-TRAC code: a driver program for assessment of constitutive equations of two-fluid model

International Nuclear Information System (INIS)

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

1991-05-01

MINI-TRAC code, a driver program for assessment of constitutive equations of two-fluid model, has been developed to perform assessment and improvement of constitutive equations of two-fluid model widely and efficiently. The MINI-TRAC code uses one-dimensional conservation equations for mass, momentum and energy based on the two-fluid model. The code can work on a personal computer because it can be operated with a core memory size less than 640 KB. The MINI-TRAC code includes constitutive equations of TRAC-PF1/MOD1 code, TRAC-BF1 code and RELAP5/MOD2 code. The code is modulated so that one can easily change constitutive equations to perform a test calculation. This report is a manual of the MINI-TRAC code. The basic equations, numerics, constitutive, equations included in the MINI-TRAC code will be described. The user's manual such as input description will be presented. The program structure and contents of main variables will also be mentioned in this report. (author)

Assessment of TRAC-PD2 reflood core thermo-hydraulic model by CCTF Test C1-16

International Nuclear Information System (INIS)

Sugimoto, Jun

1982-11-01

The TRAC-PD2 reflood core thermo-hydraulic model was assessed by CCTF Test C1-16. The measured data were utilized as core boundary conditions in the TRAC calculations. The results indicate that the core inlet liquid temperature and the core heater rod temperatures are in reasonable agreement with data, but the pressure distribution in the core and water pool formation in the upper plenum are not in good agreement. The parametric effects of the droplet critical Weber number, the material properties of the heater rod, the noding of the upper plenum, and the minimum stable film boiling temperature are also discussed. (author)

Development of REFLA/TRAC code for engineering work station

International Nuclear Information System (INIS)

Ohnuki, Akira; Akimoto, Hajime; Murao, Yoshio

1994-03-01

The REFLA/TRAC code is a best-estimate code which is expected to check reactor safety analysis codes for light water reactors (LWRs) and to perform accident analyses for LWRs and also for an advanced LWR. Therefore, a high predictive capability is required and the assessment of each physical model becomes important because the models govern the predictive capability. In the case of the assessment of three-dimensional models in REFLA/TRAC code, a conventional large computer is being used and it is difficult to perform the assessment efficiently because the turnaround time for the calculation and the analysis is long. Then, a REFLA/TRAC code which can run on an engineering work station (EWS) was developed. Calculational speed of the current EWS is the same order as that of large computers and the EWS has an excellent function for multidimensional graphical drawings. Besides, the plotting processors for X-Y drawing and for two-dimensional graphical drawing were developed in order to perform efficient analyses for three-dimensional calculations. In future, we can expect that the assessment of three-dimensional models becomes more efficient by introducing an EWS with higher calculational speed and with improved graphical drawings. In this report, each outline for the following three programs is described: (1) EWS version of REFLA/TRAC code, (2) Plot processor for X-Y drawing and (3) Plot processor for two-dimensional graphical drawing. (author)

Comparative calculations on selected two-phase flow phenomena using major PWR system codes

International Nuclear Information System (INIS)

1990-01-01

In 1988 a comparative study on important features and models in six major best estimate thermal hydraulic codes for PWR systems was implemented (Comparison of thermal hydraulic safety codes for PWR Graham, Trotman, London, EUR 11522). It was a limitation of that study that the source codes themselves were not available but the comparison had to be based on the available documentation. In the present study, the source codes were available and the capability of four system codes to predict complex two-phase flow phenomena has been assessed. Two areas of investigation were selected: (a) pressurized spray phenomena; (b) boil-up phenomena in rod bundles. As regards the first area, experimental data obtained in 1972 on the Neptunus Facility (Delft University of Technology) were compared with the results of the calculations using Athlet, Cathare, Relap 5 and TRAC-PT1 and, concerning the second area, the results of two experimental facilities obtained in 1980 and 1985 on Thetis (UKEA) and Pericles (CEA-Grenoble) were considered

One dimensional neutron kinetics in the TRAC-BF1 code

International Nuclear Information System (INIS)

Weaver, W.L. III; Wagner, K.C.

1987-01-01

The TRAC-BWR code development program at the Idaho National Engineering Laboratory is developing a version of the TRAC code for the U.S. Nuclear Regulatory Commission (USNRC) to provide a best-estimate analysis capability for the simulation of postulated accidents in boiling water reactor (BWR) power systems and related experimental facilities. Recent development efforts in the TRAC-BWR program have focused on improving the computational efficiency through the incorporation of a hybrid Courant- limit-violating numerical solution scheme in the one-dimensional component models and on improving code accuracy through the development of a one-dimensional neutron kinetics model. Many other improvements have been incorporated into TRAC-BWR to improve code portability, accuracy, efficiency, and maintainability. This paper will describe the one- dimensional neutron kinetics model, the generation of the required input data for this model, and present results of the first calculations using the model

Investigation of spatial coupling aspects for coupled code application in PWR safety analysis

International Nuclear Information System (INIS)

Todorova, N.K.; Ivanov, K.N.

2003-01-01

The simulation of nuclear power plant accident conditions requires three-dimensional (3-D) modeling of the reactor core to ensure a realistic description of physical phenomena. This paper describes a part of the research activities carried out on the sensitivity of coupled neutronics/thermal-hydraulic system code's results to the spatial mesh overlays used for modeling pressurized water reactor (PWR) cores for analysis of different transients. The coupled TRAC-PF1/NEM was used to model PWR rod ejection accident (REA). Modeling schemes for pressurized water reactor are described in detail, followed by a comparative analysis of both steady state and transient calculations. By using different TRAC-PF1/NEM vessel modeling options it was demonstrated that the geometric refinement plays a great role in determining the local parameters and control rod worth in the case of spatially asymmetric transients. The capability of TRAC-PF1/NEM to introduce local refinement of heat structure models was explored while preserving the original coarse-mesh structure of the hydraulic model. The obtained results indicated that the thermal-hydraulic feedback phenomenon is non-linear and cannot be separated even in rod ejection accident analysis, where the Doppler feedback plays a dominant role. While the impact of neutronics mesh refinement is well known, this research found that the local predictions, as well as the global predictions are also very sensitive to the thermal-hydraulic refinement

TRAC analysis of the Crystal River Unit-3 Plant transient of February 26, 1980

International Nuclear Information System (INIS)

Coddington, P.; Willcutt, G.J.E. Jr.

1983-01-01

This paper describes the application of the TRAC-PD2 and TRAC-PF1 codes to analyze the Crystal River transient. The PD2 and PF1 analyses used the three-dimensional and one-dimensional vessel models, respectively. Both calculations predicted the plant depressurization caused by the open PORV and the subsequent repressurization caused by closing the PORV and continuing high-pressure injection flow. Also, natural circulation was calculated in loop B following reestablishment of feedwater to the loop-B steam generator. After system repressurization, the codes calculated that pressure was relieved through the safety valves, and an intermittent flow occurred in loop A because of high-pressure-injection-driven density variations

Post-processing of the TRAC code's results

International Nuclear Information System (INIS)

Baron, J.H.; Neuman, D.

1987-01-01

The TRAC code serves for the analysis of accidents in nuclear installations from the thermohydraulic point of view. A program has been developed with the aim of processing the information rapidly generated by the code, with screening graph capacity, both in high and low resolution, or either in paper through printer or plotter. Although the programs are intended to be used after the TRAC runs, they may be also used even when the program is running so as to observe the calculation process. The advantages of employing this type of tool, its actual capacity and its possibilities of expansion according to the user's needs are herein described. (Author)

TRAC-PF1 code verification with data from the OTIS test facility

International Nuclear Information System (INIS)

Childerson, M.T.; Fujits, R.K.

1985-01-01

A computer code (TRAC-PFI/MODI; denoted as TRAC) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the Once-Through Integral Systems (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and saturation, intermittent reactor coolant system circulation, boiler-condenser mode and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool- and auxiliary- feedwater initiated boiler-condenser mode heat transfer

Fluid dynamics and heat transfer methods for the TRAC code

International Nuclear Information System (INIS)

Reed, W.H.; Kirchner, W.L.

1977-01-01

A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, new highly implicit difference techniques are developed that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained

Introduction and immigration of TRAC-PF1 code

International Nuclear Information System (INIS)

Yan Yuhua; Gao Zuying; Gao Cheng; Li Jincai

1997-01-01

TRAC-PF1 code performs best-estimate predictions of postulated accidents for pressurized light water reactors. It is one of the few system analyse codes which use two fluid model to treat two phase problems in nuclear system. In order to use this advanced software in China and make it possible to be run in different compute systems, IBM version of TRAC-PF1 code, imported from USA National Energy Software Center, is immigrated to CDC NOS/VE system and SUN workstation. The differences in computer languages from IBM 370 to CDC NOS/VE and to SUN workstation are modified properly. All the benchmark problems are calculated, and the results show that the immigration is successful

TRAC User's Guide

International Nuclear Information System (INIS)

Boyack, B.E.; Stumpf, H.; Lime, J.F.

1985-11-01

This guide has been prepared to assist users in applying the Transient Reactor Analysis Code (TRAC). TRAC is an advanced best-estimate systems code for analyzing transients in thermal-hydraulic systems. The code is very general. Because it is general, efforts to model specific nuclear power plants or experimental facilities often present a challenge to the TRAC user. This guide has been written to assist first-time or intermediate users. It is specifically written for the TRAC version designated TRAC-PF1/MOD1. The TRAC User's Guide should be considered a companion document to the TRAC Code Manual; the user will need both documents to use TRAC effectively. 18 refs., 45 figs., 19 tabs

Fluid dynamics and heat transfer methods for the TRAC code

International Nuclear Information System (INIS)

Reed, W.H.; Kirchner, W.L.

1977-01-01

A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, we have developed new highly implicit difference techniques that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained. (author)

Simulation of LOFT anticipated-transient experiments L6-1, L6-2, and L6-3 using TRAC-PF1/MOD1

International Nuclear Information System (INIS)

Sahota, M.S.

1984-01-01

Anticipated-transient experiments L6-1, L6-2, and L6-3, performed at the Loss-of-fluid Test (LOFT) facility, are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1/MOD1). The results are used to assess TRAC-PF1/MOD1 trip and control capabilities, and predictions of thermal-hydraulic phenomena during slow transients. Test L6-1 simulated a loss-of-stream load in a large pressurized-water reactor (PWR), and was initiated by closing the main steam-flow control valve (MSFCV) at its maximum rate, which reduced the heat removal from the secondary-coolant system and increased the primary-coolant system pressure that initiated a reactor scram. Test L6-2 simulated a loss-of-primary coolant flow in a large PWR, and was initiated by tripping the power to the primary-coolant pumps (PCPs) allowing the pumps to coast down. The reduced primary-coolant flow caused a reactor scram. Test L6-3 simulated an excessive-load increase incident in a large PWR, and was initiated by opening the MSFCV at its maximum rate, which increased the heat removal from the secondary-coolant system and decreased the primary-coolant system pressure that initiated a reactor scram. The TRAC calculations accurately predict most test events. The test data and the calculated results for most parameters of interest also agree well

TRAC-PF1/MOD2 status and plans

International Nuclear Information System (INIS)

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

1989-01-01

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

TRAC-PF1 code verification with data from the OTIS test facility

International Nuclear Information System (INIS)

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer

Development of a graphical user interface for the TRAC plant/safety analysis code

Energy Technology Data Exchange (ETDEWEB)

Kelly, A.E.; Harkins, C.K.; Smith, R.J.

1995-09-01

A graphical user interface (GUI) for the Transient Reactor Analysis Code (TRAC) has been developed at Knolls Atomic Power Laboratory. This X Window based GUI supports the design and analysis process, acting as a preprocessor, runtime editor, help system and post processor to TRAC-PF1/MOD2. TRAC was developed at the Los Alamos National Laboratory (LANL). The preprocessor is an icon-based interface which allows the user to create a TRAC model. When the model is complete, the runtime editor provides the capability to execute and monitor TRAC runs on the workstation or supercomputer. After runs are made, the output processor allows the user to extract and format data from the TRAC graphics file. The TRAC GUI is currently compatible with TRAC-PF1/MOD2 V5.3 and is available with documentation from George Niederauer, Section Leader of the Software Development Section, Group TSA-8, at LANL. Users may become functional in creating, running, and interpreting results from TRAC without having to know Unix commands and the detailed format of any of the data files. This reduces model development and debug time and increases quality control. Integration with post-processing and visualization tools increases engineering effectiveness.

Development of a graphical user interface for the TRAC plant/safety analysis code

International Nuclear Information System (INIS)

Kelly, A.E.; Harkins, C.K.; Smith, R.J.

1995-01-01

A graphical user interface (GUI) for the Transient Reactor Analysis Code (TRAC) has been developed at Knolls Atomic Power Laboratory. This X Window based GUI supports the design and analysis process, acting as a preprocessor, runtime editor, help system and post processor to TRAC-PF1/MOD2. TRAC was developed at the Los Alamos National Laboratory (LANL). The preprocessor is an icon-based interface which allows the user to create a TRAC model. When the model is complete, the runtime editor provides the capability to execute and monitor TRAC runs on the workstation or supercomputer. After runs are made, the output processor allows the user to extract and format data from the TRAC graphics file. The TRAC GUI is currently compatible with TRAC-PF1/MOD2 V5.3 and is available with documentation from George Niederauer, Section Leader of the Software Development Section, Group TSA-8, at LANL. Users may become functional in creating, running, and interpreting results from TRAC without having to know Unix commands and the detailed format of any of the data files. This reduces model development and debug time and increases quality control. Integration with post-processing and visualization tools increases engineering effectiveness

Nodal kinetics model upgrade in the Penn State coupled TRAC/NEM codes

International Nuclear Information System (INIS)

Beam, Tara M.; Ivanov, Kostadin N.; Baratta, Anthony J.; Finnemann, Herbert

1999-01-01

The Pennsylvania State University currently maintains and does development and verification work for its own versions of the coupled three-dimensional kinetics/thermal-hydraulics codes TRAC-PF1/NEM and TRAC-BF1/NEM. The subject of this paper is nodal model enhancements in the above mentioned codes. Because of the numerous validation studies that have been performed on almost every aspect of these codes, this upgrade is done without a major code rewrite. The upgrade consists of four steps. The first two steps are designed to improve the accuracy of the kinetics model, based on the nodal expansion method. The polynomial expansion solution of 1D transverse integrated diffusion equation is replaced with a solution, which uses a semi-analytic expansion. Further the standard parabolic polynomial representation of the transverse leakage in the above 1D equations is replaced with an improved approximation. The last two steps of the upgrade address the code efficiency by improving the solution of the time-dependent NEM equations and implementing a multi-grid solver. These four improvements are implemented into the standalone NEM kinetics code. Verification of this code was accomplished based on the original verification studies. The results show that the new methods improve the accuracy and efficiency of the code. The verification of the upgraded NEM model in the TRAC-PF1/NEM and TRAC-BF1/NEM coupled codes is underway

Analysis on ingress of coolant event in vacuum vessel using modified TRAC-BF1 code

International Nuclear Information System (INIS)

Ajima, Toshio; Kurihara, Ryoichi; Seki, Yasushi

1999-08-01

The Transient Reactor Analysis Code (TRAC-BF1) was modified on the basis of ICE experimental results so as to analyze the Ingress of Coolant Event (ICE) in the vacuum vessel of a nuclear fusion reactor. In the previous report, the TRAC-BF1 code, which was originally developed for the safety analysis of a light water reactor, had been modified for the ICE of the fusion reactor. And the addition of the flat structural plate model to the VESSEL component and arbitrary appointment of the gravity direction had been added in the TRAC-BF1 code. This TRAC-BF1 code was further modified. The flat structural plate model of the VESSEL component was enabled to divide in multi layers having different materials, and a part of the multi layers could take a buried heater into consideration. Moreover, the TRAC-BF1 code was modified to analyze under the low-pressure condition close to vacuum within range of the steam table. This paper describes additional functions of the modified TRAC-BF1 code, analytical evaluation using ICE experimental data and the ITER model with final design report (FDR) data. (author)

Methodology, status, and plans for development and assessment of the TRAC code

International Nuclear Information System (INIS)

Boyack, B.E.; Nelson, R.A.; Jolly-Woodruff, S.

1996-01-01

The Transient Reactor Analysis Code (TRAC) is a state-of-the-art, best-estimate, transient system analysis computer code for analyzing geometrically complex multidimensional thermal hydraulic systems, primarily nuclear reactor power plants. TRAC is used by government and industry organizations for design and safety analysis, phenomenological studies, operational transient analysis, evaluating emergency operating procedures, simulator support and operator training, and for assessment of data involving basic experiments, separate effects tests, and plant operations. TRAC will calculate one- and three-dimensional (rectilinear and cylindrical coordinates) fluid flow involving gas, liquid, and mixture states. Although TRAC has many capabilities, it also has limitations. Some limitations arise from its implementation, dating from the 1970s. Rapid advances in hardware and software engineering highlight TRAC's inefficiencies; however, other limitations relate to the level of scientific knowledge regarding two-phase flow physics. These limitations will continue until such time as the fundamental understanding of two-phase flows is extended. Presently, several development activities are either in progress or soon to begin that will fundamentally improve TRAC. Foremost among these are reimplementation of the current TRAC data structures in Fortran 90 and the integrated development of closure packages for large-break loss-of-coolant accident applications

TRAC-BD1: transient reactor analysis code for boiling-water systems

International Nuclear Information System (INIS)

Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

1981-01-01

The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented

TRAC analyses for CCTF and SCTF tests and UPTF design/operation

International Nuclear Information System (INIS)

Williams, K.A.

1983-01-01

The 2D/3D Program is a multinational (Germany, Japan, and the United States) experimental and analytical nuclear reactor safety research program. The Los Alamos analysis effort is functioning as a vital part of the 2D/3D program. The CCTF and SCTF analyses have demonstrated that TRAC-PF1 can correctly predict multidimensional, nonequilibrium behavior in large-scale facilities prototypical of actual PWR's. Through these and future TRAC analyses the experimental findings can be related from facility to facility, and the results of this research program can be directly related to licensing concerns affecting actual PWR's

TRAC code assessment using data from SCTF Core-III, a large-scale 2D/3D facility

International Nuclear Information System (INIS)

Boyack, B.E.; Shire, P.R.; Harmony, S.C.; Rhee, G.

1988-01-01

Nine tests from the SCTF Core-III configuration have been analyzed using TRAC-PF1/MOD1. The objectives of these assessment activities were to obtain a better understanding of the phenomena occurring during the refill and reflood phases of a large-break loss-of-coolant accident, to determine the accuracy to which key parameters are calculated, and to identify deficiencies in key code correlations and models that provide closure for the differential equations defining thermal-hydraulic phenomena in pressurized water reactors. Overall, the agreement between calculated and measured values of peak cladding temperature is reasonable. In addition, TRAC adequately predicts many of the trends observed in both the integral effect and separate effect tests conducted in SCTF Core-III. The importance of assessment activities that consider potential contributors to discrepancies between the measured and calculated results arising from three sources are described as those related to (1) knowledge about the facility configuration and operation, (2) facility modeling for code input, and (3) deficiencies in code correlations and models. An example is provided. 8 refs., 7 figs., 2 tabs

TRAC-PF1 code assessment using OECD LOFT LP-FP-1 experiment

International Nuclear Information System (INIS)

Barbero, F.J.

1992-04-01

This report assesses thermal-hydraulic aspects of LOFT LP-FP-1 experiment making use of TRAC-PF1/MOD1. LP-FP-1 experiment studies the system thermal-hydraulic and core thermal response for initial and boundary conditions similar to a large-break design basis LOCA leading to fission product release from the fuel cladding gap region. It also assesses the fission product retention effectiveness of the PWR-ECCS in best estimate conditions

Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C. [LOCA Integrated Services I, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

Tests of the TRAC code against known analytical solutions for stratified flow

International Nuclear Information System (INIS)

Black, P.S.; Leslie, D.C.; Hewitt, G.F.

1987-01-01

The area averaged equations for gas-liquid flow are briefly summarized and related, for the specific case of stratified flow, to the shallow water equations commonly used in hydraulics. These equations are then compared to the equations used in TRAC-PF/MOD1 and are shown to differ in their treatment of the gravity head terms. A modification of the TRAC code is therefore necessary to bring it into line with established shallow water theory. The corrected form of the code was compared with a number of specific cases, each of which throws further light on the code behavior. The following areas are discussed in the paper: (1) the dam break problem; (2) Kelvin-Helmholtz instability; (3) counter-current flow; and (4) slug flow. It is concluded that detailed comparisons of the code with known analytic solutions and with a number of the more complex phenomenological experiments can give useful insights into its behavior

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

International Nuclear Information System (INIS)

Bubelis, Evaldas; Coddington, Paul; Leung, Waihung

2006-01-01

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

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

International Nuclear Information System (INIS)

Schultz, R.R.

1993-12-01

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

Study plan for the sensitivity analysis of the Terrain-Responsive Atmospheric Code (TRAC)

International Nuclear Information System (INIS)

Restrepo, L.F.; Deitesfeld, C.A.

1987-01-01

Rocky Flats Plant, Golden, Colorado is presently developing a computer code to model the dispersion of potential or actual releases of radioactive or toxic materials to the environment, along with the public consequences from these releases. The model, the Terrain-Responsive Atmospheric Code (TRAC), considers several complex features which could affect the overall dispersion and consequences. To help validate TRAC, a sensitivity analysis is being planned to determine how sensitive the model's solutions are to input variables. This report contains a brief description of the code, along with a list of tasks and resources needed to complete the sensitivity analysis

Development of Transient-Reactor Analysis Code (TRAC) for real-time applications

International Nuclear Information System (INIS)

Niederauer, G.F.; Giguere, P.T.; Lime, J.F.; Knight, T.D.; Ashy, O.; Fakory, R.

1997-01-01

This is the final report of a six-month, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). Nuclear-plant training simulators employ simplified one-dimensional thermal-hydraulics codes because of the demands to run in real time and with limited computing power. The objective of this project was to investigate the feasibility of using the advanced Transient-Reactor Analysis Code (TRAC) in a simulator to increase the fidelity of a simulator. Many issues need to be addressed to take such a complex code from a batch engineering environment to a real-time environment. Working with simulator vendor, GSE, the authors investigated the technical issues relating to integrating TRAC into a real-time environment. They also modified a nuclear power plant model for simulator purposes and investigated its performance in real time

Independent assessment of TRAC and RELAP5 codes through separate effects tests

International Nuclear Information System (INIS)

Saha, P.; Rohatgi, U.S.; Jo, J.H.; Neymotin, L.; Slovik, G.; Yuelys-Miksis, C.; Pu, J.

1983-01-01

Independent assessment of TRAC-PF1 (Version 7.0), TRAC-BD1 (Version 12.0) and RELAP5/MOD1 (Cycle 14) that was initiated at BNL in FY 1982, has been completed in FY 1983. As in the previous years, emphasis at Brookhaven has been in simulating various separate-effects tests with these advanced codes and identifying the areas where further thermal-hydraulic modeling improvements are needed. The following six catetories of tests were simulated with the above codes: (1) critical flow tests (Moby-Dick nitrogen-water, BNL flashing flow, Marviken Test 24); (2) Counter-Current Flow Limiting (CCFL) tests (University of Houston, Dartmouth College single and parallel tube test); (3) level swell tests (G.E. large vessel test); (4) steam generator tests (B and W 19-tube model S.G. tests, FLECHT-SEASET U-tube S.G. tests); (5) natural circulation tests (FRIGG loop tests); and (6) post-CHF tests (Oak Ridge steady-state test)

Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

International Nuclear Information System (INIS)

Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

2004-01-01

An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect

Implantation of TRAC-PF1 computer code of VAX-11/750

International Nuclear Information System (INIS)

Madeira, A.A.; Souza Gouvea, L. de; Galetti, M.R.S.

1988-01-01

The implantation of TRAC-PF1 code, IBM version, in VAX-11/750 is described. This work provides Reator Department with an advanced best-estimate tool to perform loss-of-coolant accident analysis. (Author) [pt

TRAC-P validation test matrix. Revision 1.0

International Nuclear Information System (INIS)

Hughes, E.D.; Boyack, B.E.

1997-01-01

This document briefly describes the elements of the Nuclear Regulatory Commission's (NRC's) software quality assurance program leading to software (code) qualification and identifies a test matrix for qualifying Transient Reactor Analysis Code (TRAC)-Pressurized Water Reactor Version (-P), or TRAC-P, to the NRC's software quality assurance requirements. Code qualification is the outcome of several software life-cycle activities, specifically, (1) Requirements Definition, (2) Design, (3) Implementation, and (4) Qualification Testing. The major objective of this document is to define the TRAC-P Qualification Testing effort

PREP-PWR-1.0: a WIMS-D/4 pre-processor code for the generation of data for PWR fuel assemblies

International Nuclear Information System (INIS)

Ball, G.

1991-06-01

The PREP-PWR-1.0 computer code is a substantially modified version of the PREWIM code which formed part of the original MARIA System (Report J.E.N. 543). PREP-PWR-1.0 is a comprehensive pre-processor code which generates input data for the WIMS-D/4.1 code (Report PEL 294) for PWR fuel assemblies, with or without control and burnable poison rods. This data is generated at various base and off-base conditions. The overall cross section generation methodology is described, followed by a brief overview of the model. Aspects of the base/off-base calculational scheme are outlined. Additional features of the code are described while the input data format of PREP-PWR-1.0 is listed. The sample problems and suggestions for further improvements to the code are also described. 2 figs., 2 tabs., 12 refs

Four-fluid model of PWR degraded cores

International Nuclear Information System (INIS)

Dearing, J.F.

1985-01-01

This paper describes the new two-dimensional, four-fluid fluid dynamics and heat transfer (FLUIDS) module of the MELPROG code. MELPROG is designed to give an integrated, mechanistic treatment of pressurized water reactor (PWR) core meltdown accidents from accident initiation to vessel melt-through. The code has a modular data storage and transfer structure, with each module providing the others with boundary conditions at each computational time step. Thus the FLUIDS module receives mass and energy source terms from the fuel pin module, the structures module, and the debris bed module, and radiation energy source terms from the radiation module. MELPROG, which models the reactor vessel, is also designed to model the vessel as a component in the TRAC/PF1 networking solution of a PWR reactor coolant system (RCS). The coupling between TRAC and MELPROG is implicit in the fluid dynamics of the reactor coolant (liquid water and steam) allowing an accurate simulation of the coupling between the vessel and the rest of the RCS during an accident. This paper deals specifically with the numerical model of fluid dynamics and heat transfer within the reactor vessel, which allows a much more realistic simulation (with less restrictive assumptions on physical behavior) of the accident than has been possible before

TRAC analyses of severe overcooling transients for the Oconee-1 PWR

Energy Technology Data Exchange (ETDEWEB)

Ireland, J R [comp.

1985-05-01

This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs.

TRAC analyses of severe overcooling transients for the Oconee-1 PWR

International Nuclear Information System (INIS)

Ireland, J.R.

1985-05-01

This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs

TRAC-P validation test matrix. Revision 1.0

Energy Technology Data Exchange (ETDEWEB)

Hughes, E.D.; Boyack, B.E.

1997-09-05

This document briefly describes the elements of the Nuclear Regulatory Commission`s (NRC`s) software quality assurance program leading to software (code) qualification and identifies a test matrix for qualifying Transient Reactor Analysis Code (TRAC)-Pressurized Water Reactor Version (-P), or TRAC-P, to the NRC`s software quality assurance requirements. Code qualification is the outcome of several software life-cycle activities, specifically, (1) Requirements Definition, (2) Design, (3) Implementation, and (4) Qualification Testing. The major objective of this document is to define the TRAC-P Qualification Testing effort.

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

International Nuclear Information System (INIS)

Newbon, S.

1982-05-01

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

Independent assessment of the TRAC-BD1/MOD1 computer code at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Wilson, G.E.; Charboneau, B.L.; Dallman, R.J.; Kullberg, C.M.; Wagner, K.C.; Wheatley, P.D.

1984-01-01

Under auspices of the United States Nuclear Regulatory Commission, their primary boiling water reactor safety analysis code (TRAC-BWR) is being assessed with simulations of a wide range of experimental data. The FY-1984 assessment activities were associated with the latest version (TRAC-BD1/MOD1) of this code. Typical results of the assessment studies are given. Conclusions formulated from these results are presented. These calculations relate to the overall applicability of the current code to safety analysis, and to future work which would further enhance the code's quality and ease of use

Prediction of reflood behavior for tests with differing axial power shapes using WCOBRA/TRAC

International Nuclear Information System (INIS)

Bajorek, S.M.; Hochreiter, L.E.

1991-01-01

The rector core power shape can vary over the fuel cycle due to load follow, control rod movement, burnup effects and Xenon transients. a best estimate thermal-hydraulic code must be able to accurately predict the reflooding behavior for different axial power shapes in order to find the power shapes effects on the loss-of-coolant peak cladding temperature. Several different reflood heat transfer experiments have been performed at the same or similar PWR reflood conditions with different axial power shapes. These experiments have different rod diameters, were full length, 3.65 m (12 feet) in height, and had simple egg crate grids. The WCOBRA/TRAC code has been used to model several different tests from these three experiments to examine the code's capability to predict the reflood transient for different power shapes, with a consistent model and noding scheme. This paper describes these different experiments, their power shapes, and the test conditions. The WCOBRA/TRAC code is described as well as the noding scheme, and the calculated results will be compared in detail with the test data rod temperatures. An overall assessment of the code's predictions of these experiments is presented

TRAC-BWR development

International Nuclear Information System (INIS)

Weaver, W.L.; Rouhani, S.Z.

1983-01-01

The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

The verification of PWR-fuel code for PWR in-core fuel management

International Nuclear Information System (INIS)

Surian Pinem; Tagor M Sembiring; Tukiran

2015-01-01

In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Comparisons of TRAC-PF-1 calculations with semiscale Mod-3 small-break tests S-SB-P1 and S-SB-P7

International Nuclear Information System (INIS)

Sahota, M.S.

1982-01-01

Semiscale Tests S-SB-P1 and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Tests S-SB-P1 and S-SB-P7 simulated an equivalent pressurized-water-reactor (PWR) 2.5% communicative cold-leg break for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics

Comparison of TRAC calculations with experimental data

International Nuclear Information System (INIS)

Jackson, J.F.; Vigil, J.C.

1980-01-01

TRAC is an advanced best-estimate computer code for analyzing postulated accidents in light water reactors. This paper gives a brief description of the code followed by comparisons of TRAC calculations with data from a variety of separate-effects, system-effects, and integral experiments. Based on these comparisons, the capabilities and limitations of the early versions of TRAC are evaluated

Comparison of SCDAP/RELAP5/MOD3 to TRAC-PF1/MOD1 for timing analysis of PWR fuel pin failures

International Nuclear Information System (INIS)

Jones, K.R.; Katsma, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.

1991-01-01

A comparison has been made of SCDAP/RELAP5/MOD3- and TRAC-PF1/MOD1- based calculations of the fuel pin failure timing (time from containment isolation signal to first fuel pin failure) in a loss-of-coolant accident (LOCA). The two codes were used to calculate the thermal-hydraulic boundary conditions for a complete, double-ended, offset-shear break of a cold leg in a Westinghouse 4-loop pressurized water reactor. Both calculations used the FRAPCON-2 code to calculate the steady-state fuel rod behavior and the FRAP-T6 code to calculate the transient fuel rod behavior. The analysis was performed for 16 combinations of fuel burnups and power peaking factors extending up to the Technical Specifications limits. While all calculations were made on a best-estimate basis, the SCDAP/RELAP5/MOD3 code has not yet been fully assessed for large-break LOCA analysis. The results indicate that SCDAP/RELAP5/MOD3 yields conservative fuel pin failure timing results in comparison to those generated using TRAC-PF1/MOD1. 7 refs., 5 figs

Ballooning analysis for the Sizewell B PWR using symmetric MABEL calculations

International Nuclear Information System (INIS)

Sweet, D.W.; Gibson, I.H.; Fell, J.

1982-12-01

An analysis of the fuel clad ballooning potential associated with the Sizewell B PWR following a design basis large break cold leg LOCA is described. Calculations employ MABEL-2C code. No allowance has been made for asymmetries in power or geometry, thus precluding any amelioration offered by early clad rupture. Thermal hydraulic data were derived from a TRAC-PD2 best estimate analysis of the LOCA and the work includes a detailed sensitivity study which leads to a correlation between peak clad temperature and clad strain. For the best estimate start of cycle 1 peak rod rating, no loss of coolability is expected within 95 percent confidence limits on peak clad temperature. No loss of coolability is expected either for rods at the design basis peak rod rating. The temperature does not have to be much higher than the 95 percent confidence limit on the best estimate rating or much beyond that of the design basis rating for rod contact and severe blockage to follow. This indicates that to establish a complete safety case the added complexity of pellet eccentricity and rod to rod power variations must be considered. (U.K.)

Application programming interface document for the modernized Transient Reactor Analysis Code (TRAC-M)

International Nuclear Information System (INIS)

Mahaffy, J.; Boyack, B.E.; Steinke, R.G.

1998-05-01

The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization

The PWR spectral code GELS. Pt. 1

International Nuclear Information System (INIS)

Penndorf, K.; Schult, F.; Schulz, G.

1976-01-01

The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

TRAC development at General Electric

International Nuclear Information System (INIS)

Andersen, J.G.M.; Shaug, J.C.; Shiralkar, B.S.

1987-01-01

TRAC is a computer code for transient analysis of light water reactors. The BWR version of TRAC has been developed as a result of a close cooperation between General Electric Company and Idaho National Engineering Laboratory. Up through 1985 the development work at General Electric was jointly funded by General Electric, the Nuclear Regulatory Commission and Electric Power Research Institute under the Refill-Reflood and FIST programs. At INEL (which has the main responsibility for the NRC version of TRAC-BWR) this work has led to the development of TRACBD1 and TRACBF1, while at GE, TRACB04 was the final product of the Refill-Reflood and FIST programs. TRAC development has continued at General Electric after the completion of these programs with the evolution of the TRACG code. The purpose of the paper is to describe this work. The TRAC development at General Electric can be divided into two main categories: extended benchmark capability and improved user convenience

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

International Nuclear Information System (INIS)

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations

TRAC-PF1 choked-flow model

International Nuclear Information System (INIS)

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

1983-01-01

The two-phase, two-component choked-flow model implemented in the latest version of the Transient Reactor analysis Code (TRAC-PF1) was developed from first principles using the characteristic analysis approach. The subcooled choked-flow model in TRAC-PF1 is a modified form of the Burnell model. This paper discusses these choked-flow models and their implementation in TRAC-PF1. comparisons using the TRAC-PF1 choked-flow models are made with the Burnell model for subcooled flow and with the homogeneous-equilibrium model (HEM) for two-phae flow. These comparisons agree well under homogeneous conditions. Generally good agreements have been obtained between the TRAC-PF1 results from models using the choking criteria and those using a fine mesh (natural choking). Code-data comparisons between the separate-effects tests of the Marviken facility and the Edwards' blowdown experiment also are favorable. 10 figures

An economic analysis code used for PWR fuel cycle

International Nuclear Information System (INIS)

Liu Dingqin

1989-01-01

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

TRAC posttest calculations of Semiscale Test S-06-3

International Nuclear Information System (INIS)

Ireland, J.R.; Bleiweis, P.B.

1980-01-01

A comparison of Transient Reactor Analysis Code (TRAC) steady-state and transient results with Semiscale Test S-06-3 (US Standard Problem 8) experimental data is discussed. The TRAC model used employs fewer mesh cells than normal data comparison models so that TRAC's ability to obtain reasonable results with less computer time can be assessed. In general, the TRAC results are in good agreement with the data and the major phenomena found in the experiment are reproduced by the code with a substantial reduction in computing times

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Calculations of Edwards' pipe blowdown tests using the code TRAC P1

International Nuclear Information System (INIS)

O'Mahoney, R.

1979-05-01

The paper describes the results obtained using the non-thermal equilibrium LOCA code TRAC-P1 for two of a series of Pipe Blowdown Tests. Comparisons are made with the experimental values and RELAP-UK Mark IV predictions. Some discrepancies between prediction and experiment are observed, and certain aspects of the model are considered to warrant possible further attention. (U.K.)

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Improved gap conductance model for the TRAC code

International Nuclear Information System (INIS)

Hatch, S.W.; Mandell, D.A.

1980-01-01

The purpose of the present work, as indicated earlier, is to improve the present constant fuel clad spacing in TRAC-P1A without significantly increasing the computer costs. It is realized that the simple model proposed may not be accurate enough for some cases, but for the initial calculations made the DELTAR model improves the predictions over the constant Δr results of TRAC-P1A and the additional computing costs are negligible

Zebra: An advanced PWR lattice code

Energy Technology Data Exchange (ETDEWEB)

Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

2012-07-01

This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

Zebra: An advanced PWR lattice code

International Nuclear Information System (INIS)

Cao, L.; Wu, H.; Zheng, Y.

2012-01-01

This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

Energy Technology Data Exchange (ETDEWEB)

Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States). Dept. of Nuclear Engineering; Wang, W. [SCIENTECH, Inc., Rockville, MD (United States); Mousseau, V.A.; Ebert, D.D. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

1999-03-01

A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model.

A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

International Nuclear Information System (INIS)

Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J.; Mousseau, V.A.; Ebert, D.D.

1999-01-01

A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model

Assessment of TRAC-PF1/MOD1 against a loss-of-grid transient in Ringhals 4 power plant

International Nuclear Information System (INIS)

Sjoberg, A.; Almberger, J.; Sandervag, O.

1989-07-01

A loss of grid transient in a three loop Westinghouse PWR has been simulated with the frozen version of TRAC-PF1/MOD1 computer code. The results reveal the capability of the code to qualitatively predict the different pertinent phenomena and the data comparison was quite encouraging. Accurate predictions of the system response required careful determination of the boundary conditions simulating the turbine governor valves and steam dump valves behaviour. An explicit modeling of the steam generator internals was also found to be important for the results. It was also revealed that the pressurizer system including spray and heaters and their operation should be modeled in some detail for proper response. 4 refs., 18 figs

Software design implementation document for TRAC-M data structures

Energy Technology Data Exchange (ETDEWEB)

Jolly-Woodruff, S. [Ogden Environmental and Energy Services (United States); Mahaffy, J. [Pennsylvania State Univ., University Park, PA (United States); Giguere, P.; Dearing, J.; Boyack, B. [Los Alamos National Lab., NM (United States)

1997-07-01

The Transient Reactor Analysis Code (TRAC)-M system-wide and component data structures are to be reimplemented by using the new features of Fortran 90 (F90). There will be no changes to the conceptual design, data flow, or computational flow with respect to the current TRAC-P, except that readability, maintainability, and extensibility will be improved. However, the task described here is a basic step that does not meet all future needs of the code, especially regarding extensibility. TRAC-M will be fully functional and will produce null computational changes with respect to TRAC-P, Version 5.4.25; computational efficiency will not be degraded significantly. The existing component and functional modularity and possibilities for coarse-grained parallelism will be retained.

Software design implementation document for TRAC-M data structures

International Nuclear Information System (INIS)

Jolly-Woodruff, S.; Mahaffy, J.; Giguere, P.; Dearing, J.; Boyack, B.

1997-07-01

The Transient Reactor Analysis Code (TRAC)-M system-wide and component data structures are to be reimplemented by using the new features of Fortran 90 (F90). There will be no changes to the conceptual design, data flow, or computational flow with respect to the current TRAC-P, except that readability, maintainability, and extensibility will be improved. However, the task described here is a basic step that does not meet all future needs of the code, especially regarding extensibility. TRAC-M will be fully functional and will produce null computational changes with respect to TRAC-P, Version 5.4.25; computational efficiency will not be degraded significantly. The existing component and functional modularity and possibilities for coarse-grained parallelism will be retained

TRAC-CFD code integration and its application to containment analysis

International Nuclear Information System (INIS)

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

2004-01-01

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

TRAC Searchable Research Library

Science.gov (United States)

2016-05-01

Relational Data Modeling (VRDM) computational paradigm. VRDM has the key attributes of being cloud available, using domain semantics for configured...Figure 1. Methodology for TRAC Searchable Research Library Development. ........................... 5 Figure 2. The conceptual model for the cloud ...TRAC Searchable Research Library project was initiated by TRAC- HQ to address a current capability gap in the TRAC organization. Currently TRAC does not

Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRAC{sub R}T

Energy Technology Data Exchange (ETDEWEB)

Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto, E-mail: prey@tecnatom.e, E-mail: jaruiz@tecnatom.e, E-mail: nrivero@tecnatom.e [Tecnatom S.A., Madrid (Spain)

2011-07-01

In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC{sub R}T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC{sub R}T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC{sub R}T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

Turbine trip transient analysis in peach bottom NPP with TRAC-BF1 code and Simtab-1D methodology

International Nuclear Information System (INIS)

Barrachina, T.; Miro, R.; Verdu, G.; Collazo, I.; Gonzalez, P.; Concejal, A.; Ortego, P.; Melara, J.

2010-01-01

In TRAC-BF1 nuclear cross-sections are specified in the input deck in as a polynomial expansion. Therefore, it is necessary to obtain the coefficients of this polynomial function. One of the methods proposed in the literature is the KINPAR methodology. This methodology uses the results from different perturbations of the original state to obtain the coefficients of the polynominal expansion. The simulations are performed using the SIMULATE3 code. In this work, a new methodology to obtain the cross-sections set in 1D is presented. The first step consists of the application of the SIMTAB methodology, developed in UPV, to obtain the 3D cross-sections sets from CASMO4/SIMULATE3. These 3D cross-sections sets are collapsed to 1D, using as a weighting factor the 3D thermal and rapid neutron fluxes obtained from SIMULATE3. The 1D cross-sections obtained are in the same format as the 3D sets, hence, it has been necessary to modify the TRAC-BF1 code in order to be able to read and interpolate between these tabulated 1D cross-sections. With this new methodology it is not necessary to perform simulations of different perturbations of the original state, and also the variation range of the moderator density can be higher than using the former KINPAR methodology. This is important for simulating severe accidents in which the variables vary in a wide range. This new methodology is applied to the simulation of the turbine trip transient Benchmark in Peach Bottom NPP using the TRAC-BF1 code. The results of the transient simulation in TRAC-BF1 using the KINPAR methodology and the new methodology, SIMTAB-1D, are compared. (author)

Application of CATE 2.0 code on evaluating activated corrosion products in a PWR cooling loop

Energy Technology Data Exchange (ETDEWEB)

Zhang, Jingyu; Li, Lu; Chen, Yixue [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering

2017-03-15

In PWR plants, most Occupational Radiation Exposure (ORE) for personnel results from Activated Corrosion Products (ACPs) in the cooling loop. In order to evaluate the ACPs in the cooling loop, a three-region transport model is built up based on the theory of driving force from the concentration difference in CATE 2.0 code. In order to analyze the nuclide composition of ACPs, the EAF-2007 nuclear database is embedded in CATE 2.0. The case of MIT PCCL test loop is simulated to test the availability of CATE 2.0 on PWR ACPs evaluation, and the activity of Co-58 and Co-60 after operation for 42 days calculated by CATE 2.0 is consistent with that from the code CRUDSIM adopted by MIT. Then, the nuclide composition of ACPs is analyzed in detail respectively for operation of 42 days and 12 months using CATE 2.0. The results show that the short-lived nuclides contribute a majority of the activity in the regions of in-flux wall and coolant, while the long-lived nuclides contribute most of the activity in the region of out-flux wall.

Horizontal stratified flow model for the 1-D module of WCOBRA/TRAC-TF2: modeling and validation

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Frepoli, C.; Ohkawa, K., E-mail: liaoj@westinghouse.com [Westinghouse Electric Company LLC, LOCA Integrated Services I, Cranberry Twp, Pennsylvania (United States)

2011-07-01

For a two-phase flow in a horizontal pipe, the individual phases may separate by gravity. This horizontal stratification significantly impacts the interfacial drag, interfacial heat transfer and wall drag of the two phase flow. For a PWR small break LOCA, the horizontal stratification in cold legs is a highly important phenomenon during loop seal clearance, boiloff and recovery periods. The low interfacial drag in the stratified flow directly controls the time period for the loop clearance and the level of residual water in the loop seal. Horizontal stratification in hot legs also impacts the natural circulation stage of a small break LOCA. In addition, the offtake phenomenon and cold leg condensation phenomenon are also affected by the occurrence of horizontal stratification in the cold legs. In the 1-D module of the WCOBRA/TRAC-TF2 computer code, a horizontal stratification criterion was developed by combining the Taitel-Dukler model and the Wallis-Dobson model, which approximates the viscous Kelvin-Helmholtz neutral stability boundary. The objective of this paper is to present the horizontal stratification model implemented in the code and its assessment against relevant data. The adequacy of the horizontal stratification transition criterion is confirmed by examining the code-predicted flow regime in a horizontal pipe with the measured data in the flow regime map. The void fractions (or liquid level) for the horizontal stratified flow in cold leg or hot leg are predicted with a reasonable accuracy. (author)

Assessment of flooding in a best estimate thermal hydraulic code (WCOBRA/TRAC)

International Nuclear Information System (INIS)

Takeuchi, K.; Young, M.Y.

1998-01-01

The performance of WCOBRA/TRAC code in predicting the flooding, the counter-current flow limit, is evaluated in three geometries important to nuclear reactor loss-of-coolant accident evaluation; a vertical pipe, a perforated plate, and a downcomer annulus. These flow limits are computationally evaluated through transient conditions. The flooding in the vertical pipe is compared with the classical Wallis flooding limit. The flooding on the perforated plate is compared with the Northwestern flooding data correlation. The downcomer flooding in 1/15th and 1/5th scale model is compared with the Creare data. Finally, full scale downcomer flooding is compared with the UPTF test data. The prediction capability of the code for the flooding is found to be very good. (orig.)

TRAC analysis of steam-generator overfill transients for TMI-1

International Nuclear Information System (INIS)

Bassett, B.

1983-01-01

A reactor safety issue concerning the overfilling of once-through steam generators leading to combined primary/secondary blowdown has been raised recently. A series of six calculations, performed with the LWR best-estimate code, TRAC-PD2, on a Babcock and Wilcox Plant (TMI-1), was performed to investigate this safety issue. The base calculation assumed runaway main feedwater to one steam generator causing it to overfill and to break the main steam line. Four additional calculations build onto the base case with combinations of a pump-seal failure, a steam-generator tube rupture, and the pilot-operated relief valve not reseating. A sixth calculation involved only the rupture of a single steam-generator tube. The results of these analyses indicate that for the transients investigated, the emergency cooling system provided an adequate make-up coolant flow to mitigate the accidents

Improved timestep-size diagnostic edits for TRAC-P

International Nuclear Information System (INIS)

Giguere, P.T.

1996-04-01

Improvements have been made to the timestep-size selection logic diagnostic edits of the Transient Reactor Analysis Code (TRAC), specifically to the TRAC-P version. These include both a precise account of the reason for the selection for individual timesteps and thermal-hydraulic information on mesh cells that control the timestep size. The new edits can be specified by user input as a range of timestep numbers, problem time, or both. A description of the current timestep controls in effect in TRAC-P is also given

Improved timestep-size diagnostic edits for TRAC-P

Energy Technology Data Exchange (ETDEWEB)

Giguere, P.T. [Los Alamos National Lab., NM (United States). Technology and Safety Assessment Div.

1996-04-01

Improvements have been made to the timestep-size selection logic diagnostic edits of the Transient Reactor Analysis Code (TRAC), specifically to the TRAC-P version. These include both a precise account of the reason for the selection for individual timesteps and thermal-hydraulic information on mesh cells that control the timestep size. The new edits can be specified by user input as a range of timestep numbers, problem time, or both. A description of the current timestep controls in effect in TRAC-P is also given.

TRAC-PF1 analyses of potential pressurized-thermal-shock transients at a Combustion-Engineering PWR

International Nuclear Information System (INIS)

Koenig, J.E.; Spriggs, G.D.; Smith, R.C.

1984-01-01

Los Alamos is participating in a program to assess the risk of pressurized thermal shock (PTS) to a reactor vessel. Our role is to provide best-estimate thermal-hydraulic analyses of 12 postulated overcooling transients using TRAC-PF1. These transients are hypothetical and include multiple operator/equipment failures. Calvert Cliffs/Unit-1, a Combustion-Engineering plant, is the pressurized water reactor modeled for this study. The utility and the vendor supplied information for the comprehensive TRAC-PF1 model. Secondary and primary breaks from both hot-zero-power and full-power conditions were simulated for 7200 s (2 h). Low bulk temperatures and loop-flow stagnation while the system was at a high pressure were of particular interest for PTS analysis. Three transients produced primary temperatures below 405 K (270 0 F - the NRC screening criterion) with system repressurization. Six transients indicated flow stagnation would occur in one loop but not both. One transient showed flow stagnation might occur in both loops. Oak Ridge National Laboratory will do fracture-mechanics analysis using these TRAC-PF1 results and make the final determination of the risk of PTS

Integrated TRAC/MELPROG analysis of core damage from a severe feedwater transient in the Oconee-1 PWR

International Nuclear Information System (INIS)

Henninger, R.J.; Boyack, B.E.

1986-01-01

A postulated complete loss-of-feedwater event in the Oconee-1 pressurized water reactor has been analyzed. With an initial version of the lonked TRAC and MELPROG codes, we have modeled the loss-of-feedwater event from initiation to the time of complete disruption of the core, which was calculated to occur by 6800 s. The highest structure temperatures otuside the vessel are on the flow path from the vessel to the pressurizer relief valve. Temperatures in excess of 1200 K could result in failure and depressurization of the primary system before vessel failure

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

TRAC-PF1 analysis of LOFT steam-generator feedwater transient test L9-1

International Nuclear Information System (INIS)

Meier, J.K.

1983-01-01

The Transient Reactor Analysis Code (TRAC-PF1) calculations were compared to test data from Loss-of-Fluid Test (LOFT) L9-1, which was a loss-of-feedwater transient. This paper includes descriptions of the test and the TRAC input and compares the TRAC-calculated results with the test data. We conclude that the code predicted the experiment well, given the uncertainties in the boundary conditions. The analysis indicates the need to model all the flow paths and heat structures, and to improve the TRAC wall condensation heat-transfer model

Application of the MELCOR code to design basis PWR large dry containment analysis.

Energy Technology Data Exchange (ETDEWEB)

Phillips, Jesse; Notafrancesco, Allen (USNRC, Office of Nuclear Regulatory Research, Rockville, MD); Tills, Jack Lee (Jack Tills & Associates, Inc., Sandia Park, NM)

2009-05-01

The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

Computational simulation of natural circulation and rewetting experiments using the TRAC/PF1 code

International Nuclear Information System (INIS)

Silva, J.D. da.

1994-05-01

In this work the TRAC code was used to simulate experiments of natural circulation performed in the first Brazilian integral test facility at (COPESP), Sao Paulo and a rewetting experiment in a single tube test section carried out at CDTN, Belo Horizonte, Brazil. In the first simulation the loop behavior in two transient conditions with different thermal power, namely 20 k W and 120 k W, was verified in the second one the quench front propagation, the liquid mass collected in the carry over measuring tube and the wall temperature at different elevations during the flooding experiment was measured. A comparative analysis, for code consistency, shows a good agreement between the code results and experimental data, except for the quench from velocity. (author). 15 refs, 19 figs, 12 tabs

Assessment of PWR Steam Generator modelling in RELAP5/MOD2

International Nuclear Information System (INIS)

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

1993-06-01

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

Post-test analysis of semiscale large-break test S-06-3 using TRAC-PF1

International Nuclear Information System (INIS)

Boyack, B.E.

1982-01-01

The Transient Reactor Analysis Code (TRAC) is an advanced systems code for light-water-reactor accident analysis. The code was developed originally to analyze large-break loss-of-coolant accidents (LOCAs) and running time was not a primary development criterion. TRAC-PF1 was developed because increased application of the code to long transients such as small-break LOCAs required a faster-running code version. Although developed for long transients, its performance on large-break transients is still important. This paper assesses the ability of TRAC-PF1 to predict large-break-LOCA Test S-06-3 conducted in the Semiscale Mod-1 facility

TRAC calculations of a loss-of-coolant accident in a reactor scale model

International Nuclear Information System (INIS)

Pyun, J.J.

1981-01-01

The TRAC (Transient Reactor Analysis Code) is being developed at the Los Alamos National Laboratory as an advanced best-estimate computer program for analysis of postulated hypothetical accidents in pressurized water reactors. As a part of the TRAC developmental verification efforts, a TRAC posttest analysis of Semiscale Mod-3 Test S-07-6 was conducted. The results of this analysis show that the agreement between TRAC calculations and experimental data is not very good. In particular, TRAC does not predict the long term doncomer and core liquid level oscillations during the reflood phase

Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

International Nuclear Information System (INIS)

Shimooke, Takanori; Yoshida, Kazuo

1979-05-01

The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Presentation of geometries and transient results of TRAC-calculations

International Nuclear Information System (INIS)

Lutz, A.; Lang, U.; Ruehle, R.

1985-02-01

The computer code TRAC is used to analyze the transient behaviour of nuclear reactors. The input of a TRAC-Calculation, as well as the produced result files serve for the graphical presentation of the geometries and transient results. This supports the search for errors during input generation and the understanding of complex processes by dynamic presentation of calculational result in colour. (orig.) [de

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

International Nuclear Information System (INIS)

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

1994-11-01

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

Evaluation of CRUDTRAN code to predict transport of corrosion products and radioactivity in the PWR primary coolant system

International Nuclear Information System (INIS)

Lee, C.B.

2002-01-01

CRUDTRAN code is to predict transport of the corrosion products and their radio-activated nuclides such as cobalt-58 and cobalt-60 in the PWR primary coolant system. In CRUDTRAN code the PWR primary circuit is divided into three principal sections such as the core, the coolant and the steam generator. The main driving force for corrosion product transport in the PWR primary coolant comes from coolant temperature change throughout the system and a subsequent change in corrosion product solubility. As the coolant temperature changes around the PWR primary circuit, saturation status of the corrosion products in the coolant also changes such that under-saturation in steam generator and super-saturation in the core. CRUDTRAN code was evaluated by comparison with the results of the in-reactor loop tests simulating the PWR primary coolant system and PWR plant data. It showed that CRUDTRAN could predict variations of cobalt-58 and cobalt-60 radioactivity with time, plant cycle and coolant chemistry in the PWR plant. (author)

BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

International Nuclear Information System (INIS)

Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

1983-09-01

TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

Resuspension parameters for TRAC dispersion model

International Nuclear Information System (INIS)

Langer, G.

1987-01-01

Resuspension factors for the wind erosion of soil contaminated with plutonium are necessary to run the Rocky Flats Plant Terrain Responsive Atmospheric Code (TRAC). The model predicts the dispersion and resulting population dose due to accidental plutonium releases

International Code Assessment and Applications Program: Annual report

International Nuclear Information System (INIS)

Ting, P.; Hanson, R.; Jenks, R.

1987-03-01

This is the first annual report of the International Code Assessment and Applications Program (ICAP). The ICAP was organized by the Office of Nuclear Regulatory Research, United States Nuclear Regulatory Commission (USNRC) in 1985. The ICAP is an international cooperative reactor safety research program planned to continue over a period of approximately five years. To date, eleven European and Asian countries/organizations have joined the program through bilateral agreements with the USNRC. Seven proposed agreements are currently under negotiation. The primary mission of the ICAP is to provide independent assessment of the three major advanced computer codes (RELAP5, TRAC-PWR, and TRAC-BWR) developed by the USNRC. However, program activities can be expected to enhance the assessment process throughout member countries. The codes were developed to calculate the reactor plant response to transients and loss-of-coolant accidents. Accurate prediction of normal and abnormal plant response using the codes enhances procedures and regulations used for the safe operation of the plant and also provides technical basis for assessing the safety margin of future reactor plant designs. The ICAP is providing required assessment data that will contribute to quantification of the code uncertainty for each code. The first annual report is devoted to coverage of program activities and accomplishments during the period between April 1985 and March 1987

The new French code for PWR in service inspection

Energy Technology Data Exchange (ETDEWEB)

Noel, R; Hutin, J P [Electricite de France (EDF), 75 - Paris (France)

1988-12-31

This document presents the new french code for pressured water reactor in service inspection. The historic regulatory basis is presented, together with the new regulatory act (dating back to the 26 february 1974) and the major guidelines of the french practice for in service inspection of PWR components. (TEC).

Analysis of the FIST integral tests 4DBA1, 6SB2C and T1QUV with TRAC-BFl/v2001.2

International Nuclear Information System (INIS)

Analytis, G.Th.

2004-01-01

As part of the assessment of the frozen version of the PSU TRAC-BFl/v2001.2 and its qualification as a LB-LOCA and SB-LOCA code, in this work, we shall outline the comparisons between measurements and code predictions for three FIST tests: The LB-LOCA test 4DBA1, the SB-LOCA test 6SB2C and the failure to maintain water level test T1QUV. We shall study the effect of the number of axial levels in the active core as well as (in the case of the SB-LOCA test 6SB2C) the effect of the timing of the activation of the reflooding options/heat transfer package on the code predictions. Furthermore, we shall show that by using the upwinding option of some terms of the three-dimensional momentum equations, severe mass-error problems appearing in the analysis of the test T1QUV can be resolved. Generally, we shall show that although there are some differences between measurements and predictions, TRAC-BF1 captures quite well the overall behaviour of the LB-LOCA transient (depending on the number of axial nodes in the core) but underpredicts the rod surface temperatures of the SB-LOCA test 6SB2C

Numerical analysis on ingress-of-coolant events in fusion reactors with TRAC-PF1 code

International Nuclear Information System (INIS)

Ose, Yasuo; Takase, Kazuyuki; Akimoto, Hajime

2000-01-01

As for accident events related with thermal-hydraulics, in a fusion experimental reactor an ingress-of-coolant event (ICE) and a loss-of-vacuum-accident event (LOVA) should be considered. An integrated ICE/LOVA test apparatus is under planning in order to estimate quantitatively heat transfer and fluid flow characteristics under ICE and LOVA events. This study was carried out to predict numerically the thermal-hydraulic characteristics in fusion reactors at the ICE events before construction of the integrated ICE/LOVA test apparatus. The TRAC-PF1 code, which was originally developed for the thermal-hydraulic safety analysis in light water reactors, was used. The numerical analyses were performed for two kinds of system configuration with/without a pressure-suppression tank:the former for is investigation of the pressure rise characteristics and two-phase flow behavior; the latter for estimation of an effect of the pressure reduction due to the pressure-suppression tank. From the present analytical results, effects of the ingress water flow rate and vessel temperatures on the pressure rise ware clarified quantitatively. Furthermore, the pressure-rise suppression effect due to the vapor condensation in the pressure-suppression tank was predicted numerically. In addition, the useful information regarding to the design of the integrated ICE/LOVA test apparatus and the knowledge with respect to the effective usage of the TRAC-PF1 code were obtained through the present numerical study. (author)

76 FR 32227 - Core Industries, Inc., DBA Star Trac and/or Unisen, Inc., DBA STAR TRAC and/or Trac Strength...

Science.gov (United States)

2011-06-03

... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-75,192; TA-W-75,192A] Core Industries, Inc., DBA Star Trac and/or Unisen, Inc., DBA STAR TRAC and/or Trac Strength, Including On-Site Leased Workers From Aerotek, Helpmates, Mattson, and Empire Staffing, Irvine, CA; Core Industries, Inc., DBA Star Trac and/ar Unisen, Inc., DBA Star...

Updated TRAC analysis of an 80% double-ended cold-leg break for the AP600 design

International Nuclear Information System (INIS)

Lime, J.F.; Boyack, B.E.

1995-01-01

An updated TRAC 80% large-break loss-of-coolant accident (LBLOCA) has been calculated for the Westinghouse AP600 advanced reactor design, The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The 80% break size was calculated by Westinghouse to be the most severe large-break size for the AP600 design. The LBLOCA transient was calculated to 144 s. Peak cladding temperatures (PCTS) were well below the Appendix K limit of 1,478 K (2,200 F), but very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCT for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown phase. The reasons for these differences are still being investigated. Additional break sizes and break locations need to be analyzed to confirm the most severe break postulated by Westinghouse

Conversion of the thermal hydraulics components of Almaraz NPP model from RELAP5 into TRAC-M

International Nuclear Information System (INIS)

Queral, C.; Mulas, J.; Collazo, I.; Concejal, A.; Burbano, N.; Gallego, I.; Lopez Lechas, A.

2002-01-01

In the scope of a joint project between the Spanish Nuclear Regulatory Commission (CSN) and the electric energy industry of Spain (UNESA) on the USNRC state-of-the-art thermal hydraulic code, TRAC-M, there is a task relating to the translation of the Spanish NPP models from other TH codes to the new one. As part of this project, our team is working on the translation of Almaraz NPP model from RELAP5/MOD3.2 to TRAC-M. At present, several portions of the input deck have been converted to TRAC-M, and the output data have also been compared with RELAP5 data. This paper refers to the translation of the following thermal hydraulic models: pressurizer, hot and cold legs (including the pumps and the injection systems), and steam generators. The comparison of the results obtained with both codes shows a good agreement. However, some difficulties have been found in the translation of some code components, like pipes, heat structures, pumps, branchs, valves and boundary conditions. In this paper, these translation problems and their solutions are described.(author)

Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems

International Nuclear Information System (INIS)

Baiocco, G.; Petruzzi, A.; Bznuni, S.; Kozlowski, T.

2017-01-01

Highlights: • The consistency between Helios and Serpent few-group cross sections is shown. • The PARCS model is validated against a Monte Carlo 3D model. • The fission and capture rates are compared. • The influence of the spacer grids on the axial power distribution is shown. - Abstract: Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as k eff , axial and radial power, fission and capture rates were compared and found to be in good agreement.

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.

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

International Nuclear Information System (INIS)

Kmetyk, L.N.

1986-02-01

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

Evaluations on power ramp data of PWR fuels by FROST and THERMOST codes

International Nuclear Information System (INIS)

Murai, K.; Ogawa, S.; Nuno, H.; Kondo, Y.

1987-01-01

An evaluation is presented of power ramp data of Mitsubishi's PWR fuel rods tested in R-2, Studsvik, which was analysed by FROST and THERMOST codes. The analyses give good predictions for measured diameter changes and on-power rod elongations. The work indicates that FROST is capable of analysing both radial and axial pellet-cladding mechanism interaction (PCMI) appropriately, and that predicted states of PCMI (i.e. stress and strain which cannot be measured directly) are considered to be reliable. The ramp data used in the present analyses were obtained in two joint programmes with five Japanese PWR utilities (KEPCO, KYEPCO, SEPCO, HEPCO, and JAPCO). (UK)

Development of a coupling code for PWR reactor cavity radiation streaming calculation

International Nuclear Information System (INIS)

Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

2012-01-01

PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

TRAC analysis of passive containment cooling system performance

International Nuclear Information System (INIS)

Arai, Kenji; Kataoka, Kazuyoshi; Nagasaka, Hideo

1993-01-01

A passive containment cooling system (PCCS) is a promising concept to improve the reliability of decay heat removal during an accident. Toshiba has carried out analytical studies for PCCS development in addition to experimental studies, using a best estimate thermal hydraulic computer code TRAC. In order to establish an analytical model for the PCCS performance analysis, it is necessary for the analytical model to be qualified against experimental results and thoroughly address the phenomena important for PCCS performance analysis. In this paper, the TRAC qualification for PCCS application is reported. A TRAC model has been verified against a drain line break simulation test conducted at the PCCS integral test facility, GIRAFFE. The result shows that the TRAC model can accurately predict the major system response and the PCCS performance in the drain line break test. In addition, the results of several sensitivity analyses, showing various points concerning the modeling in the PCCS performance analysis, have been reported. The analyses have been carried out for the SBWR and the analytical points are closely related to important phenomena which can affect PCCS performance

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)

SCAR - Post-Accident Simulator SIPA with safety analysis code CATHARE-2 and PWR cold shutdown state simulation

International Nuclear Information System (INIS)

Farvacque, M.; Faydide, B.; Dufeil, Ph.; Raimond, E.

2003-01-01

The use of Cathare in the simulators of pressurized water reactors has been effective since the beginning of the nineties. Scar project is the second stage of the Cathare strategy for the simulators, its main objective is the extension of the field of simulation to the accident situations in cold shutdown states. Work was carried out in 3 major areas: modelling, optimization and integration in the simulator. Throughout the project, the developments were part of a 3 stages validation strategy: -) elementary tests of the developments of new model on the N4 (1450 MW PWR); -) analytical tests and systems to ensure non regression of the validation of the physical laws of the Cathare code during the modifications carried out within the optimization stage; and -) overall tests of the SIPA-CP1 (900 MW PWR) simulator, controlled automatically by programmed scenarios including the transients which are carried out in PWR, the transients of the Regulatory Guides and the accident transients

MELPROG/TRAC: update and applications

International Nuclear Information System (INIS)

Henninger, R.J.; Kelly, J.E.

1986-01-01

The first complete, coupled, and mechanistic analysis of a core meltdown sequence has been made with MELPROG-PWR/MOD1 and MELPROG/TRAC. The sequence analyzed was a station blackout accident for the Surry plant. Through vessel failure, all important aspects of the meltdown sequence were calculated. This version of MELPROG permits a full two-dimensional treatment of the in-vessel phenomena. Natural circulation can thus be modeled. Comparison to one-dimensional MELPROG and MARCH calculations shows that natural circulation reduces the rate of core heating, but increases the rate of heating of upper plenum structures and primary piping. This increased heating can inhibit fission product deposition and may lead to an early failure of the primary system. Because of uncertainty, sensitivity studies were performed to assess the relative importance of modeling assumptions. Changes in the modeling of the initial fuel rod melting and relocation were found to vary the hydrogen source by a factor of 2 and alter the timing of key events. These results imply that accurate and mechanistic modeling is important for severe accident sequence analysis

Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

International Nuclear Information System (INIS)

Peng Hong Liem; Surian Pinem; Tagor Malem Sembiring; Tran Hoai Nam

2015-01-01

A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA). The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducting to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed. (author)

Post-test analysis of semiscale tests S-UT-6 and S-UT-7 using TRAC PF1

International Nuclear Information System (INIS)

Boyack, B.E.

1983-01-01

A posttest study of Semiscale Tests S-UT-6 and S-UT-7 has been completed to assess TRAC-PFl predictions of pressurized water-reactor (PWR) small-break transients. The comparisons of the TRAC calculations and experimental results show that the correct qualitative influence of upper-head injection (UHI) was predicted. The major phenomenological difference predicted was the mode of core voiding. The data show a slow boiloff from the top of the core resulting in a dryout near the top of the core only. TRAC predicted a more extensive voiding with fluid forced from the bottom of the core by a pressure increase in the upper vessel plenum. The pressure increase was the primary consequence of a failure to predict a complete clearance of the seal in the intact-loop pump-suction upflow leg. Further review of the interphasic drag correlations, entrainment correlations, and critical-flow model is recommended. 20 figures

Contribution to study and design of PWR plant simulation code

International Nuclear Information System (INIS)

Delourme, Didier.

1980-11-01

This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr

A PWR hot-rod model: Relap5/mod3.2.2.{gamma} as a subchannel code

Energy Technology Data Exchange (ETDEWEB)

Kirsten, I.C.; Jones, J.R. [British Energy, Barnwood, Gloucester (United Kingdom); Kimber, G.R. [Atomic Energy Authority Technology, Winfrith, Dorset (United Kingdom); Page, R. [National Nuclear Corp. Ltd., Cheshire (United Kingdom)

2001-07-01

The use of the PWR transient analysis code RELAP5 for detailed assessment of Departure from Nucleate Boiling (DNB) has previously implied coupling it in some way to a subchannel code, either by direct code-to-code coupling or by transferring core boundary conditions to the subchannel code. This paper shows an alternative by using a group of subchannels modelled in RELAP5 to represent a hot rod. The model consists of three parallel channels, each more refined than its neighbour: The first channel represents a quadrant of the core; the second a quadrant on a fuel assembly and the final channel represents a passage adjacent to a single fuel pin. The model is intended for use as part of point kinetics assessments and each channel is assigned a radial form factor designed to conservatively represent the hottest fuel pins in the reactor core. The main outputs from the model are minimum Departure from Nucleate Boiling Ratio (DNBR) and clad oxidation for the hot rod (lead pin). The DNBR results from the hot-rod model are benchmarked against the subchannel code COBRA 3-CP and the results are presented in this paper. Some of the modelling problems that needed to be resolved are also highlighted. (author)

Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

International Nuclear Information System (INIS)

Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

2017-01-01

A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

Energy Technology Data Exchange (ETDEWEB)

Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

One-dimensional nodal neutronics routines for the TRAC-BD1 thermal-hydraulics program

International Nuclear Information System (INIS)

Nigg, D.W.

1983-09-01

Nuclear reactor core transient neutronic behavior is currently modeled in the TRAC-BD1 code using a point-reactor kinetics formulation. This report describes a set of subroutines based on the Analytic Nodal Method that were written to provide TRAC-BD1 with a one-dimensional space-dependent neutronics capability. Use of the routines is illustrated with several test problems. The results of these problems show that the Analytic Nodal neutronics routines have desirable accuracy and computing time characteristics and should be a useful addition to TRAC-BD1

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

International Nuclear Information System (INIS)

Mur, J.; Larrauri, D.

1998-07-01

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

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

Directory of Open Access Journals (Sweden)

Surian Pinem

2014-01-01

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

Assessment of the turbine trip transient in Cofrentes NPP with TRAC-BF1

International Nuclear Information System (INIS)

Castrillo, F.; Gomez, A.; Gallego, I.

1993-06-01

This report presents the results of the assessment of TRAC-BF1 (G1-J1) code with the model of C. N. Cofrentes for simulation of the transient originated by the manual trip of the main turbine. C. N. Cofrentes is a General Electric designed BWR/6 plant, with a nominal core thermal power of 2894 Mwt, in commercial operation since 1985, owned and operated by Hidroelectrica Espanola, S. A. The plant incorporates all the characteristics of BWR/6 reactors, with two turbine driven FW pumps. As a result of this assessment a model of C. N. Cofrentes has been developed for TRAC-BF1 that fairly reproduces operational transient behavior of the plant. A special purpose code was generated to obtain reactivity coefficients, as required by TRAC-BF1, from the 3D simulator

The KINA neutronic module of the LEGO code for steady-state and transient PWR plant simulations

International Nuclear Information System (INIS)

Nicolopoulos, D.; Pollacchini, L.; Vimercati, G.; Spelta, S.

1989-01-01

The Automation Research Center (CRA) of ENEl has implemented some models for analyzing both incidental and operational transients in PWR power plants. For such models an axial neutron kinetics module characterized by high computational efficency with adequate results accuracy was called for. CISE has been entrusted with the task of implementing such a module named KINA and based on IQS (Improved Quasi Static) method, to be included in the library of LEGO modular code used by CRA to set up PWR power models. Moreover, The KINA module has been adapted to the neutron constants computing model developed by the EdF-SEPTEN, which has been using and improving the LEGO code for a long time in cooperation with ENEL-CRA. In this paper, after some remarks on the LEGO code, a general description of KINA neutronic module is given. The resylts of a preliminary validation activity of KINA for an EdF 1300 MWe PWR plant are also presented

TRAC analysis and support of Oconee-1 PTS studies

International Nuclear Information System (INIS)

Ireland, J.R.

1983-01-01

This paper describes the overall pressurized thermal shock (PTS) program at Los Alamos with emphasis on TRAC-PF1 calculations of severe overcooling transients for the Oconee-1 pressurized water reactor (PWR). A summary of results for several calculations are presented for the Oconee-1 PWR along with detailed discussions of two of the most severe overcooling transients predicted [main steam-line break and turbine-bypass valve (TBV) failures]. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to a detailed model of the plant Integrated Control System (ICS). For the Oconee-1 main steam-line break transient, a minimum downcomer fluid temperature of approx. 405 K was predicted. For the TBV transient involving the failure of one bank of TBVs to close after initially opening following reactor and turbine trips, an extrapolated downcomer fluid temperature of approx. 365 K was estimated. The latter temperature is at the nil-ductility temperature (NDT) limit (approx. 365 K) for Oconee-1

Computer codes for the study of the loss of coolant accident of PWR reactors

International Nuclear Information System (INIS)

Gomolinski, M.; Menessier, D.; Tellier, N.

1975-01-01

The CEA has undertaken a large programme to study the consequence on the core of the LOCA of a PWR. In the programme, simultaneously carried out experiments and the development of the calculations means are described. Several experiments such as OMEGA, ERSEC and PHEBUS tests, which provide data to check the computer codes are outlined briefly in the paper. For analysis of the LOCA of a PWR, a series of computer codes, which are at present in use or under development, are linked with each other. The codes are DANAIDES for blowdown, CERES for refill and reflood, THETA-1B and FLIRA for heat up calculation during the blow-down and the reflooding period respectively. FLIRA-PASTEL, a combination of FLIRA and PASTEL which calculate the stress and deformations of material using the finite element method, will be used in place of FLIRA. The basic models and flowcharts of the above codes are described in the paper

Computational simulation of natural circulation and rewetting experiments using the TRAC/PF1 code; Simulacao computacional dos experimentos: circulacao natural no CTE-150 e remolhamento na ITR utilizando o TRAC-PF1

Energy Technology Data Exchange (ETDEWEB)

Silva, J.D. da

1994-05-01

In this work the TRAC code was used to simulate experiments of natural circulation performed in the first Brazilian integral test facility at (COPESP), Sao Paulo and a rewetting experiment in a single tube test section carried out at CDTN, Belo Horizonte, Brazil. In the first simulation the loop behavior in two transient conditions with different thermal power, namely 20 k W and 120 k W, was verified in the second one the quench front propagation, the liquid mass collected in the carry over measuring tube and the wall temperature at different elevations during the flooding experiment was measured. A comparative analysis, for code consistency, shows a good agreement between the code results and experimental data, except for the quench from velocity. (author). 15 refs, 19 figs, 12 tabs.

Development of the computer code system for the analyses of PWR core

International Nuclear Information System (INIS)

Tsujimoto, Iwao; Naito, Yoshitaka.

1992-11-01

This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Assessment of a new interfacial friction correlation in TRAC-BD1/MOD1

International Nuclear Information System (INIS)

Analytis, G.T.

1986-01-01

Analysis of a number of 5-bar (and one 1-bar) boiloff experiments in the electrically heated 33-rod bundle NEPTUN at the Swiss Federal Institute for Reactor Research by TRAC-BD1 version 12 and MOD1 has shown that the code consistently underpredicts the collapsed liquid level histories, hence predicting earlier critical heat fluxes and higher peak rod surface temperatures than the measurements showed. Moreover, recent work has demonstrated that these differences can be attributed to the bubbly/churn interfacial friction model in TRAC-BD1 (resulting in rather large interfacial drag) whose appropriateness for rod bundles is questionable; decreasing the interfacial drag resulted in excellent agreement between measurements and code predictions. Recent analysis of boiloff experiments with the French code CATHARE and with a bubbly/churn interfacial drag force f/sub i/ similar to the one of TRAC-BD1 has also resulted in the underprediction of collapsed liquid level histories. In this work, it was shown that the usual vapor drift velocity correlations (through which f/sub i/ is derived) for this flow regime developed for tubes are not appropriate for rod bundles. Moreover, a new f/sub i/ correlation for bubbly/churn flow in rod bundles was developed based on the Froude number. The authors have modified this correlation slightly, implemented it in TRAC-BD1/MOD1, and reanalyzed most of the boiloff experiments in NEPTUN

Study of the noise propagation in PWR with coupled codes

International Nuclear Information System (INIS)

Verdu, G.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Barrachina, T.

2011-01-01

The in-core detectors provide signals of the power distribution monitoring for the Reactor Protection System (RPS). The advanced fuel management strategies (high exposure) and the power upratings for PWR reactor types have led to an increase in the noise amplitude in detectors signals. In the present work a study of the propagation along the reactor core and the effects on the core power evolution of a small perturbation on the moderator density, using the coupled code RELAP5-MOD3.3/PARCSv2.7 is presented. The purpose of these studies is to be able to reproduce and analyze the in-core detector simulated signals. (author)

A methodology for the coupling of RAMONA-3B neutron kinetics and TRAC-BF1 thermal-hydraulics

International Nuclear Information System (INIS)

Lopez, Arsenio Procopio; Morales Sandoval, Jaime B.

2005-01-01

The initial objective of this project was to directly couple the RAMONA and TRAC codes running on different PCs. The idea was to use the best part of each one and eliminate some of their limitations and widen the applicability of these codes to simulate different BWR and system components. It was required to try to minimize the amount of changes to present code subroutines and calculation procedures. If possible, just substitute values obtained in the parallel code. Preliminary results indicated that using a CHAN component of TRAC to model thermal-hydraulic phenomena for each neutronic channel modeled in RAMONA is rather difficult. Large amounts of CPU time consumption are obtained and lots of PCs would make this solution difficult, besides considerable large transients are introduced by the differences in thermal-hydraulic results of these codes. The substitution of the thermal-hydraulics of RAMONA, by the TRAC channel calculations, is possible but simulation of a null transient on both codes must be planed and a gradual change must be controlled by an additional supervisory subroutine. An indirect coupling of these codes, it is therefore proposed, in order to eliminate most of these limitations. In this indirect coupling, a thermal-hydraulic model of the average tube in a bundle and the global channel cooling fluid dynamics is programmed for each neutronic channel while the global reactor vessel and core is modeled by TRAC with just four channels and four rings. Results are more reliable, coupling is simpler and faster simulations are possible

Verification of NUREC Code Transient Calculation Capability Using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem

International Nuclear Information System (INIS)

Joo, Hyung Kook; Noh, Jae Man; Lee, Hyung Chul; Yoo, Jae Woon

2006-01-01

In this report, we verified the NUREC code transient calculation capability using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem. The benchmark problem consists of Part 1, a 2-D problem with given T/H conditions, Part 2, a 3-D problem at HFP condition, Part 3, a 3-D problem at HZP condition, and Part 4, a transient state initiated by a control rod ejection at HZP condition in Part 3. In Part 1, the results of NUREC code agreed well with the reference solution obtained from DeCART calculation except for the pin power distributions at the rodded assemblies. In Part 2, the results of NUREC code agreed well with the reference DeCART solutions. In Part 3, some results of NUREC code such as critical boron concentration and core averaged delayed neutron fraction agreed well with the reference PARCS 2G solutions. But the error of the assembly power at the core center was quite large. The pin power errors of NUREC code at the rodded assemblies was much smaller the those of PARCS code. The axial power distribution also agreed well with the reference solution. In Part 4, the results of NUREC code agreed well with those of PARCS 2G code which was taken as the reference solution. From the above results we can conclude that the results of NUREC code for steady states and transient states of the MOX loaded LWR core agree well with those of the other codes

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

International Nuclear Information System (INIS)

Knight, T.D.

1996-07-01

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

Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code

Directory of Open Access Journals (Sweden)

Tagor Malem Sembiring

2017-01-01

Full Text Available The in-house coupled neutronic and thermal-hydraulic (N/T-H code of BATAN (National Nuclear Energy Agency of Indonesia, NODAL3, based on the few-group neutron diffusion equation in 3-dimensional geometry using the polynomial nodal method, has been verified with static and transient PWR benchmark cases. This paper reports the verification of NODAL3 code in the NEA-NSC PWR uncontrolled control rods withdrawal at zero power benchmark. The objective of this paper is to determine the accuracy of NODAL3 code in solving the continuously slow and fast reactivity insertions due to single and group of control rod bank withdrawn while the power and temperature increment are limited by the Doppler coefficient. The benchmark is chosen since many organizations participated using various methods and approximations, so the calculation results of NODAL3 can be compared to other codes’ results. The calculated parameters are performed for the steady-state, transient core averaged, and transient hot pellet results. The influence of radial and axial nodes number was investigated for all cases. The results of NODAL3 code are in very good agreement with the reference solutions if the radial and axial nodes number is 2 × 2 and 2 × 18 (total axial layers, respectively.

Implantation of FRAPCON-2 code in HB computer

International Nuclear Information System (INIS)

Silva, C.F. da.

1987-05-01

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

Translation of the model plant of the CN code TRAC-BF1 Cofrentes of a SNAP-TRACE

International Nuclear Information System (INIS)

Escriva, A.; Munuz-Cobo, J. L.; Concejal, A.; Melara, J.; Albendea, M.

2012-01-01

It aims to develop a three-dimensional model of the CN Cofrentes whose consistent results Compared with those in current use programs (TRAC-BFl, RETRAN) validated with data of the plant. This comparison should be done globally and that you can not carry a compensation of errors. To check the correct translation of the results obtained have been compared with TRACE and the programs currently in use and the relevant adjustments have been made, taking into account that both the correlations and models are different codes. During the completion of this work we have detected several errors that must be corrected in future versions of these tools.

PWR core follow calculations using the ELCOS code system

International Nuclear Information System (INIS)

Grimm, P.; Paratte, J.M.

1990-01-01

The ELCOS code system developed at PSI is used to simulate a cycle of a PWR in which one fifth of the assemblies are MOX fuel. The reactor and the calculational methods are briefly described. The calculated critical boron concentrations and power distributions are compared with the measurements at the plant. Although the critical boron concentration is somewhat overpredicted and the computed power distributions are slightly flatter than the measured ones the results of the calculations agree generally well with the measured data. (author) 1 tab., 8 figs., 6 refs

Improvements to TRAC models of condensing stratified flow. Pt. 1

International Nuclear Information System (INIS)

Zhang, Q.; Leslie, D.C.

1991-12-01

Direct contact condensation in stratified flow is an important phenomenon in LOCA analyses. In this report, the TRAC interfacial heat transfer model for stratified condensing flow has been assessed against the Bankoff experiments. A rectangular channel option has been added to the code to represent the experimental geometry. In almost all cases the TRAC heat transfer coefficient (HTC) over-predicts the condensation rates and in some cases it is so high that the predicted steam is sucked in from the normal outlet in order to conserve mass. Based on their cocurrent and countercurrent condensing flow experiments, Bankoff and his students (Lim 1981, Kim 1985) developed HTC models from the two cases. The replacement of the TRAC HTC with either of Bankoff's models greatly improves the predictions of condensation rates in the experiment with cocurrent condensing flow. However, the Bankoff HTC for countercurrent flow is preferable because it is based only on the local quantities rather than on the quantities averaged from the inlet. (author)

Conversion of control systems, protection and engineering safeguard system signals of Almaraz NPP model from RELAP5 into TRAC-M

International Nuclear Information System (INIS)

Mulas, J.; Queral, C.; Collazo, I.; Concejal, A.; Burbano, N.; Lopez Lechas, A.; Tarrega, I.

2002-01-01

In the scope of a joint project between the Spanish Regulatory Commission (CSN) and the electric energy industry of Spain (UNESA) about the USNRC state-of-art thermal hydraulic code, TRAC-M, there is a task relating to the translation of the Spanish NPP models from other TH codes to the new one. As a part of this project, our team is working on the translation of Almaraz NPP model from RELAP5/MOD3.2 to TRAC-M. One of the goals of the project is to analyze the conversion of control blocks, signal variables and trips in order to correct modelling all instrumentation and control systems, and also protection and engineering safeguard system-signals of the NPP. At present, several portions of the input deck have been converted to TRAC-M, and the output data have also been compared with RELAP5 data. This paper describes the problems found in the conversion and the solutions achieved.(author)

Comparison of Serpent and HELIOS-2 as applied for the PWR few-group cross section generation

International Nuclear Information System (INIS)

Fridman, E.; Leppaenen, J.; Wemple, C.

2013-01-01

This paper discusses recent modifications to the Serpent Monte Carlo code methodology and related to the calculation of few-group diffusion coefficients and reflector discontinuity factors The new methods were assessed in the following manner. First, few-group homogenized cross sections calculated by Serpent for a reference PWR core were compared with those generated 1 commercial deterministic lattice transport code HELIOS-2. Second, Serpent and HELIOS-2 fe group cross section sets were later employed by nodal diffusion code DYN3D for the modeling the reference PWR core. Finally, the nodal diffusion results obtained using the both cross section sets were compared with the full core Serpent Monte Carlo solution. The test calculations show that Serpent can calculate the parameters required for nodal analyses similar to conventional deterministic lattice codes. (authors)

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

Energy Technology Data Exchange (ETDEWEB)

Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: stephane.mimouni@edf.fr [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: giovanni.manzini@rse-web.it [RSE, Milano (Italy); Xiao, J., E-mail: jianjun.xiao@kit.edu [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: vyl@ujv.cz [UJV Rez (Czech Republic); Siccama, N.B., E-mail: siccama@nrg.eu [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: risto.huhtanen@vtt.fi [VTT, PO Box 1000, FI-02044 VTT (Finland)

2015-02-15

Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

International Nuclear Information System (INIS)

Malet, J.; Mimouni, S.; Manzini, G.; Xiao, J.; Vyskocil, L.; Siccama, N.B.; Huhtanen, R.

2015-01-01

Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety

Application of a generalized interface module to the coupling of PARCS with both RELAPS and TRAC-M

International Nuclear Information System (INIS)

Barber, D.A.; Wang, W.; Miller, R.M.; Downar, T.J.; Joo, H.G.; Mousseau, V.A.; Ebert, D.E.

1999-01-01

In an effort to more easily assess various combinations of 3-D neutronic/thermal-hydraulic codes, the USNRC has sponsored the development of a generalized interface module for the coupling of any thermal-hydraulics code to any spatial kinetics code. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine (PVM) software to manage inter-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCS, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for an OECD/NEA main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated; nonetheless, the capabilities of the coupled code are presented for the OECD/NEA main steam line break benchmark problem

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

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

TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

International Nuclear Information System (INIS)

Lime, J.F.; Boyack, B.E.

1996-01-01

An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

International Nuclear Information System (INIS)

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

1984-02-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided

Application of a generalized interface module to the coupling of PARCS with both RELAP5 and TRAC-M

Energy Technology Data Exchange (ETDEWEB)

Barber, D.A.; Wang, W. [SCIENTECH, Inc. (United States); Miller, R.M.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States); Joo, H.G. [Korean Atomic Energy Research Inst., Seoul (Korea, Republic of); Mousseau, V.A. [Los Alamos National Lab., NM (United States); Ebert, D.E. [Nuclear Regulatory Commission, Washington, DC (United States)

1999-04-01

In an effort to more easily assess various combinations of 3-D neutronic/thermal-hydraulic codes, the USNRC has sponsored the development of a generalized interface module for the coupling of any thermal-hydraulics code to any spatial kinetics code. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine (PVM) software to manage inter-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCS, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for an OECD/NEA main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated; nonetheless, the capabilities of the coupled code are presented for the OECD/NEA main steam line break benchmark problem.

Development of a computer code for transients simulation in PWR type reactors

International Nuclear Information System (INIS)

Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de

1981-01-01

A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt

Posttest analysis of MIST Test 330302 using TRAC-PF1/MOD1

International Nuclear Information System (INIS)

Boyack, B.E.

1992-09-01

This report discusses a posttest analysis of Multi-Loop Integral System Test (MIST) 330302 which has been performed using TRAC-PF1/MOD1. This test was one of group performed in the MIST facility to investigate high-pressure injection (HPI)-power-operated relief valve (PORV) cooling, also known as feed-and-bleed cooling. In Test 330302, HPI cooling was delayed 20 min after opening and locking the PORV open to induce extensive system voiding. We have concluded that the TRAC-calculated results are in reasonable overall agreement with the data for Test 330302. All major trends and phenomena were correctly predicted. Differences observed between the measured and calculated results have been traced and related, in part, to deficiencies in our knowledge of the facility configuration and operation. We have identified two models forwhich additional review is appropriate. However, in general, the TRAC closure models and correlations appear to be adequate for the prediction of the phenomena expected to occur during feed-and-bleed transientsin the MIST facility. We believe that the correct conclusions about trends and phenomena will be reached if the code is used in similar applications. Conclusions reached regarding use of the code to calculate similar phenomena in full-size plants (scaling implications) and regulatory implications of this work are also presented

A probabilistic analysis of PWR and BWR fuel rod performance using the code CASINO-SLEUTH

International Nuclear Information System (INIS)

Bull, A.J.

1987-01-01

This paper presents a brief description of the Monte Carlo and response surface techniques used in the code, and a probabilistic analysis of fuel rod performance in PWR and BWR applications. The analysis shows that fission gas release predictions are very sensitive to changes in certain of the code's inputs, identifies the most dominant input parameters and compares their effects in the two cases. (orig./HP)

Improvement of neutron kinetics module in TRAC-BF1code: one-dimensional nodal collocation method

Energy Technology Data Exchange (ETDEWEB)

Jambrina, Ana; Barrachina, Teresa; Miro, Rafael; Verdu, Gumersindo, E-mail: ajambrina@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidade Politecnica de Valencia (UPV), Valencia (Spain); Soler, Amparo, E-mail: asoler@iberdrola.es [SEA Propulsion S.L., Madrid (Spain); Concejal, Alberto, E-mail: acbe@iberdrola.es [Iberdrola Ingenieria y Construcion S.A.U., Madrid (Spain)

2013-07-01

The TRAC-BF1 one-dimensional kinetic model is a formulation of the neutron diffusion equation in the two energy groups' approximation, based on the analytical nodal method (ANM). The advantage compared with a zero-dimensional kinetic model is that the axial power profile may vary with time due to thermal-hydraulic parameter changes and/or actions of the control systems but at has the disadvantages that in unusual situations it fails to converge. The nodal collocation method developed for the neutron diffusion equation and applied to the kinetics resolution of TRAC-BF1 thermal-hydraulics, is an adaptation of the traditional collocation methods for the discretization of partial differential equations, based on the development of the solution as a linear combination of analytical functions. It has chosen to use a nodal collocation method based on a development of Legendre polynomials of neutron fluxes in each cell. The qualification is carried out by the analysis of the turbine trip transient from the NEA benchmark in Peach Bottom NPP using both the original 1D kinetics implemented in TRAC-BF1 and the 1D nodal collocation method. (author)

A thermal hydraulic analysis in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code

International Nuclear Information System (INIS)

Santos, Thiago A. dos; Maiorino, José R.; Stefanni, Giovanni L. de

2017-01-01

In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O 2 . For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O 2 .The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO 2 was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

The development of a transient neutron flux solution in the PANTHER code

International Nuclear Information System (INIS)

Hutt, P.K.; Knight, M.P.

1990-01-01

In the United Kingdom a new three-dimensional, two-group, homogeneous reactor diffusion code, PANTHER, has been developed for the analysis of pressurized water reactors (PWRs) and advanced gas-cooled reactors (AGRs). The code can perform a comprehensive range of calculations, steady state, depletion, and transient with either a finite difference or analytic nodal flux solution. The nodal solution allows the representation of within-node burnup variation and pin-power reconstruction in either steady-state or transient mode. Specific steady-state and transient thermal feedback modules are included for both PWRs and AGRs. The code is being developed to perform a complete range of reactor calculations from online operational support to fuel management and fault transient analysis. In the area of transient analysis, the code is currently being used for a number of PWR fault transient assessments, including rod ejection and steam-line break. In addition, work is proceeding to incorporate the PANTHER 3D nodal transient solution in the TRAC-P code. This paper outlines the development of the transient flux solutions within PANTHER

A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

International Nuclear Information System (INIS)

Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

2002-01-01

A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

Application of Melcor code for the calculo of TMLB sequence in PWR with natural circulating into the vessel

International Nuclear Information System (INIS)

Marten-Fuertes, F.

1995-01-01

The use of computer codes to analyze the phenomena of severe accidents is very important to take decisions in Nuclear Safety. This paper presents the MELCOR code used to calculate the TMLB sequence of PWR with natural circulation into the vessels. The main goal of this code is its application for the PSA (probabilistic safety analysis)

Translation of the model plant of the CN code TRAC-BF1 Cofrentes of a SNAP-TRACE; Traduccion del modelo de planta de CN Cofrentes del codigo TRAC-BF1 a SNAP-TRACE

Energy Technology Data Exchange (ETDEWEB)

Escriva, A.; Munuz-Cobo, J. L.; Concejal, A.; Melara, J.; Albendea, M.

2012-07-01

It aims to develop a three-dimensional model of the CN Cofrentes whose consistent results Compared with those in current use programs (TRAC-BFl, RETRAN) validated with data of the plant. This comparison should be done globally and that you can not carry a compensation of errors. To check the correct translation of the results obtained have been compared with TRACE and the programs currently in use and the relevant adjustments have been made, taking into account that both the correlations and models are different codes. During the completion of this work we have detected several errors that must be corrected in future versions of these tools.

Thermal-hydraulic study of integrated steam generator in PWR

International Nuclear Information System (INIS)

Osakabe, Masahiro

1989-01-01

One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC

International Nuclear Information System (INIS)

Gomez Garcia-Torano, I.; Jimenez, G.

2013-01-01

The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

A computer code PACTOLE to predict activation and transport of corrosion products in a PWR

International Nuclear Information System (INIS)

Beslu, P.; Frejaville, G.; Lalet, A.

1978-01-01

Theoretical studies on activation and transport of corrosion products in a PWR primary circuit have been concentrated, at CEA on the development of a computer code : PACTOLE. This code takes into account the major phenomena which govern corrosion products transport: 1. Ion solubility is obtained by usual thermodynamics laws in function of water chemistry: pH at operating temperature is calculated by the code. 2. Release rates of base metals, dissolution rates of deposits, precipitation rates of soluble products are derived from solubility variations. 3. Deposition of solid particles is treated by a model taking into account particle size, brownian and turbulent diffusion and inertial effect. Erosion of deposits is accounted for by a semi-empirical model. After a review of calculational models, an application of PACTOLE is presented in view of analyzing the distribution of in core. (author)

Evaluations of the CCFL and critical flow models in TRACE for PWR LBLOCA analysis

Energy Technology Data Exchange (ETDEWEB)

Yang, Jung-Hua; Lin, Hao Tzu [National Tsing Hua Univ., HsinChu, Taiwan (China). Dept. of Engineering and System Science; Wang, Jong-Rong [Atomic Energy Council, Taoyuan County, Taiwan (China). Inst. of Nuclear Energy Research; Shih, Chunkuan [National Tsing Hua Univ., HsinChu, Taiwan (China). Inst. of Nuclear Engineering and Science

2012-12-15

This study aims to develop the Maanshan Pressurized Water Reactor (PWR) analysis model by using the TRACE (TRAC/RELAP Advanced Computational Engine) code. By analyzing the Large Break Loss of Coolant Accident (LBLOCA) sequence, the results are compared with the Maanshan Final Safety Analysis Report (FSAR) data. The critical flow and Counter Current Flow Limitation (CCFL) play an important role in the overall performance of TRACE LBLOCA prediction. Therefore, the sensitivity study on the discharge coefficients of critical flow model and CCFL modeling among different regions are also discussed. The current conclusions show that modeling CCFL in downcomer has more significant impact on the peak cladding temperature than modeling CCFL in hot-legs does. No CCFL phenomena occurred in the pressurizer surge line. The best value for the multipliers of critical flow model would be 0.5 and the TRACE could consistently predict the break flow rate in the LBLOCA analysis as shown in FSAR. (orig.)

Summary of transient analysis

International Nuclear Information System (INIS)

Saha, P.

1984-01-01

This chapter reviews the papers on the pressurized water reactor (PWR) and boiling water reactor (BWR) transient analyses given at the American Nuclear Society Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors. Most of the papers were based on the systems calculations performed using the TRAC-PWR, RELAP5 and RETRAN codes. The status of the nuclear industry in the code applications area is discussed. It is concluded that even though comprehensive computer codes are available for plant transient analysis, there is still a need to exercise engineering judgment, simpler tools and even hand calculations to supplement these codes

TRAC-PF1 MOD1 post test calculations of the OECD LOFT Experiment LP-SB-1

International Nuclear Information System (INIS)

Allen, E.J.

1990-04-01

Analysis of the small, hot leg break, OECD LOFT Experiment LP-SB-1. using the ''best-estimate'' computer code TRAC-PF1/MOD1 is presented. Descriptions of the LOFT facility and the LP-SB-1 experiment are given and development of the TRAC-PF1/MOD1 input model is detailed. The calculations performed in achieving the steady state conditions, from which the experiment was initiated, and the specification of experimental boundary conditions are outlined. 24 refs., 66 figs., 12 tabs

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

International Nuclear Information System (INIS)

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

1978-12-01

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

FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

Directory of Open Access Journals (Sweden)

Thiollay Nicolas

2016-01-01

Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

Hamor-2: a computer code for LWR inventory calculation

International Nuclear Information System (INIS)

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

1985-01-01

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

Analysis of dynamic behavior of a PWR utilizing the computer program SARDAN 2

International Nuclear Information System (INIS)

Pessanha, J.A.O.

1982-07-01

In the design of a PWR nuclear plant it is necessary to verify if the design limits are respected, even under abnormal operation condition. An evolution of SARDAN code, developed to simulate transients in PWR, are presented. The new aspects incorporeted in SARDAN 2 are: the fuel ROD analysis in finite-diference, an open channel model for the critic subchannel analysis and the introduction of a simplified model for the automatic control system. The program has been tested in accident condition II, in special, uncontrolled ROD cluster assembly bank withoraw, dropped full-length assembly group, uncontrolled Boron dilution, and the results obtained were considered satisfactory. (Author) [pt

Development a computer codes to couple PWR-GALE output and PC-CREAM input

Science.gov (United States)

Kuntjoro, S.; Budi Setiawan, M.; Nursinta Adi, W.; Deswandri; Sunaryo, G. R.

2018-02-01

Radionuclide dispersion analysis is part of an important reactor safety analysis. From the analysis it can be obtained the amount of doses received by radiation workers and communities around nuclear reactor. The radionuclide dispersion analysis under normal operating conditions is carried out using the PC-CREAM code, and it requires input data such as source term and population distribution. Input data is derived from the output of another program that is PWR-GALE and written Population Distribution data in certain format. Compiling inputs for PC-CREAM programs manually requires high accuracy, as it involves large amounts of data in certain formats and often errors in compiling inputs manually. To minimize errors in input generation, than it is make coupling program for PWR-GALE and PC-CREAM programs and a program for writing population distribution according to the PC-CREAM input format. This work was conducted to create the coupling programming between PWR-GALE output and PC-CREAM input and programming to written population data in the required formats. Programming is done by using Python programming language which has advantages of multiplatform, object-oriented and interactive. The result of this work is software for coupling data of source term and written population distribution data. So that input to PC-CREAM program can be done easily and avoid formatting errors. Programming sourceterm coupling program PWR-GALE and PC-CREAM is completed, so that the creation of PC-CREAM inputs in souceterm and distribution data can be done easily and according to the desired format.

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

International Nuclear Information System (INIS)

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

1984-01-01

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

A thermal hydraulic analysis in PWR reactors with UO{sub 2} or (U-Th)O{sub 2} fuel rods employing a simplified code

Energy Technology Data Exchange (ETDEWEB)

Santos, Thiago A. dos; Maiorino, José R., E-mail: thiago.santos@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil); Stefanni, Giovanni L. de, E-mail: giovanni.stefanni@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh-100 reactor thermal limits. This PWR is a project developed in Universidade Federal do ABC (UFABC) using fuel composed of Uranium and Thorium oxide mixed (U,Th)O{sub 2}. For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named STC-MOX-Th 'Simplified Thermal-hydraulics Code-Mixed Oxide Thorium'. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O{sub 2}.The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficient as a temperature function. In order to solve this equation, the finite elements method was used. Furthermore, the proportion of 36% of UO{sub 2} was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middle and ending. The program has proven to be efficient in every condition and the results evidenced that the APTh-1000 reactor, in an initial analysis, has its thermal limits within the recommended security parameters. (author)

ASTEC-CATHARE2 benchmarks on French PWR 1300MWe reactors

International Nuclear Information System (INIS)

Tregoures, Nicolas; Philippot, Marc; Foucher, Laurent; Guillard, Gaetan; Fleurot, Joelle

2009-01-01

The French Institut de Radioprotection et de Surete Nucleaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe reactors. This PSA-2 is heavily relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, an important series of benchmarks with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe accident scenarios. The present paper details 2 out of the 14 studied scenarios: a 12 inches cold leg Loss of Coolant Accident (LOCA) and a 2 tubes Steam Generator Tube Rupture (SGTR). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and the ASTEC results of the core degradation phase are presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results. (author)

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

2006-01-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

Energy Technology Data Exchange (ETDEWEB)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

2006-07-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

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

The CAIN computer code for the generation of MABEL input data sets: a user's manual

International Nuclear Information System (INIS)

Tilley, D.R.

1983-03-01

CAIN is an interactive FORTRAN computer code designed to overcome the substantial effort involved in manually creating the thermal-hydraulics input data required by MABEL-2. CAIN achieves this by processing output from either of the whole-core codes, RELAP or TRAC, interpolating where necessary, and by scanning RELAP/TRAC output in order to generate additional information. This user's manual describes the actions required in order to create RELAP/TRAC data sets from magnetic tape, to create the other input data sets required by CAIN, and to operate the interactive command procedure for the execution of CAIN. In addition, the CAIN code is described in detail. This programme of work is part of the Nuclear Installations Inspectorate (NII)'s contribution to the United Kingdom Atomic Energy Authority's independent safety assessment of pressurized water reactors. (author)

Assessment of Ultimate Load Capacity for Pre-Stressed Concrete Containment Vessel Model of PWR Design With BARC Code ULCA

International Nuclear Information System (INIS)

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

2002-01-01

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

Study on Reactor Physics Characteristic of the PWR Core Using UO2

International Nuclear Information System (INIS)

Tukiran Surbakti

2009-01-01

Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

Laguna Verde simulator: A new TRAC-RT based application

International Nuclear Information System (INIS)

Munoz Cases, J.J.; Tanarro Onrubia, A.

2006-01-01

In a partnership with GSE Systems, TECNATOM is developing a full scope training simulator for Laguna Verde Unit 2 (LV2). The simulator design is based upon the current 'state-of-the art technology' regarding the simulation platform, instructor station, visualization tools, advanced thermalhydraulics and neutronics models, I/O systems and automated model building technology. When completed, LV2 simulator will achieve a remarkable level of modeling fidelity by using TECNATOM's TRAC-RT advanced thermalhydraulic code for the reactor coolant and main steam systems, and NEMO neutronic model for the reactor core calculations. These models have been utilized up to date for the development or upgrading of nine NPP simulators in Spain and abroad, with more than 8000 hours of training sessions, and have developed an excellent reputation for its robustness and high fidelity. (author)

Three-dimensional two-fluid numerical treatment of a reactor vessel in TRAC

International Nuclear Information System (INIS)

Liles, D.R.

1979-01-01

A three-dimensional two-fluid finite difference model has been used in TRAC (Transient Reactor Analysis Code) to represent a pressurized water reactor vessel. Mesh cells may be blocked off completely to represent large flow obstructions such as downcomer walls. The hydrodynamic volumes and flow areas may also be reduced in order to provide a porous matrix simulation of smaller scale strucuture. The finite difference equations are semi-implicit so that stability time scales are associated with material movement and not wave propagation. The block matrix structure is reduced during the implicit pass to a single element seven stripe system which is easily solved iteratively. This procedure has successfully performed numerous simulations of both full sized reactor accidents and smaller scale experments. It has proven to be a useful feature of the TRAC effort

Analysis of PWR control rod ejection accident with the coupled code system SKETCH-INS/TRACE by incorporating pin power reconstruction model

International Nuclear Information System (INIS)

Nakajima, T.; Sakai, T.

2010-01-01

The pin power reconstruction model was incorporated in the 3-D nodal kinetics code SKETCH-INS in order to produce accurate calculation of three-dimensional pin power distributions throughout the reactor core. In order to verify the employed pin power reconstruction model, the PWR MOX/UO_2 core transient benchmark problem was analyzed with the coupled code system SKETCH-INS/TRACE by incorporating the model and the influence of pin power reconstruction model was studied. SKETCH-INS pin power distributions for 3 benchmark problems were compared with the PARCS solutions which were provided by the host organisation of the benchmark. SKETCH-INS results were in good agreement with the PARCS results. The capability of employed pin power reconstruction model was confirmed through the analysis of benchmark problems. A PWR control rod ejection benchmark problem was analyzed with the coupled code system SKETCH-INS/ TRACE by incorporating the pin power reconstruction model. The influence of pin power reconstruction model was studied by comparing with the result of conventional node averaged flux model. The results indicate that the pin power reconstruction model has significant effect on the pin powers during transient and hence on the fuel enthalpy

Assessment of selected TRAC and RELAP5 calculations for Oconee-1 pressurized thermal shock study

International Nuclear Information System (INIS)

Rohatgi, U.S.; Pu, J.; Saha, P.; Jo, J.

1984-11-01

Several Oconee-1 overcooling transients that were computed by LANL and INEL using the latest versions of TRAC-PF1 and RELAPS/MOD1.5 codes have been reviewed by BNL. Three of these transients were selected for detailed review as they either had the potential of challenging the integrity of the pressure vessel or highlighted the effect of code differences. These are: (1) Main Steam Line Break (MSLB); (2) All Turbine Bypass Valves Stuck Open; and (3) 2-Inch Small Break LOCA. Both codes were reasonably successful in modeling these transients. However, there were differences in the code results even though the specified scenarios were exactly the same for two transients (MSLB and Small Break LOCA). This report compares the code results and explains the possible reasons for these differences. Recommendations have been made regarding which result seems more reasonable for a specific transient

Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

International Nuclear Information System (INIS)

Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

2013-10-01

The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

Assessment of TRAC-BD1/MOD1 using FIST data

International Nuclear Information System (INIS)

Jo, J.H.; Connell, H.R.

1985-01-01

This report is concerned with the assessment of the TRAC-BD1/MOD1 Code, developed at Idaho National Engineering Laboratory. The assessment was conducted using data from the FIST (Full Integral Simulation Test) facility, which is a BWR safety test facility which was built to investigate small break LOCA and operational transients in BWR's and to complement earlier large break LOCA test results from TLTA (Two-Loop Test Apparatus). 21 figs

The effect of hot spots upon swelling of Zircaloy cladding as modelled by the code CANSWEL-2

International Nuclear Information System (INIS)

Haste, T.J.; Gittus, J.H.

1980-12-01

The code CANSWEL-2 models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised-water reactor (PWR). It can treat azimuthal non-uniformities in cladding thickness and temperature, and model the mechanical restraint imposed by the nearest neighbouring rods, including situations where cladding is forced into non-circular shapes. The physical and mechanical models used in the code are presented. Applications of the code are described, both as a stand-alone version and as part of the PWR LOCA code MABEL-2. Comparison with a limited number of relevant out-of-reactor creep strain experiments has generally shown encouraging agreement with the data. (author)

Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

International Nuclear Information System (INIS)

Croff, A.G.; Bjerke, M.A.; Morrison, G.W.; Petrie, L.M.

1978-09-01

Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given

Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

International Nuclear Information System (INIS)

Sadek, Sinisa; Amizic, Milan; Grgic, Davor

2013-01-01

The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

Fuel behaviour calculations with version 2.0 of the code FUROM

International Nuclear Information System (INIS)

Kulacsy, K.

2011-01-01

The fuel modelling code FUROM (FUel ROd Model), suitable for calculating the normal operation condition behaviour of PWR and WWER fuels, has been developed at AEKI for several years. In 2010 the new version of the code, FUROM-2.0 was released. Calculations performed with this version and results are presented. (author)

Quality assurance aspects of the computer code CODAR2

International Nuclear Information System (INIS)

Maul, P.R.

1986-03-01

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

Design study of a PWR of 1.300 MWe of Angra-2 type operating in the thorium cycle

International Nuclear Information System (INIS)

Andrade, E.P.; Carneiro, F.A.N.; Schlosser, G.J.

1984-01-01

The utilization of the thorium-highly enriched uranium and thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. The PWR of 1300 MWe from KWU (Angra-2 type) is taken as the reference reactor for the study. Reactor core design calculations for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM 2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. (Author) [pt

Special small-break applications with TRAC

International Nuclear Information System (INIS)

Dobranich, D.; DeMuth, N.S.; Henninger, R.J.; Burns, R.D. III.

1981-01-01

Input models for the Transient Reactor Analysis Code (TRAC) are described and applications of these models to reactor transients involving small breaks in the primary coolant pressure boundary are demonstrated. The operation of the primary overpressure protection system (relief and safety valves) and the thermal-hydraulic response of the reactor to these transients are obtained from numerical simulations. Also, the effects of steam generator recirculation, steam generator tube rupture, Emergency Core Cooling (ECC) injection and reactivity feedback on the course and consequences of these transients are investigated. These models allow reliable predictions of accident signatures that can help determine the adequacy of equipment and procedures at nuclear power plants to prevent and to control severe accidents

Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Min; Sabundjian, Gaianê, E-mail: smlee@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

International Nuclear Information System (INIS)

Lee, Seung Min; Sabundjian, Gaianê

2017-01-01

The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents

International Nuclear Information System (INIS)

Fermandjian, J.; Beonio-Brocchieri, F.

1986-09-01

The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes used in reactor safety in order to assess their capability of realistically describing the aerosol behavior in PWR reactor containment buildings during severe accidents. The codes included in the present study are the following: AEROSIM-M, AEROSOLS/Bl, CORRAL-2, NAUA Mod5. In AEROSIM-M, AEROSOLS/Bl and NAUA Mod5, the integro-differential equation for the evolution of the particle mass distribution is approximated by a set of coupled first order differential equations. To this end, the particle distribution function is replaced by a number of discrete monodisperse fractions. The CORRAL-2 has an essentially empirical basis (processes not explicitely modelled, but their net effects accounted for). The physical processes taken into account in the codes are shown finally

Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

International Nuclear Information System (INIS)

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

2014-01-01

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

Analyses of PWR spent fuel composition using SCALE and SWAT code systems to find correction factors for criticality safety applications adopting burnup credit

Energy Technology Data Exchange (ETDEWEB)

Shin, Hee Sung; Suyama, Kenya; Mochizuki, Hiroki; Okuno, Hiroshi; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-01-01

The isotopic composition calculations were performed for 26 spent fuel samples from the Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using the SAS2H module of the SCALE4.4 code system with 27, 44 and 238 group cross-section libraries and the SWAT code system with the 107 group cross-section library. For the analyses of samples from the Obrigheim PWR reactor, geometrical models were constructed for each of SCALE4.4/SAS2H and SWAT. For the analyses of samples from 7 PWR reactors, the geometrical model already adopted in the SCALE/SAS2H was directly converted to the model of SWAT. The four kinds of calculation results were compared with the measured data. For convenience, the ratio of the measured to calculated values was used as a parameter. When the ratio is less than unity, the calculation overestimates the measurement, and the ratio becomes closer to unity, they have a better agreement. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed, however: (1) Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from the Obrigheim reactor (10 samples were deselected because their burnups were measured with Cs-137 non-destructive method, less reliable than Nd-148 method the rest 16 samples were measured with); (2) Larger than unity ratios were found for Am-241 and Cm-242 for both the 16 and 55 samples; (3) Larger than unity ratios were found for Sm-149 for the 55 samples; (4) SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor

Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

1986-02-01

The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

Energy Technology Data Exchange (ETDEWEB)

Sekhri, A.; Baumann, P. [KernkraftwerkLeibstadt AG, 5325 Leibstadt (Switzerland); Wicaksono, D. [Swiss Federal Inst. of Technology Zurich ETH, 8092 Zurich (Switzerland); Miro, R.; Barrachina, T.; Verdu, G. [Inst. for Industrial, Radiophysical and Environmental Safety ISIRYM, Universitat Politecnica de Valencia UPV, Cami de Vera s/n, 46021 Valencia (Spain)

2012-07-01

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

International Nuclear Information System (INIS)

Sekhri, A.; Baumann, P.; Wicaksono, D.; Miro, R.; Barrachina, T.; Verdu, G.

2012-01-01

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

TRAC-PF1/MOD3 calculations of Savannah River Laboratory Rig FA single-annulus heated experiments

International Nuclear Information System (INIS)

Fischer, S.R.; McDaniel, C.K.

1992-01-01

This paper presents the results of TRAC-PF1/MOD3 benchmarks of the Rig FA experiments performed at the Savannah River Laboratory to simulate prototypic reactor fuel assembly behavior over a range of fluid conditions typical of the emergency cooling system (ECS) phase of a loss-of-coolant accident (LOCA). The primary purpose of this work was to use the SRL Rig FA tests to qualify the TRAC-PF1/MOD3 computer code and models for computing Mark-22 fuel assembly LOCA/ECS power limits. This qualification effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to independently confirm power limits for the Savannah River Site K Reactor. The results of this benchmark effort as discussed in this paper demonstrate that TRAC/PF1/MOD3 coupled with proper modeling is capable of simulating thermal-hydraulic phenomena typical of that encountered in Mark-22 fuel assembly during LOCA/ECS conditions

TRAC-BDl/MOD1 post-dryout wall heat transfer

International Nuclear Information System (INIS)

Shumway, R.W.

1984-01-01

A comparison of TRAC-BWR heat transfer package with 766 data points is presented. On the average, TRAC-BWR provides a better prediction of the data compared to any single correlation although there is still a large scatter in TRAC-BWR prediction. Regarding any potential changes in the TRAC-BD1/MOD1 wall heat transfer package, it is concluded that no significant improvement in the film boiling area can be made until data with better measurements are obtained and analyzed. Specifically, data is needed which has a wide range of accurately measured void fractions. Heated tube data is also needed which addresses the countercurrent flow transition conditions

Criticality analysis for mixed thorium-uranium fuel in the Angra-2 PWR reactor using KENO-VI

Energy Technology Data Exchange (ETDEWEB)

Wichrowski, Caio C.; Gonçalves, Isadora C.; Oliveira, Claudio L.; Vellozo, Sergio O.; Baptista, Camila O., E-mail: wichrowski@ime.eb.br, E-mail: isadora.goncalves@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

2017-07-01

The increasing energy demand associated to the current sustainability challenges have given the thorium nuclear fuel cycle renewed interest in the scientific community. Studies have focused on energy production in different reactor designs through the fission of uranium 233, the product of thorium fertilization by neutrons. In order to make it possible for near future applications a strategy based on the adaptation of current nuclear reactors for the use of thorium fuels is being considered. In this work, bearing in mind these limitations, a code was used to evaluate the effect on criticality (k{sub inf}) of the mixing of thorium and uranium in different proportions in the fuel of a PWR, the German designed Angra-2 Brazilian reactor in order to scrutinise its behaviour and determine the feasibility of an adapted ThO{sub 2}-UO{sub 2} mixed fuel cycle using current PWR technology. The analysis is performed using the KENO-VI module in the SCALE 6.1 nuclear safety analysis simulation code and the information is taken from the Angra-2 FSAR (Final Security Analysis Report). (author)

Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

International Nuclear Information System (INIS)

Kheshtpaz, H.; Alison, C.

2006-01-01

As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

TRAC Innovative Visualization Techniques

Science.gov (United States)

2016-11-14

5Benjamin Jotham Fry. “Computational information design”. PhD thesis . Massachusetts Institute of Technology, 2004, p. 39. 6Ibid., p. 39. 7Tamara Munzner...central capacity limit to the simultaneous storage of visual and auditory arrays in working memory.” In: Journal of Experimental Psychology : General...GAT scores. 18 August 2016 TRAC Brief Template (AE Method) 2 1Q 2Q 3Q SpiritualService 4Q Rank Gender Spiritual LowVery Low High Initial Model Final

Generalized drift-flux correlation

International Nuclear Information System (INIS)

Takeuchi, K.; Young, M.Y.; Hochreiter, L.E.

1991-01-01

A one-dimensional drift-flux model with five conservation equations is frequently employed in major computer codes, such as TRAC-PD2, and in simulator codes. In this method, the relative velocity between liquid and vapor phases, or slip ratio, is given by correlations, rather than by direct solution of the phasic momentum equations, as in the case of the two-fluid model used in TRAC-PF1. The correlations for churn-turbulent bubbly flow and slug flow regimes were given in terms of drift velocities by Zuber and Findlay. For the annular flow regime, the drift velocity correlations were developed by Ishii et al., using interphasic force balances. Another approach is to define the drift velocity so that flooding and liquid hold-up conditions are properly simulated, as reported here. The generalized correlation is used to reanalyze the MB-2 test data for two-phase flow in a large-diameter pipe. The results are applied to the generalized drift flux velocity, whose relationship to the other correlations is discussed. Finally, the generalized drift flux correlation is implemented in TRAC-PD2. Flow reversal from countercurrent to cocurrent flow is computed in small-diameter U-shaped tubes and is compared with the flooding curve

GPS-based Microenvironment Tracker (MicroTrac) Model to ...

Science.gov (United States)

A critical aspect of air pollution exposure assessment is the estimation of the time spent by individuals in various microenvironments (ME). Accounting for the time spent in different ME with different pollutant concentrations can reduce exposure misclassifications, while failure to do so can add uncertainty and bias to risk estimates. In this study, a classification model, called MicroTrac, was developed to estimate time of day and duration spent in eight ME (indoors and outdoors at home, work, school; inside vehicles; other locations) from global positioning system (GPS) data and geocoded building boundaries. Based on a panel study, MicroTrac estimates were compared to 24 h diary data from 7 participants on workdays and 2 participants on nonworkdays, with corresponding GPS data and building boundaries of home, school, and work. MicroTrac correctly classified the ME for 99.5% of the daily time spent by the participants. The capability of MicroTrac could help to reduce the time-location uncertainty in air pollution exposure models and exposure metrics for individuals in health studies. The National Exposure Research Laboratory’s (NERL’s) Human Exposure and Atmospheric Sciences Division (HEASD) conducts research in support of EPA’s mission to protect human health and the environment. HEASD’s research program supports Goal 1 (Clean Air) and Goal 4 (Healthy People) of EPA’s strategic plan. More specifically, our division conducts research to characterize

Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

Energy Technology Data Exchange (ETDEWEB)

Gomez Garcia-Torano, I.; Jimenez, G.

2013-07-01

The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

Methods for the development of large computer codes under LTSS

International Nuclear Information System (INIS)

Sicilian, J.M.

1977-06-01

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

Development of computer aided engineering system for TRAC applications

International Nuclear Information System (INIS)

Arai, Kenji; Itoya, Seihiro; Uematsu, Hitoshi; Tsunoyama, Shigeaki

1990-01-01

An advanced best estimate computer program for nuclear reactor transient analysis, TRAC has been extensively used to carry out various thermal hydraulic calculations in the nuclear engineering field, because of its versatility. To perform efficiently a wide variety of TRAC calculation, the efficient utilization of computers and the convenient environment for input and output processing is necessary. We have applied a computer network comprising a super-computer, engineering work stations and personal computers to TRAC calculations and have assigned the appropriate functions to each computer. We have also been developing an interactive graphics system for input and output processing on an EWS. This hardware and software environment can improve the effectiveness of TRAC utilization for various thermal hydraulic calculations. (author)

COBRA/TRAC analysis of two-dimensional thermal-hydraulic behavior in SCTF reflood tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Ohnuki, Akira; Sobajima, Makoto; Adachi, Hiromichi

1987-01-01

The effects of radial power distribution and non-uniform upper plenum water accumulation on thermal-hydraulic behavior in the core were observed in the reflood tests with Slab Core Test Facility (SCTF). In order to examine the predictability of these two effects by a multi-dimensional analysis code, the COBRA/TRAC calculations were made. The calculated results indicated that the heat transfer enhancement in high power bundles above quench front was caused by high vapor flow rate in those bundles due to the radial power distribution. On the other hand, the heat transfer degradation in the peripheral bundles under the condition of non-uniform upper plenum water accumulation was caused by the lower flow rates of vapor and entrained liquid above the quench front in those bundles by the reason that vapor concentrated in the center bundles due to the cross flow induced by the horizontal pressure gradient in the core. The above-mentioned two-dimensional heat transfer behaviors calculated with the COBRA/TRAC code is similar to those observed in SCTF tests and therefore those calculations are useful to investigate the mechanism of the two-dimensional effects in SCTF reflood tests. (author)

Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

International Nuclear Information System (INIS)

Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

2016-01-01

To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

Hazardous chemical tracking system (HAZ-TRAC)

International Nuclear Information System (INIS)

Bramlette, J.D.; Ewart, S.M.; Jones, C.E.

1990-07-01

Westinghouse Idaho Nuclear Company, Inc. (WINCO) developed and implemented a computerized hazardous chemical tracking system, referred to as Haz-Trac, for use at the Idaho Chemical Processing Plant (ICPP). Haz-Trac is designed to provide a means to improve the accuracy and reliability of chemical information, which enhances the overall quality and safety of ICPP operations. The system tracks all chemicals and chemical components from the time they enter the ICPP until the chemical changes form, is used, or becomes a waste. The system runs on a Hewlett-Packard (HP) 3000 Series 70 computer. The system is written in COBOL and uses VIEW/3000, TurboIMAGE/DBMS 3000, OMNIDEX, and SPEEDWARE. The HP 3000 may be accessed throughout the ICPP, and from remote locations, using data communication lines. Haz-Trac went into production in October, 1989. Currently, over 1910 chemicals and chemical components are tracked on the system. More than 2500 personnel hours were saved during the first six months of operation. Cost savings have been realized by reducing the time needed to collect and compile reporting information, identifying and disposing of unneeded chemicals, and eliminating duplicate inventories. Haz-Trac maintains information required by the Superfund Amendment Reauthorization Act (SARA), the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) and the Occupational Safety and Health Administration (OSHA)

Hazardous chemical tracking system (HAZ-TRAC)

Energy Technology Data Exchange (ETDEWEB)

Bramlette, J D; Ewart, S M; Jones, C E

1990-07-01

Westinghouse Idaho Nuclear Company, Inc. (WINCO) developed and implemented a computerized hazardous chemical tracking system, referred to as Haz-Trac, for use at the Idaho Chemical Processing Plant (ICPP). Haz-Trac is designed to provide a means to improve the accuracy and reliability of chemical information, which enhances the overall quality and safety of ICPP operations. The system tracks all chemicals and chemical components from the time they enter the ICPP until the chemical changes form, is used, or becomes a waste. The system runs on a Hewlett-Packard (HP) 3000 Series 70 computer. The system is written in COBOL and uses VIEW/3000, TurboIMAGE/DBMS 3000, OMNIDEX, and SPEEDWARE. The HP 3000 may be accessed throughout the ICPP, and from remote locations, using data communication lines. Haz-Trac went into production in October, 1989. Currently, over 1910 chemicals and chemical components are tracked on the system. More than 2500 personnel hours were saved during the first six months of operation. Cost savings have been realized by reducing the time needed to collect and compile reporting information, identifying and disposing of unneeded chemicals, and eliminating duplicate inventories. Haz-Trac maintains information required by the Superfund Amendment Reauthorization Act (SARA), the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) and the Occupational Safety and Health Administration (OSHA).

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Model for transient simulation in a PWR steam circuit

International Nuclear Information System (INIS)

Mello, L.A. de.

1982-11-01

A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

2D/3D Program work summary report, [January 1988--December 1992

International Nuclear Information System (INIS)

Damerell, P.S.; Simons, J.W.

1993-06-01

The 2D/3D Program was carried out by Germany, Japan and the United States to investigate the thermal-hydraulics of a PWR large-break LOCA. A contributory approach was utilized in which each country contributed significant effort to the program and all three countries shared the research results. Germany constructed and operated the Upper Plenum Test Facility (UPTF), and Japan constructed and operated the Cylindrical Core Test Facility (CCTF) and the Slab Core Test Facility (SCTF). The US contribution consisted of provision of advanced instrumentation to each of the three test facilities, and assessment of the TRAC computer code against the test results. Evaluations of the test results were carried out in all three countries. This report summarizes the 2D/3D Program in terms of the contributing efforts of the participants

System calculations related to the accident at Three-Mile Island using TRAC

International Nuclear Information System (INIS)

Ireland, J.R.

1980-01-01

The Three Mile Island nuclear plant (Unit 2) was modeled using the Transient Reactor Analysis Code (TRAC-P1A) and a base case calculation, which simulated the initial part of the accident that occurred on March 28, 1979, was performed. In addition to the base case calculation, several parametric calculations were performed in which a single hypothetical change was made in the system conditions, such as assuming the high pressure injection (HPI) system operated as designed rather than as in the accident. Some of the important system parameter comparisons for the base case as well as some of the parametric case results are presented

Nuclear Plant Analyzer: an interactive TRAC/RELAP Power-Plant Simulation Program

International Nuclear Information System (INIS)

Steinke, R.; Booker, C.; Giguere, P.; Liles, D.; Mahaffy, J.; Turner, M.; Wiley, R.

1984-01-01

The Nuclear Plant Analyzer (NPA) is a computer-software interface for executing the TRAC or RELAP5 power-plant systems codes. The NPA is designed to use advanced supercomputers, long-distance data communications, and a remote workstation terminal with interactive computer graphics to analyze power-plant thermal-hydraulic behavior. The NPA interface simplifies the running of these codes through automated procedures and dialog interaction. User understanding of simulated-plant behavior is enhanced through graphics displays of calculational results. These results are displayed concurrently with the calculation. The user has the capability to override the plant's modeled control system with hardware adjustment commands. This gives the NPA the utility of a simulator, and at the same time, the accuracy of an advanced, best-estimate, power-plant systems code for plant operation and safety analysis

PWR-to-PWR fuel cycle model using dry process

International Nuclear Information System (INIS)

Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

2002-03-01

PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

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

International Nuclear Information System (INIS)

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

1985-01-01

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

TRAC analysis of design basis events for the accelerator production of tritium target/blanket

International Nuclear Information System (INIS)

Lin, J.C.; Elson, J.

1997-01-01

A two-loop primary cooling system with a residual heat removal system was designed to mitigate the heat generated in the tungsten neutron source rods inside the rungs of the ladders and the shell of the rungs. The Transient Reactor Analysis Code (TRAC) was used to analyze the thermal-hydraulic behavior of the primary cooling system during a pump coastdown transient; a cold-leg, large-break loss-of-coolant accident (LBLOCA); a hot-leg LBLOCA; and a target downcomer LBLOCA. The TRAC analysis results showed that the heat generated in the tungsten neutron source rods can be mitigated by the primary cooling system for the pump coastdown transient and all the LBLOCAs except the target downcomer LBLOCA. For the target downcomer LBLOCA, a cavity flood system is required to fill the cavity with water at a level above the large fixed headers

OECD benchmark a of MOX fueled PWR unit cells using SAS2H, triton and mocup

International Nuclear Information System (INIS)

Ganda, F.; Greenspan, A.

2005-01-01

Three code systems are tested by applying them to calculate the OECD PWR MOX unit cell benchmark A. The codes tested are the SAS2H code sequence of the SCALE5 code package using 44 group library, MOCUP (MCNP4C + ORIGEN2), and the new TRITON depletion sequence of SCALE5 using 238 group cross sections generated using CENTRM with continuous energy cross sections. The burnup-dependent k ∞ and actinides concentration calculated by all three code-systems were found to be in good agreement with the OECD benchmark average results. Limited results were calculated also with the WIMS-ANL code package. WIMS-ANL was found to significantly under-predict k ∞ as well as the concentration of Pu 242 , consistently with the predictions of the WIMS-LWR reported by two of the OECD benchmark participants. Additionally, SAS2H is benchmarked against MOCUP for a hydride fuel containing unit cell, giving very satisfactory agreement. (authors)

TRAC-Monterey FY16 Work Program Development and Report of Research Elicitation

Science.gov (United States)

2016-01-01

any changes to priorities or additional projects that require immediate research. Work Program; Research Elicitation Unclassified UU UU UU UU 35 MAJ...conduct analysis for the Army. 1 Marks, Chris, Nesbitt, Peter. TRAC FY14 Research Requirements Elicitation . Technical Report TRAC-M-TM-13-059. 700 Dyer... Requirements Elicitation Interviews Interview Guide: 1. Describe a research requirement in the areas of topics, techniques, and methodologies. 2

Validation of the REL2005 code package on Gd-poisoned PWR type assemblies through the CAMELEON experimental program

International Nuclear Information System (INIS)

Blaise, Patrick; Vidal, Jean-Francois; Santamarina, Alain

2009-01-01

This paper details the validation of Gd-poisoned 17x17 PWR lattices, through several configurations of the CAMELEON experimental program, by using the newly qualified REL2005 French code package. After a general presentation of the CAMELEON program that took place in the EOLE critical Facility in Cadarache, one describes the new REL2005 code package relying on the deterministic transport code APOLLO2.8 based on characteristics method (MOC), and its new CEA2005 library based on the latest JEFF-3.1.1 nuclear data evaluation. For critical masses, the average Calculation-to-Experiment C/E's on the k eff are (136 ± 80) pcm and (300 ± 76) pcm for the reference 281 groups MOC and optimized 26 groups MOC schemes respectively. These values include also a drastic improvement of about 250 pcm due to the change in the library from JEF2.2 to JEFF3.1. For pin-by-pin radial power distributions, reference and REL2005 results are very close, with maximum discrepancies of the order of 2%, i.e., in the experimental uncertainty limits. The Optimized REL2005 code package allows to predict the reactivity worth of the Gd-clusters (averaged on 9 experimental configurations) to be C/E Δρ(Gd clusters) = +1.3% ± 2.3%. (author)

Thermohydraulic calculations of PWR primary circuits

International Nuclear Information System (INIS)

Botelho, D.A.

1984-01-01

Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

MELCOR 1.8.2 assessment: Surry PWR TMLB' (with a DCH study)

International Nuclear Information System (INIS)

Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

1994-02-01

MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified

Uncertainties in modelling and scaling of critical flows and pump model in TRAC-PF1/MOD1

International Nuclear Information System (INIS)

Rohatgi, U.S.; Yu, Wen-Shi.

1987-01-01

The USNRC has established a Code Scalability, Applicability and Uncertainty (CSAU) evaluation methodology to quantify the uncertainty in the prediction of safety parameters by the best estimate codes. These codes can then be applied to evaluate the Emergency Core Cooling System (ECCS). The TRAC-PF1/MOD1 version was selected as the first code to undergo the CSAU analysis for LBLOCA applications. It was established through this methodology that break flow and pump models are among the top ranked models in the code affecting the peak clad temperature (PCT) prediction for LBLOCA. The break flow model bias or discrepancy and the uncertainty were determined by modelling the test section near the break for 12 Marviken tests. It was observed that the TRAC-PF1/MOD1 code consistently underpredicts the break flow rate and that the prediction improved with increasing pipe length (larger L/D). This is true for both subcooled and two-phase critical flows. A pump model was developed from Westinghouse (1/3 scale) data. The data represent the largest available test pump relevant to Westinghouse PWRs. It was then shown through the analysis of CE and CREARE pump data that larger pumps degrade less and also that pumps degrade less at higher pressures. Since the model developed here is based on the 1/3 scale pump and on low pressure data, it is conservative and will overpredict the degradation when applied to PWRs

TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

International Nuclear Information System (INIS)

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

1986-07-01

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients

TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

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

1986-07-01

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

Short-term calculations to supplement the RS 16 B PWR experiments with internals (PWR1 to PWR5), using the LECK 4 computer code

International Nuclear Information System (INIS)

Hughes, G.; Mueller, R.

1980-03-01

Within the framework of research project RS 16 B sponsored by the German BMFT a series of a blowdown experiments, DWR1 to DWR5, were performed using a vessel with dummy internals under conditions similar to those in a PWR. The prime objective of these experiments was the investigation of the highly transient blowdown phenomena in the discharge nozzle and the determination of the induced loads on the internals. As a partner in the project, KWU carried out both pre-test predictions and post-test analyses of these experiments using, among others, the computer code LECK 4. For the most severe blowdown test DWR5, the influence of the most important model parameters on the blowdown analysis was investigated in detail. These investigations suggest that, similar to the long-term analyses, calculations using the homogeneous critical flow model would improve agreement between calculation and experiment. (orig./RW) [de

TRAC Innovation Report (FY15 to FY18)

Science.gov (United States)

2017-11-01

Operations and Reports, 1215 Jefferson Davis Highway, Suite 1204, Arlington, VA 22202-4302, and to the Office of Management and Budget, Paperwork Reduction...inside and outside of TRAC. Rapid, successful adoption of ideas requires leadership and support, but innovation is fundamentally a bottom up phenomena...the organizational environment. TRAC provides a knowledge management system and communities of interest and practice to foster teamwork

Validation of thermohydraulic codes by comparison of experimental results with computer simulations

International Nuclear Information System (INIS)

Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

1989-01-01

The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

ORIGEN-2 libraries based on JENDL-3.2 for PWR-MOX fuel

Energy Technology Data Exchange (ETDEWEB)

Matsumoto, Hideki; Onoue, Masaaki; Tahara, Yoshihisa [Mitsubishi Heavy Industries Ltd., Tokyo (Japan)

2001-08-01

A set of ORIGEN-2 libraries for PWR MOX fuel was developed based on JENDL-3.2 in the Working Group on Evaluation of Nuclide Production, Japanese Nuclear Data Committee. The calculational model generating ORIGEN-2 libraries of PWR MOX is explained here in detail. The ORIGEN-2 calculation with the new ORIGEN-2 MOX library can predict the nuclides contents within 10% for U and Pu isotopes and 20% for both minor actinides and main FPs. (author)

Parameterized representation of macroscopic cross section for PWR reactor

International Nuclear Information System (INIS)

Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

2015-01-01

Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement

Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

Energy Technology Data Exchange (ETDEWEB)

Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2013-10-15

The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

Fuel models and results from the TRAC-PF1/MIMAS TMI-2 accident calculation

International Nuclear Information System (INIS)

Schwegler, E.C.; Maudlin, P.J.

1983-01-01

A brief description of several fuel models used in the TRAC-PF1/MIMAS analysis of the TMI-2 accident is presented, and some of the significant fuel-rod behavior results from this analysis are given. Peak fuel-rod temperatures, oxidation heat production, and embrittlement and failure behavior calculated for the TMI-2 accident are discussed. Other aspects of fuel behavior, such as cladding ballooning and fuel-cladding eutectic formation, were found not to significantly affect the accident progression

Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

1986-01-01

The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

Energy Technology Data Exchange (ETDEWEB)

Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

2014-07-01

In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

Technical Requirements Analysis and Control Systems (TRACS) Initial Operating Capability (IOC) documentation

Science.gov (United States)

Hammond, Dana P.

1991-01-01

The Technical Requirements Analysis and Control Systems (TRACS) software package is described. TRACS offers supplemental tools for the analysis, control, and interchange of project requirements. This package provides the fundamental capability to analyze and control requirements, serves a focal point for project requirements, and integrates a system that supports efficient and consistent operations. TRACS uses relational data base technology (ORACLE) in a stand alone or in a distributed environment that can be used to coordinate the activities required to support a project through its entire life cycle. TRACS uses a set of keyword and mouse driven screens (HyperCard) which imposes adherence through a controlled user interface. The user interface provides an interactive capability to interrogate the data base and to display or print project requirement information. TRACS has a limited report capability, but can be extended with PostScript conventions.

PROF-TRAC D.4.1 Overview and structure of available material

DEFF Research Database (Denmark)

Larsen, Olena Kalyanova; Heiselberg, Per Kvols

This report introduces available educational and training material from the recent IEE projects adapted in the PROF/TRAC repository. An explanation to the categorisation principle of the material in the repository is the key part of this report. The categorisation schema for educational material...... in the repository is developed using the skills and qualifications structure, developed in WP 2 of PROF/TRAC project. In this way the continuity of the work is well established and the achievements from between the work packages are harmonised....

An integrated PWR for marine propulsion

International Nuclear Information System (INIS)

Letouze, A.; Marecaux, A.; Rollason, J.; Heap, S.; Foster, A.; Jewer, S.; Thompson, A. C.; Williams, A. M.; Beeley, P. A.

2008-01-01

Results from a design study for a nuclear propulsion plant utilising a small integrated PWR using many of the inherent safety features of the IRIS design. The design consists of a single pass, low enrichment core housed, together with all associated primary circuit components, within a reactor pressure vessel 10.3 m high and 4.1 m in diameter. Reactor physics calculations were conducted with the codes WIMS9a and MONK8b. The core design contains 21 fuel assemblies each containing 264 UO 2 fuel pins. Each fuel module has a cluster of 24 boron carbide control rods and a central instrumentation channel. The fuel enrichment was 9% in order to achieve the core lifetime requirement of 3000 EFPD at a reactor power of 120 MWth. This gives a discharge burnup of 51,000 MWd/t. To control excess reactivity, two forms of burnable poison are employed: a zirconium dibromide (ZrB 2 ) coating on the fuel compacts, and gadolinium oxide homogeneously mixed in the fuel. Thermal hydraulic calculations were performed using TRAC-P(ND) for steady-state operation and for a number of fault transients. The helical once through steam generators were modelled using heat structure and pipe components and their performance compared to independent calculations including heat transfer correlations for the helical coiled geometry. Intact circuit calculations for steady state were followed by a small break LOCA calculation including the effect of a containment volume which reproduced the gain of coolant effect reported for IRIS. It was demonstrated that the thermal limits were not exceeded for the identified key transients. The dynamic response of the reactor plant to typical power demands was modelled using AcslXtreme software. Several schemes for limiting the power overshoot that was found on rapid increase to full power were examined. It was concluded that the SG must be operated with variable secondary pressure and the best means of reducing power overshoot is to step back the throttle opening

PKL

International Nuclear Information System (INIS)

Probst, P.; Szabo, I.

1985-01-01

This study concerns the simulation of the behavior of a PWR during a LOCA. One studies more particularly the refilling and reflooding phase (PKLI) and the end of the depressurization phase (PKLII). The aim of this test program is to provide the information needed for the validation/verification of computer codes (RELAP4 MOD6, TRAC PF1) and to test an instrumentation realized for the 2D/3D program. A comparison calculation/experiment is presented. 5 refs [fr

Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4

International Nuclear Information System (INIS)

Martinez F, M.A.; Valle G, E. del; Alonso V, G.

2007-01-01

In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO 2 ) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

Data for use in UKAEA PWR plant studies

International Nuclear Information System (INIS)

Kinnersly, S.R.; Richards, C.G.; O'Mahoney, R.

1983-05-01

Plant data represented by the RETRAN, RELAP4 and TRAC models used at Winfrith for studies of pressurised faults and small and large break loss of coolant accidents for the UK PWR are presented together with comparable data for the Sizewell B design taken from the Pre-Construction Safety Report (PCSR). The main components of the plant are described, and modelling issues, which may affect the interpretation and assessment of the data, and the historical development and use of the models, are outlined. The bulk of the report consists of tables of data with supporting figures and text for all the main items of plant modelled in the Winfrith accident studies. The data presented should be adequate to allow assessments of the Winfrith models and results to be carried out and provide a firm basis for the development of models more representative of the Sizewell B PCSR design. (U.K.)

Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

Energy Technology Data Exchange (ETDEWEB)

Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

2015-07-01

Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

Input data preparation and simulation of the second standard problem of IAEA using the Trac/PF1 code

International Nuclear Information System (INIS)

Madeira, A.A.; Pontedeiro, A.C.; Silva Galetti, M.R. da; Borges, R.C.

1989-10-01

The second Standard Problem sponsored by IAEA consists in the simulation of a small LOCA located in the downcomer of a PMK-NVH integral test facility, which models WWER/440 type reactor. This report presents input data preparation and comparison between TRAC-PF1 results and experimental measurements. (author) [pt

Study on thermal-hydraulics during a PWR reflood phase

Energy Technology Data Exchange (ETDEWEB)

Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

A PWR PCI failure criterion to burnups of 60 GW·d/t using the ENIGMA code

International Nuclear Information System (INIS)

Clarke, A.P.; Tempest, P.A.; Shea, J.H.

2000-01-01

A fuel performance modelling code (ENIGMA) has been used to analyse the empirical PCI failure criterion in terms of a clad failure stress as a function of burnup and fast neutron dose. The Studsvik database has been analysed. Results indicate a rising and then saturating failure stress with burnup and fast neutron dose. Using the PCI failure limits, equivalent to 95/95 confidence limits, an ENIGMA stress-based methodology is used to derive PWR PCI failure limits up to 60 GW·d/t U using a conservative assumption that the failure stress does not increase at high burnup and neutron dose. In addition experimental ramp data on gadolinia-doped fuel rods do not indicate any increased susceptibility to PCI failure implying that the UO 2 criterion can be used for gadolinia doped fuel. (author)

Experimental study of effect of initial clad temperature on reflood phenomena during PWR-LOCA

International Nuclear Information System (INIS)

Sugimoto, Jun; Murao, Yoshio

1983-01-01

Integral system tests with the Cylindrical Core Test Facility (CCTF) were performed to investigate the effect of the initial clad temperature on the reflood phenomena in a PWR-LOCA. The initial peak clad temperatures in these three tests were 871, 968 and 1,047K, respectively. The feedback of the system on the core inlet mass flow rate was estimated to be little influenced by the variation of the initial clad temperature except for the first 20s in the transient. The observed temperature rise from the reflood initiation was lower with the higher initial clad temperature. This qualitatively agreed with the results of the small scale forced feed reflood experiments. However, the magnitude of the temperature rise in CCTF was significantly low due to the high initial core inlet mass flow rate. Also observed were the multi-dimensional thermal behaviors for the three cases in the CCTF wide core. The analysis codes REFLA and TRAC reasonably predicted the effect of the initial clad temperature on the core thermo-hydraulics under the simulated core inlet flow conditions. However, the calculated temperature rise of the maximum powered rod based on the one-dimensional core analysis was higher than that of the average powered rod, which contradicts the tendency observed in CCTF tests. (author)

TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor

International Nuclear Information System (INIS)

Martin, R.P.

1993-05-01

Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BF1 computer code. This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with future analysis work at the US Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions

Modelling of coupled self-actuating safety, relief and damped check valve systems with the codes TRAC-PF1 and ROLAST

International Nuclear Information System (INIS)

Neumann, U.; Puzalowski, R.; Grimm, I.

1985-01-01

Numerical valve models for simulation of selfactuating safety valves and damped check valves are introduced for the computer programs TRAC-PF1 and ROLAST. As examples of application post-test calculations and stability analysis are given. (orig.)

2D/3D Program work summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

The 2D/3D Program was carried out by Germany, Japan and the United States to investigate the thermal-hydraulics of a PWR large-break LOCA. A contributory approach was utilized in which each country contributed significant effort to the program and all three countries shared the research results. Germany constructed and operated the Upper Plenum Test Facility (UPTF), and Japan constructed and operated the Cylindrical Core Test Facility (CCTF) and the Slab Core Test Facility (SCTF). The US contribution consisted of provision of advanced instrumentation to each of the three test facilities, and assessment of the TRAC computer code against the test results. Evaluations of the test results were carried out in all three countries. This report summarizes the 2D/3D Program in terms of the contributing efforts of the participants, and was prepared in a coordination among three countries. US and Germany have published the report as NUREG/IA-0126 and GRS-100, respectively. (author).

2D/3D Program work summary report

International Nuclear Information System (INIS)

1995-09-01

The 2D/3D Program was carried out by Germany, Japan and the United States to investigate the thermal-hydraulics of a PWR large-break LOCA. A contributory approach was utilized in which each country contributed significant effort to the program and all three countries shared the research results. Germany constructed and operated the Upper Plenum Test Facility (UPTF), and Japan constructed and operated the Cylindrical Core Test Facility (CCTF) and the Slab Core Test Facility (SCTF). The US contribution consisted of provision of advanced instrumentation to each of the three test facilities, and assessment of the TRAC computer code against the test results. Evaluations of the test results were carried out in all three countries. This report summarizes the 2D/3D Program in terms of the contributing efforts of the participants, and was prepared in a coordination among three countries. US and Germany have published the report as NUREG/IA-0126 and GRS-100, respectively. (author)

TracWorks - global fuel assembly data management

International Nuclear Information System (INIS)

Cooney, B.F.

1997-01-01

The TracWorks Data Management System is a workstation-based software product that provides a utility with a single, broadly available, regularly updated source for virtually every data item available for a fuel assembly or core component. TracWorks is designed to collect, maintain and provide information about assembly and component locations and movements during the refuelling process and operation, assembly burnup and isotopic inventory (both in-core and out-of-core), pin burnup and isotopics for pins that have been removed from their original assemblies, assembly and component inspection results (including video) and manufacturing data provided by the fabrication plant. (UK)

Analysis of the return to power scenario following a LBLOCA in a PWR

Energy Technology Data Exchange (ETDEWEB)

Macian, R.; Tyler, T.N.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

1995-09-01

The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus, the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.

Estimation of PWR spent fuel composition using SCALE and SWAT code systems

Energy Technology Data Exchange (ETDEWEB)

Shin, Hee Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kenya, Suyama; Hiroshi, Okuno [Japan Atomic Energy Research Institute, Tokyo (Japan)

2001-05-01

The isotopic composition calculations were performed for 26 spent fuel samples from Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using SCALE4.4 SAS2H with 27, 44 and 238 group cross-section libraries and SWAT with 107 group cross-section library. For convenience, the ratio of the measured to calculated value was used as a parameter. The four kinds of the calculation results were compared with the measured data. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed the following results. Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from Obrigheim reactor. Larger than unity ratios were found for Am-241 for both the 16 and 55 samples. Larger than unity ratios were found for Sm-149 for the 55 samples. In the case of 26 sample SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor of a system containing PWR spent fuel, taking burnup credit into account.

Burnup calculation methodology in the serpent 2 Monte Carlo code

International Nuclear Information System (INIS)

Leppaenen, J.; Isotalo, A.

2012-01-01

This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)

SWAT2: The improved SWAT code system by incorporating the continuous energy Monte Carlo code MVP

International Nuclear Information System (INIS)

Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi

2003-01-01

SWAT is a code system, which performs the burnup calculation by the combination of the neutronics calculation code, SRAC95 and the one group burnup calculation code, ORIGEN2.1. The SWAT code system can deal with the cell geometry in SRAC95. However, a precise treatment of resonance absorptions by the SRAC95 code using the ultra-fine group cross section library is not directly applicable to two- or three-dimensional geometry models, because of restrictions in SRAC95. To overcome this problem, SWAT2 which newly introduced the continuous energy Monte Carlo code, MVP into SWAT was developed. Thereby, the burnup calculation by the continuous energy in any geometry became possible. Moreover, using the 147 group cross section library called SWAT library, the reactions which are not dealt with by SRAC95 and MVP can be treated. OECD/NEA burnup credit criticality safety benchmark problems Phase-IB (PWR, a single pin cell model) and Phase-IIIB (BWR, fuel assembly model) were calculated as a verification of SWAT2, and the results were compared with the average values of calculation results of burnup calculation code of each organization. Through two benchmark problems, it was confirmed that SWAT2 was applicable to the burnup calculation of the complicated geometry. (author)

Improved coupling of the conduction and flow equations in TRAC

International Nuclear Information System (INIS)

Addessio, F.L.

1981-01-01

Recent nuclear-reactor-systems modeling efforts have been directed toward the development of computer codes capable of simulating transients in short computational times. For this reason, a stability enhancing two-stem method (SETS) has been applied to the two-phase flow equations in the Transient Reactor Analysis Code (TRAC) allowing the Courant limit to be violated. Unfortunately, the coupling between the wall conduction equation and the fluid-dynamics equations is performed semi-implicitly, that is, the wall-heat transfer term, is evaluated using old-time heat-transfer coefficients and wall temperatures and new-time coolant temperatures. This coupling may lead to numerical instabilities at large time steps because of large variations in the heat-transfer coefficient in certain regimes of the boiling curve. Consequently, simply using new-time wall temperatures is not sufficient. A technique that also incorporates new-time heat-transfer coefficients must be used

Modeling of the thermo-mechanical behaviour of the PWR fuel

International Nuclear Information System (INIS)

Mailhe, P.

2014-01-01

This article reviews the various physical phenomena that take place in an irradiated fuel rod and presents the development of the thermo-mechanical codes able to simulate them. Though technically simple the fuel rod is the place where appear 4 types of process: thermal, gas behaviour, mechanical and corrosion that combine involving 5 elements: the fuel pellet, the fuel clad, the fuel-clad gap, the inside volume and the coolant. For instance the pellet is the place where the following mechanical processes took place: thermal dilatation, elastic deformation, creep deformation, densification, solid swelling, gaseous swelling and cracking. The first industrial code simulating the behaviour of the fuel rod was COCCINEL, it was developed by AREVA teams from the American PAD code that was included in the Westinghouse license. Today the GALILEO code has replaced the COPERNIC code that was developed in the beginning of the 2000 years. GALILEO is a synthesis of the state of the art of the different models used in the codes validated for PWR and BWR. GALILEO has been validated on more than 1500 fuel rods concerning PWR, BWR and specific reactors like Siloe, Osiris, HFR, Halden, Studsvik, BR2/3,...) and also for extended burn-ups. (A.C.)

MVP utilization for PWR design code

International Nuclear Information System (INIS)

Matsumoto, Hideki; Tahara, Yoshihisa

2001-01-01

MHI studies the method of the spatially dependent resonance cross sections so as to predict the power distribution in a fuel pellet accurately. For this purpose, the multiband method and the Stoker/Weiss method were implemented to the 2 dimensional transport code PHOENIX-P, and the methods were validated by comparing them with MVP code. Although the appropriate reference was not obtain from the deterministic codes on the resonance cross section study, now the Monte Carlo code MVP result is available and useful as reference. It is shown here how MVP is used to develop the multiband method and the Stoker/Weiss method, and how effective the result of MVP is on the study of the resonance cross sections. (author)

Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

International Nuclear Information System (INIS)

Soares, Alexandre de Souza

2014-01-01

The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

Implementation and assessment of improved models and options in TRAC-BF1

International Nuclear Information System (INIS)

Analytis, G.Th.

1996-07-01

A summary of modifications and options introduced in TRAC-BF1 is presented and it is shown that the predicting capabilities of the modified version of the core are greatly improved. These changes include the introduction of a different heat transfer package during reflooding, the implementation of a simple single-phase limit procedure for forcing the two phases to acquire the same velocity if one phase disappears, a close assessment of the annular flow interfacial shear correlation, implementation of a simple radiation model which seems to alleviate some numerical-oscillations problems induced by the existing highly complex model. Furthermore, different options were introduced and tested like upwinding some terms of the momentum equations (which seems to solve a number of problems reported in the past), the second upwind scheme for the convective terms of the phasic momentum equations and the implementation and assessment of a completely different annular flow interfacial shear correlation. The modified TRAC-BF1 is assessed against some bottom-flooding separate-effect experiments, a 'benchmark' top flooding simulation as well as against the TLTA test Nr. 6423. In the process of this task, the different options are assessed and discussed and is shown that the predictions of the modified code are physically sound and close to the measurements, while almost all the predicted variables are free of unphysical spurious oscillations. The modifications introduced solve a number of problems associated with the frozen version of the code and result in a version which can be confidently used for LB-LOCA analyses. (author) 19 figs., 16 refs

PWR accident management realated tests: some Bethsy results

International Nuclear Information System (INIS)

Clement, P.; Chataing, T.; Deruaz, R.

1993-01-01

The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

Balance of plant modeling in TRAC-BD1/MOD1

International Nuclear Information System (INIS)

Weaver, W.L.; Giles, M.M.; Mohr, C.M.

1983-01-01

The mission of the TRAC-BD1/MOD1 code is to provide a best-estimate analysis capability for Boiling Water Reactor systems and related experimental facilities for the full range of accidents from large and small break Loss-of-Coolant accidents to operational transients including anticipated transients without scram (ATWS), for which point reactor kinetics is adequate (as a first approximation). Recent model developments allow a complete reactor system including the containment and the balance of plant to be modeled. This paper describes the balance of plant models and presents the results of a simulation of a loss-of-feedwater heater transients which was used to assess the performance of the balance of plant models

Analyses of plant behaviors at the secondary side depressurization during LOCA of PWR

Energy Technology Data Exchange (ETDEWEB)

Kawabe, Yasuharu; Tamaki, Tomohiko; Kohriyama, Tamio; Ohtani, Masanori [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

When high pressure injection systems failed during a small break loss-of-coolant-accident (LOCA) for a PWR, main steam relief valves are opened to operate accumulator systems. However, it is pointed out that the core can be exposed since so-called counter current flow limitation (CCFL) occurs in steam generator (SG) tubes. The possibility of the core exposure by CCFL in a PWR plant was evaluated. First, RELAP5/MOD2 code was modified to be able to calculate CCFL. And then the code was applied to evaluate a 4-loop PWR plant. The LOCA with a rupture 3 inches were analyzed with the following two cases: (1) Only the main steam relief valve of the loop with the rupture is opened. (2) all of the relief valves are opened. It is seen that the CCFL phenomenon occurs in the case (1), however, the core cooling was maintained by the accumulator systems that actuated during the core exposure. On the other hand, the core exposure by CCFL is not observed in the case (2). It is shown that core cooling is promoted by operation of main steam relief valves. (author)

Validation of gadolinium burnout using PWR benchmark specification

Energy Technology Data Exchange (ETDEWEB)

Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

2014-07-01

Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

Computational methods and implementation of the 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction

International Nuclear Information System (INIS)

Aragones, J.M.; Ahnert, C.

1995-01-01

New computational methods have been developed in our 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction. They improve the accuracy and efficiency of the coupled neutronic-thermalhydraulic solution and extend its scope to provide, mainly, the calculation of: the fission reaction rates at the incore mini-detectors; the responses at the excore detectors (power range); the temperatures at the thermocouple locations; and the in-vessel distribution of the loop cold-leg inlet coolant conditions in the reflector and core channels, and to the hot-leg outlets per loop. The functional capabilities implemented in the extended SIMTRAN code for online utilization include: online surveillance, incore-excore calibration, evaluation of peak power factors and thermal margins, nominal update and cycle follow, prediction of maneuvers and diagnosis of fast transients and oscillations. The new code has been installed at the Vandellos-II PWR unit in Spain, since the startup of its cycle 7 in mid-June, 1994. The computational implementation has been performed on HP-700 workstations under the HP-UX Unix system, including the machine-man interfaces for online acquisition of measured data and interactive graphical utilization, in C and X11. The agreement of the simulated results with the measured data, during the startup tests and first months of actual operation, is well within the accuracy requirements. The performance and usefulness shown during the testing and demo phase, to be extended along this cycle, has proved that SIMTRAN and the man-machine graphic user interface have the qualities for a fast, accurate, user friendly, reliable, detailed and comprehensive online core surveillance and prediction

Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR); Implementacion en software libre del simulador universitario de nucleoelectrica tipo PWR (SU-PWR)

Energy Technology Data Exchange (ETDEWEB)

Valle H, J.; Hidago H, F.; Morales S, J.B. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

2007-07-01

Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

EDF/CIDEN - ONECTRA: PWR decontamination

International Nuclear Information System (INIS)

Fayolle, P.; Orcel, H.; Wertz, L.

2010-01-01

In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

Partial discharge testing: a progress report. Statistical evaluation of PD data

International Nuclear Information System (INIS)

Warren, V.; Allan, J.

2005-01-01

It has long been known that comparing the partial discharge results obtained from a single machine is a valuable tool enabling companies to observe the gradual deterioration of a machine stator winding and thus plan appropriate maintenance for the machine. In 1998, at the annual Iris Rotating Machines Conference (IRMC), a paper was presented that compared thousands of PD test results to establish the criteria for comparing results from different machines and the expected PD levels. At subsequent annual Iris conferences, using similar analytical procedures, papers were presented that supported the previous criteria and: in 1999, established sensor location as an additional criterion; in 2000, evaluated the effect of insulation type and age on PD activity; in 2001, evaluated the effect of manufacturer on PD activity; in 2002, evaluated the effect of operating pressure for hydrogen-cooled machines; in 2003, evaluated the effect of insulation type and setting Trac alarms; in 2004, re-evaluated the effect of manufacturer on PD activity. Before going further in database analysis procedures, it would be prudent to statistically evaluate the anecdotal evidence observed to date. The goal was to determine which variables of machine conditions greatly influenced the PD results and which didn't. Therefore, this year's paper looks at the impact of operating voltage, machine type and winding type on the test results for air-cooled machines. Because of resource constraints, only data collected through 2003 was used; however, as before, it is still standardized for frequency bandwidth and pruned to include only full-load-hot (FLH) results collected for one sensor on operating machines. All questionable data, or data from off-line testing or unusual machine conditions was excluded, leaving 6824 results. Calibration of on-line PD test results is impractical; therefore, only results obtained using the same method of data collection and noise separation techniques are compared. For

Modification of the ANC Nodal Code for analysis of PWR assembly bow

International Nuclear Information System (INIS)

Franceschini, Fausto; Fetterman, Robert J.; Little, David C.

2008-01-01

Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

Modification of the ANC Nodal Code for analysis of PWR assembly bow

Energy Technology Data Exchange (ETDEWEB)

Franceschini, Fausto; Fetterman, Robert J.; Little, David C. [Westinghouse Electric Company LLC, Pittsburgh PA (United States)

2008-07-01

Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

TracWorksTM-global fuel assembly data management

International Nuclear Information System (INIS)

Cooney, B.F.

1997-01-01

The TracWorks TM Data Management System is a workstation-based software product that provides a utility with a single, broadly available, regularly updated source for virtually every data item available for a fuel assembly or core component. TracWorks is designed to collect, maintain and provide information about assembly and component locations and movements during the refueling process and operation, assembly burnup and isotopic inventory (both in-core and out-of-core), pin burnup and isotopics for pins that have been removed from their original assemblies, assembly and component inspection results (including video) and manufacturing data provided by the fabrication plant

Nuclear Plant Analyzer development at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Laats, E.T.; Beelman, R.J.; Charlton, T.R.; Hampton, N.L.; Burtt, J.D.

1985-01-01

The Nuclear Plant Analyzer (NPA) is a state-of-the-art safety analysis and engineering tool being used to address key nuclear power plant safety issues. The NPA has been developed to integrate the NRC's computerized reactor behavior simulation codes such as RELAP5, TRAC-BWR, and TRAC-PWR, with well-developed computer graphics programs and large repositories of reactor design and experimental data. An important feature of the NAP is the capability to allow an analyst to redirect a RELAP5 or TRAC calculation as it progresses through its simulated scenario. The analyst can have the same power plant control capabilities as the operator of an actual plant. The NPA resides on the dual CDS Cyber-176 mainframe computers at the INEL and is being converted to operate on a Cray-1S computer at the LANL. The subject of this paper is the program conducted at the INEL

The coupling of the Star-Cd software to a whole-core neutron transport code Decart for PWR applications

International Nuclear Information System (INIS)

Thomas, J.W.; Lee, H.C.; Downar, T.J.; Sofu, T.; Weber, D.P.; Joo, H.G.; Cho, J.Y.

2003-01-01

As part of a U.S.- Korea collaborative U.S. Department of Energy INERI project, a comprehensive high-fidelity reactor-core modeling capability is being developed for detailed analysis of existing and advanced PWR reactor designs. An essential element of the project has been the development of an interface between the computational fluid dynamics (CFD) module, STAR-CD, and the neutronics module, DeCART. Since the computational mesh for CFD and neutronics calculations are generally different, the capability to average and decompose data on these different meshes has been an important part of code coupling activities. An averaging process has been developed to extract neutronics zone temperatures in the fuel and coolant and to generate appropriate multi group cross sections and densities. Similar procedures have also been established to map the power distribution from the neutronics zones to the mesh structure used in the CFD module. Since MPI is used as the parallel model in STAR-CD and conflicts arise during initiation of a second level of MPI, the interface developed here is based on using TCP/IP protocol sockets to establish communication between the CFD and neutronics modules. Preliminary coupled calculations have been performed for PWR fuel assembly size problems and converged solutions have been achieved for a series of steady-state problems ranging from a single pin to a 1/8 model of a 17 x 17 PWR fuel assembly. (authors)

SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

Energy Technology Data Exchange (ETDEWEB)

Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

2010-03-01

The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

Experiments on natural circulation during PWR severe accidents and their analysis

International Nuclear Information System (INIS)

Sehgal, B.R.; Stewart, W.A.; Sha, W.T.

1988-01-01

Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3

International Nuclear Information System (INIS)

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

1993-07-01

A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser

A Calculation of the radioactivity induced in PWR cluster control rods with the origin and casmo codes

International Nuclear Information System (INIS)

Ekberg, K.

1980-03-01

The radioactivity induced in PWR cluster control rods during reactor operation has been calculated using the computer programme ORIGEN. Neutron fluxes and spectrum conditions as well as the strongly shielded cross sections for the absorber materials Ag, In and Cd have been obtained by running the cell and assembly code CASMO for a couple of typical cases. The results show that Ag-110m, Fe-55 and Co-60 give the largest activity contributions in the interval 1-10 years after the end of irradiation, and Ni-63 and Cd-113m in a longer time perspective. (author)

Reactivity Impact of Difference of Nuclear Data Library for PWR Fuel Assembly Calculation by Using AEGIS Code

International Nuclear Information System (INIS)

Ohoka, Yasunori; Tatsumi, Masahiro; Sugimura, Naoki; Tabuchi, Masato

2011-01-01

In 2010, the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0) has been released by JAEA. JENDL-4.0 is major update from JENDL- 3.3, and confirmed to give good accuracy by integral test for fission reactor systems such as fast neutron system and thermal neutron system. In this study, we evaluated the reactivity impact due to difference between ENDF/B-VII.0 and JENDL-4.0 for PWR fuel assembly burnup calculation using AEGIS code which has been developed by Nuclear Engineering, Ltd. in cooperation with Nuclear Fuel Industries, Ltd. and Nagoya University

Quantifying reactor safety margins: Application of CSAU [Code Scalability, Applicability and Uncertainty] methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codes

International Nuclear Information System (INIS)

Wulff, W.; Boyack, B.E.; Duffey, R.B.

1988-01-01

Comparisons of results from TRAC-PF1/MOD1 code calculations with measurements from Separate Effects Tests, and published experimental data for modeling parameters have been used to determine the uncertainty ranges of code input and modeling parameters which dominate the uncertainty in predicting the Peak Clad Temperature for a postulated Large Break Loss of Coolant Accident (LBLOCA) in a four-loop Westinghouse Pressurized Water Reactor. The uncertainty ranges are used for a detailed statistical analysis to calculate the probability distribution function for the TRAC code-predicted Peak Clad Temperature, as is described in an attendant paper. Measurements from Separate Effects Tests and Integral Effects Tests have been compared with results from corresponding TRAC-PF1/MOD1 code calculations to determine globally the total uncertainty in predicting the Peak Clad Temperature for LBLOCAs. This determination is in support of the detailed statistical analysis mentioned above. The analyses presented here account for uncertainties in input parameters, in modeling and scaling, in computing and in measurements. The analyses are an important part of the work needed to implement the Code Scalability, Applicability and Uncertainty (CSAU) methodology. CSAU is needed to determine the suitability of a computer code for reactor safety analyses and the uncertainty in computer predictions. The results presented here are used to estimate the safety margin of a particular nuclear reactor power plant for a postulated accident. 25 refs., 10 figs., 11 tabs

Source terms associated with two severe accident sequences in a 900 MWe PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Berthion, Y.; Lhiaubet, G.; Lucas, M.

1983-12-01

Hypothetical accidents taken into account in PWR risk assessment result in fission product release from the fuel, transfer through the primary circuit, transfer into the reactor containment building (RCB) and finally release to the environment. The objective of this paper is to define the characteristics of the source term (noble gases, particles and volatile iodine forms) released from the reactor containment building during two dominant core-melt accident sequences: S 2 CD and TLB according to the ''Reactor Safety Study'' terminology. The reactor chosen for this study is a French 900 MWe PWR unit. The reactor building is a prestressed concrete containment with an internal liner. The first core-melt accident sequence is a 2-break loss-of-coolant accident on the cold leg, with failure of both system and the containment spray system. The second one is a transient initiated by a loss of offsite and onsite power supply and auxiliary feedwater system. These two sequences have been chosen because they are representative of risk dominant scenarios. Source terms associated with hypothetical core-melt accidents S 2 CD and TLB in a French PWR -900 MWe- have been performed using French computer codes (in particular, JERICHO Code for containment response analysis and AEROSOLS/31 for aerosol behavior in the containment)

Modeling of hydrogen behaviour in a PWR nuclear power plant containment with the CONTAIN code

International Nuclear Information System (INIS)

Bobovnik, G.; Kljenak, I.

2001-01-01

Hydrogen behavior in the containment during a severe accident in a two-loop Westinghouse-type PWR nuclear power plant was simulated with the CONTAIN code. The accident was initiated with a cold-leg break of the reactor coolant system in a steam generator compartment. In the input model, the containment is represented with 34 cells. Beside hydrogen concentration, the containment atmosphere temperature and pressure and the carbon monoxide concentration were observed as well. Simulations were carried out for two different scenarios: with and without successful actuation of the containment spray system. The highest hydrogen concentration occurs in the containment dome and near the hydrogen release location in the early stages of the accident. Containment sprays do not have a significant effect on hydrogen stratification.(author)

Use of the APOLLO2 transport code for PWR assembly studies

International Nuclear Information System (INIS)

Belhaffaf, D.; Coste, M.; Lenain, R.; Mathonniere, G.; Sanchez, R.; Stankovski, Z.; Zmijarevic, I.

1992-01-01

This paper presents some validation and application aspects of the APOLLO2, user oriented, modular code for multigroup transport assembly calculation which is developed at the French Commissariat a l'Energies Atomique. The main points approached in this paper are: the two dimensional collision probability convergence, critical leakage calculation schemes, self-shielding spatial discretization, and the equivalence procedure

Comparison of a TRAC calculation to the data from LSTF run SB-CL-05

International Nuclear Information System (INIS)

Motley, F.; Schultz, R.

1986-01-01

Run SB-CL-05 is a 5% break in the side of the cold leg. The test results show that the core was uncovered briefly and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam generator tubes was observed. When the loop seal cleared, the core refilled and the rods cooled. The TRAC results are in reasonable agreement with the test data, meaning that TRAC correctly predicted the major trends and phenomena. TRAC predicted the core uncovery, the resulting rod heatup, and the liquid holdup on the upflow side of the steam generator tubes correctly. The clearing of the loop seal allowed core recovery and cooled the overheated rods just as it had in the data, but TRAC predicted its occurrence 20 s late. The experimental and TRAC analysis results of run SB-CL-05 are similar to those for Semiscale Run S-UT-8. In both runs there was core uncovery, rod overheating, and steam generator liquid holdup. These results confirm scaling of these phenomena from Semiscale (1/1650) to LSTF (1/48)

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Feasibility to convert an advanced PWR from UO2 to a mixed U/ThO2 core – Part I: Parametric studies

International Nuclear Information System (INIS)

Maiorino, Jose R.; Stefani, Giovanni Laranjo; Moreira, João M.L.; Rossi, Pedro C.R.; Santos, Thiago A.

2017-01-01

Highlights: • Neutronics calculation using SERPENT code. • Conversion of an advanced PWR from a UO 2 to (U-Th)O 2 core. • AP 1000-advanced PWR. • Parametric studies to define a converted core. • Demonstration of the feasibility to convert the AP 1000 by using mixed uranium thorium oxide fuel with advantages. - Abstract: This work presents the neutronics and thermal hydraulics feasibility to convert the UO 2 core of the Westinghouse AP1000 in a (U-Th)O 2 core by performing a parametric study varying the type of geometry of the pins in fuel elements, using the heterogeneous seed blanket concept and the homogeneous concept. In the parametric study, all geometry and materials for the burnable poison were kept the same as the AP 1000, and the only variable was the fuel pin material, in which we use several mass proportion of uranium and thorium but keeping the enrichment in 235 U, as LEU (20 w/o). The neutronics calculations were made by SERPENT code, and to validate the thermal limits we used a homemade code. The optimization criteria were to maximize the 233 U, and conversion factor, and minimize the plutonium production. The results obtained showed that the homogeneous concept with three different mass proportion zones, the first containing (32% UO 2 -68%ThO 2 ); the second with (24% UO 2 -76% ThO 2 ), and the third with (20% UO 2 -80% ThO 2 ), using 235 U LEU (20 w/o), and corresponding with the 3 enrichment zones of the AP 1000 (4.45 w/o; 3.40 w/o; 2.35 w/o), satisfies the optimization criteria as well as attending all thermal constrain. The concept showed advantages compared with the original UO 2 core, such a lower power density, and keeping the same 18 months of cycle a reduction of B-10 concentration at the soluble poison as well as eliminating in the integral boron poison coated (IFBA).

The research on burnup characteristic of doping burnable poison in PWR

International Nuclear Information System (INIS)

Qiang Shenglong; Qin Dong; Chai Xiaoming; Yao Dong

2014-01-01

In PWR core design, burnable poisons are usually used for reactive compensation and power flatten. The choice of burnable poisons and how to match burnup would be the key-points for a long-life core design. We study the burnup character of doping burnable poisons (such as natural element, manual nuclide and soluble boron) in the PWR by the core burnup code MOI based on Monte Carlo method. The results show that Hf, Er and Eu doping burnable poison would be applicable for the nuclear design research on the long-life PWR core. (authors)

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

International Nuclear Information System (INIS)

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

1987-11-01

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

A comparison of the effect of the first and second upwind schemes on the predictions of the modified RELAP5/MOD3

Energy Technology Data Exchange (ETDEWEB)

Analytis, G.Th. [Paul Scherrer Institute (PSI), Villigen (Switzerland)

1995-09-01

As is well-known, both TRAC-BF1 and TRAC-PF are using the first upwind scheme when finite-differencing the phasic momentum equations. In contrast, RELAP5 uses the second upwind which is less diffusive. In this work, we shall assess the differences between the two schemes with our modified version of RELAP5/MOD3 by analyzing some transients of interest. These will include the LOFT LP-LB-1 and LOBI small break LOCA (SB-LOCA) BL34 tests, and a commercial PWR 200% hypothetical large break LOCA (LB-LOCA). In particular, we shall show that for some of these transients, the employment of the first upwind scheme results in significantly different code predictions than the ones obtained when the second upwind scheme is used.

Nuclear Plant Analyzer development at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Laats, E.T.

1986-10-01

The Nuclear Plant Analyzer (NPA) is a state-of-the-art safety analysis and engineering tool being used to address key nuclear power plant safety issues. Under the sponsorship of the US Nuclear Regulatory Commission (NRC), the NPA has been developed to integrate the NRC's computerized reactor behavior simulation codes such as RELAP5, TRAC-BWR and TRAC-PWR, with well-developed computer color graphics programs and large repositories of reactor design and experimental data. An important feature of the NPA is the capability to allow an analyst to redirect a RELAP5 or TRAC calculation as it progresses through its simulated scenario. The analyst can have the same power plant control capabilities as the operator of an actual plant. The NPA resides on the dual Control Data Corporation Cyber 176 mainframe computers at the Idaho National Engineering Laboratory and Cray-1S computers at the Los Alamos National Laboratory (LANL) and Kirtland Air Force Weapons Laboratory (KAFWL)

Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

International Nuclear Information System (INIS)

De Rosa, Felice

2006-01-01

In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when ΔTsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

Poster - 48: Clinical assessment of ExacTrac stereoscopic imaging of spine alignment for lung SBRT

International Nuclear Information System (INIS)

Sattarivand, Mike; Summers, Clare; Robar, James

2016-01-01

Purpose: To evaluate the validity of using spine as a surrogate for tumor positioning with ExacTrac stereoscopic imaging in lung stereotactic body radiation therapy (SBRT). Methods: Using the Novalis ExacTrac x-ray system, 39 lung SBRT patients (182 treatments) were aligned before treatment with 6 degrees (6D) of freedom couch (3 translations, 3 rotations) based on spine matching on stereoscopic images. The couch was shifted to treatment isocenter and pre-treatment CBCT was performed based on a soft tissue match around tumor volume. The CBCT data were used to measure residual errors following ExacTrac alignment. The thresholds for re-aligning the patients based on CBCT were 3mm shift or 3° rotation (in any 6D). In order to evaluate the effect of tumor location on residual errors, correlations between tumor distance from spine and individual residual errors were calculated. Results: Residual errors were up to 0.5±2.4mm. Using 3mm/3° thresholds, 80/182 (44%) of the treatments required re-alignment based on CBCT soft tissue matching following ExacTrac spine alignment. Most mismatches were in sup-inf, ant-post, and roll directions which had larger standard deviations. No correlation was found between tumor distance from spine and individual residual errors. Conclusion: ExacTrac stereoscopic imaging offers a quick pre-treatment patient alignment. However, bone matching based on spine is not reliable for aligning lung SBRT patients who require soft tissue image registration from CBCT. Spine can be a poor surrogate for lung SBRT patient alignment even for proximal tumor volumes.

Poster - 48: Clinical assessment of ExacTrac stereoscopic imaging of spine alignment for lung SBRT

Energy Technology Data Exchange (ETDEWEB)

Sattarivand, Mike; Summers, Clare; Robar, James [Nova Scotia Cancer Centre, Nova Scotia Cancer Centre, Nova Scotia Cancer Centre (Canada)

2016-08-15

Purpose: To evaluate the validity of using spine as a surrogate for tumor positioning with ExacTrac stereoscopic imaging in lung stereotactic body radiation therapy (SBRT). Methods: Using the Novalis ExacTrac x-ray system, 39 lung SBRT patients (182 treatments) were aligned before treatment with 6 degrees (6D) of freedom couch (3 translations, 3 rotations) based on spine matching on stereoscopic images. The couch was shifted to treatment isocenter and pre-treatment CBCT was performed based on a soft tissue match around tumor volume. The CBCT data were used to measure residual errors following ExacTrac alignment. The thresholds for re-aligning the patients based on CBCT were 3mm shift or 3° rotation (in any 6D). In order to evaluate the effect of tumor location on residual errors, correlations between tumor distance from spine and individual residual errors were calculated. Results: Residual errors were up to 0.5±2.4mm. Using 3mm/3° thresholds, 80/182 (44%) of the treatments required re-alignment based on CBCT soft tissue matching following ExacTrac spine alignment. Most mismatches were in sup-inf, ant-post, and roll directions which had larger standard deviations. No correlation was found between tumor distance from spine and individual residual errors. Conclusion: ExacTrac stereoscopic imaging offers a quick pre-treatment patient alignment. However, bone matching based on spine is not reliable for aligning lung SBRT patients who require soft tissue image registration from CBCT. Spine can be a poor surrogate for lung SBRT patient alignment even for proximal tumor volumes.

SU-E-J-34: Setup Accuracy in Spine SBRT Using CBCT 6D Image Guidance in Comparison with 6D ExacTrac

Energy Technology Data Exchange (ETDEWEB)

Han, Z; Yip, S; Lewis, J; Mannarino, E; Friesen, S; Wagar, M; Hacker, F [Brigham and Women’s Hospital and Dana-Farber Cancer Institute, Boston, MA (United States)

2015-06-15

Purpose Volumetric information of the spine captured on CBCT can potentially improve the accuracy in spine SBRT setup that has been commonly performed through 2D radiographs. This work evaluates the setup accuracy in spine SBRT using 6D CBCT image guidance that recently became available on Varian systems. Methods ExacTrac radiographs have been commonly used for Spine SBRT setup. The setup process involves first positioning patients with lasers followed by localization imaging, registration, and repositioning. Verification images are then taken providing the residual errors (ExacTracRE) before beam on. CBCT verification is also acquired in our institute. The availability of both ExacTrac and CBCT verifications allows a comparison study. 41 verification CBCT of 16 patients were retrospectively registered with the planning CT enabling 6D corrections, giving CBCT residual errors (CBCTRE) which were compared with ExacTracRE. Results The RMS discrepancies between CBCTRE and ExacTracRE are 1.70mm, 1.66mm, 1.56mm in vertical, longitudinal and lateral directions and 0.27°, 0.49°, 0.35° in yaw, roll and pitch respectively. The corresponding mean discrepancies (and standard deviation) are 0.62mm (1.60mm), 0.00mm (1.68mm), −0.80mm (1.36mm) and 0.05° (0.58°), 0.11° (0.48°), −0.16° (0.32°). Of the 41 CBCT, 17 had high-Z surgical implants. No significant difference in ExacTrac-to-CBCT discrepancy was observed between patients with and without the implants. Conclusion Multiple factors can contribute to the discrepancies between CBCT and ExacTrac: 1) the imaging iso-centers of the two systems, while calibrated to coincide, can be different; 2) the ROI used for registration can be different especially if ribs were included in ExacTrac images; 3) small patient motion can occur between the two verification image acquisitions; 4) the algorithms can be different between CBCT (volumetric) and ExacTrac (radiographic) registrations.

Evaluation of Computational Fluids Dynamics (CFD) code Open FOAM in the study of the pressurized thermal stress of PWR reactors. Comparison with the commercial code Ansys-CFX; Evaluacion del codigo de Dinamica de Fluidos Computacional (CFD) Open FOAM en el estudio del estres termico presurizado de los reactores PWR. Comparacion con el codigo comercial Ansys-CFX

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

Transient performance of flow in PWR reactor circuits

International Nuclear Information System (INIS)

Hirdes, V.R.T.R.; Carajilescov, P.

1988-12-01

Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

Contribution to the study of the conversion PWR type reactors to the thorium cycle

International Nuclear Information System (INIS)

Martins Filho, J.R.

1980-01-01

The use of the thorium cycle in PWR reactors is discussed. The fuel has been calculated in the equilibrium condition for a economic comparison with the uranium cycle (in the same condition). First of all, a code named EQUILIBRIO has been developed for the fuel equilibrium calculation. The results gotten by this code, were introduced in the LEOPARD code for the fuel depletion calculation (in the equilibrium cycle). Same important physics details of fuel depletion are studied, for instance: the neutron balance, power sharing, fuel burnup, etc. The calculations have been done taking as reference the Angra-1 PWR reactor. (Author) [pt

Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

Energy Technology Data Exchange (ETDEWEB)

Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

2016-08-01

To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

The CORSYS neutronics code system

International Nuclear Information System (INIS)

Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

1994-01-01

The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

International Nuclear Information System (INIS)

Valle H, J.; Hidago H, F.; Morales S, J.B.

2007-01-01

Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

International Nuclear Information System (INIS)

Papukchiev, Angel; Schaefer, Anselm

2008-01-01

In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

RCC-C: Design and construction rules for fuel assemblies of PWR nuclear power plants

International Nuclear Information System (INIS)

2015-01-01

The RCC-C code contains all the requirements for the design, fabrication and inspection of nuclear fuel assemblies and the different types of core components (rod cluster control assemblies, burnable poison rod assemblies, primary and secondary source assemblies and thimble plug assemblies). The design, fabrication and inspection rules defined in RCC-C leverage the results of the research and development work pioneered in France, Europe and worldwide, and which have been successfully used by industry to design and build nuclear fuel assemblies and incorporate the resulting feedback. The code's scope covers: fuel system design, especially for assemblies, the fuel rod and associated core components, the characteristics to be checked for products and parts, fabrication methods and associated inspection methods. The RCC-C code is used by the operator of the PWR nuclear power plants in France as a reference when sourcing fuel from the world's top two suppliers in the PWR market, given that the French operator is the world's largest buyer of PWR fuel. Fuel for EPR projects is manufactured according to the provisions of the RCC-C code. The code is available in French and English. The 2005 edition has been translated into Chinese. Contents of the 2015 edition of the RCC-C code: Chapter 1 - General provisions: 1.1 Purpose of the RCC-C, 1.2 Definitions, 1.3 Applicable standards, 1.4 Equipment subject to the RCC-C, 1.5 Management system, 1.6 Processing of non-conformances; Chapter 2 - Description of the equipment subject to the RCC-C: 2.1 Fuel assembly, 2.2 Core components; Chapter 3 - Design: Safety functions, operating functions and environment of fuel assemblies and core components, design and safety principles; Chapter 4 - Manufacturing: 4.1 Materials and part characteristics, 4.2 Assembly requirements, 4.3 Manufacturing and inspection processes, 4.4 Inspection methods, 4.5 Certification of NDT inspectors, 4.6 Characteristics to be inspected for the

TOOKUIL: A case study in user interface development for safety code application

International Nuclear Information System (INIS)

Gray, D.L.; Harkins, C.K.; Hoole, J.G.

1997-01-01

Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today's safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interface named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL

TOOKUIL: A case study in user interface development for safety code application

International Nuclear Information System (INIS)

Gray, D.L.; Harkins, C.K.; Hoole, J.G.; Peebles, R.C.; Smith, R.J.

1996-11-01

Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today's safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interface named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL

Analysis of reactivity worths of highly-burnt PWR fuel samples measured in LWR-PROTEUS Phase II

Energy Technology Data Exchange (ETDEWEB)

Grimm, Peter; Murphy, Michael F.; Jatuff, Fabian; Seiler, Rudolf [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland)

2008-07-01

The reactivity loss of PWR fuel with burnup has been determined experimentally by inserting fresh and highly-burnt fuel samples in a PWR test lattice in the framework of the LWR-PROTEUS Phase II programme. Seven UO{sub 2} samples irradiated in a Swiss PWR plant with burnups ranging from approx40 to approx120 MWd/kg and four MOX samples with burnups up to approx70 MWd/kg were oscillated in a test region constituted of actual PWR UO{sub 2} fuel rods in the centre of the PROTEUS zero-power experimental facility. The measurements were analyzed using the CASMO-4E fuel assembly code and a cross section library based on the ENDF/B-VI evaluation. The results show close proximity between calculated and measured reactivity effects and no trend for a deterioration of the quality of the prediction at high burnup. The analysis thus demonstrates the high accuracy of the calculation of the reactivity of highly-burnt fuel. (authors)

Status of SPACE Safety Analysis Code Development

International Nuclear Information System (INIS)

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

2009-01-01

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

A PWR Thorium Pin Cell Burnup Benchmark

Energy Technology Data Exchange (ETDEWEB)

Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

2000-05-01

As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

Thermal-hydraulic analysis best-estimate of an accident in the containment a PWR-W reactor with GOTHIC code using a 3D model detailed; Analisis termo-hidraulico best-estimate de un accidente en contencion de un reactor PWR-W con el codigo GOTHIC mediante un modelo 3D detallado

Energy Technology Data Exchange (ETDEWEB)

Bocanegra, R.; Jimenez, G.

2013-07-01

The objective of this project will be a model of containment PWR-W with the GOTHIC code that allows analyzing the behavior detailed after a design basis accident or a severe accident. Unlike the models normally used in codes of this type, the analysis will take place using a three-dimensional model of the containment, being this much more accurate.

International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

International Nuclear Information System (INIS)

1996-01-01

An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

Energy Technology Data Exchange (ETDEWEB)

Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

2015-12-15

A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

Reactor safety issues resolved by the 2D/3D Program

International Nuclear Information System (INIS)

Damerell, P.S.; Simons, J.W.

1993-07-01

The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated

Polynomial parameterized representation of macroscopic cross section for PWR reactor

International Nuclear Information System (INIS)

Fiel, Joao Claudio B.

2015-01-01

The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 U 92 enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K inf , generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)

78 FR 24313 - Proposed Collection; Comment Request for Tip Reporting Alternative Commitment Agreement (TRAC...

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2013-04-24

... Reporting Alternative Commitment Agreement (TRAC) For Use in Industries Other Than the Food and Beverage... Use in Industries Other than the Food and Beverage Industry and The Cosmetology and Barber Industry... Commitment Agreement (TRAC) For Use in Industries Other than the Food and Beverage Industry and The...

75 FR 11225 - Proposed Collection; Comment Request for Tip Reporting Alternative Commitment Agreement (TRAC...

Science.gov (United States)

2010-03-10

... Reporting Alternative Commitment Agreement (TRAC) for Use in Industries Other Than the Food and Beverage... Use in Industries Other Than the Food and Beverage Industry and the Cosmetology and Barber Industry... Alternative Commitment Agreement (TRAC) for Use in Industries Other Than the Food and Beverage Industry and...

3-D full core calculations for the long-term behaviour of PWR's

International Nuclear Information System (INIS)

Winter, H.J.; Koebke, K.; Wagner, M.R.

1987-01-01

Presently, the most realistic simulation of a PWR core is by means of three-dimensional (3-D) full core calculations. Only by such 3-D representations can the large scope of axial effects be treated in an accurate and direct way, without the need to perform various auxiliary calculations. Although the computationally efficient burnup-corrected nodal expansion method (NEM-BC) is used, the computing effort for 3-D reactor calculations becomes rather high, e.g. a storage of about 320000 words is required to describe a 1300 MWe PWR. NEM-BC was introduced (1979) into KWU's package of PWR design codes because of its high accuracy and the great reduction of computing time and storage requirements in comparison to other methods. The application of NEM-BC to 3-dimensional PWR design is strongly correlated with the progress achieved in the solution of the homogenization and dehomogenization problem. By means of suitable methods (equivalence theory) the transport-theoretical information of the pinwise power and burnup distribution for the heterogeneous fuel assemblies is transferred in a consistent manner to the full core reactor solution. The new methods and the corresponding code system are explained in some detail. (orig.)

SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

International Nuclear Information System (INIS)

Kodaira, Hideki

1990-01-01

1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

Three-Dimensional (X,Y,Z) Deterministic Analysis of the PCA-Replica Neutron Shielding Benchmark Experiment using the TORT-3.2 Code and Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

OpenAIRE

Pescarini Massimo; Orsi Roberto; Frisoni Manuela

2016-01-01

The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with ...

The latest full-scale PWR simulator in Japan

International Nuclear Information System (INIS)

Nishimuru, Y.; Tagi, H.; Nakabayashi, T.

2004-01-01

The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

PHEDRE model for the simulation of PWR reactors

International Nuclear Information System (INIS)

Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

1979-11-01

This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

75 FR 11226 - Proposed Collection; Comment Request for Tip Reporting Alternative Commitment Agreement (TRAC...

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2010-03-10

... Reporting Alternative Commitment Agreement (TRAC) for Use in the Food and Beverage Industry AGENCY: Internal... Reporting Alternative Commitment Agreement (TRAC) for Use in the Food and Beverage Industry. DATES: Written... in the Food and Beverage Industry. OMB Number: 1545-1549. Abstract: Announcement 2000-22, 2000-19 I.R...

Simulations of corrosion product transfer with the OSCAR V1.2 code

International Nuclear Information System (INIS)

Dacquait, F.; Francescatto, J.; Broutin, F.; Genin, J.B.; Benier, G.; Girard, M.; You, D.; Ranchoux, G.; Bonnefon, J.; Bachet, M.; Riot, G.

2012-09-01

Activated Corrosion Products (ACPs) generate a radiation field in PWRs, which is the major contributor to the dose absorbed by nuclear power plant staff working during shutdown operations and maintenance. Therefore, a thorough understanding of the mechanisms that control the corrosion product transfer is of the highest importance. Since the 1970's, the R and D strategy in France has been based on experiments in test loops representative of PWR conditions, on in-situ gamma spectrometry measurements of the PWR primary system contamination and on simulation code development. The simulation of corrosion product transfers in PWR primary circuits is a major challenge since it involves many physical and chemical phenomena including: corrosion, dissolution, precipitation, erosion, deposition, convection, activation... In addition to the intrinsic difficulty of multi-physics modelling, the primary systems present severe operating conditions (300 deg. C, 150 bar, neutron flux, fluid velocity up to 15 m.s -1 and very low corrosion product concentrations). The purpose of the OSCAR code, developed by the CEA in cooperation with EDF and AREVA NP, is to predict the PWR primary system contamination by corrosion and fission products. The OSCAR code is considered to be not only a tool for numerical simulations and predictions (operational practices improvements and new-built PWRs design) but also one that might combine and organise all new knowledge useful to progress on contamination. The OSCAR code for Products of Corrosion, OSCAR PC, allows researchers to analyse the corrosion product behaviour and to calculate the ACP volume and surface activities of the primary and auxiliary systems. In the new version, OSCAR PC V1.2, the corrosion product transfer in the particulate form is enhanced and a new feature is the possibility to simulate cold shutdowns. In order to validate this version, the contamination transfer has been simulated in 5 French PWRs with different operating and

Aqueous phase hydrogenation of phenol catalyzed by Pd and PdAg on ZrO 2

Energy Technology Data Exchange (ETDEWEB)

Resende, Karen A.; Hori, Carla E.; Noronha, Fabio B.; Shi, Hui; Gutierrez, Oliver Y.; Camaioni, Donald M.; Lercher, Johannes A.

2017-11-01

Hydrogenation of phenol in aqueous phase was studied over a series of ZrO2-supported Pd catalysts in order to explore the effects of particle size and of Ag addition on the activity of Pd. Kinetic assessments were performed in a batch reactor, on monometallic Pd/ZrO2 samples with different Pd loadings (0.5%, 1% and 2%), as well as on a 1% PdAg/ZrO2 sample. The turnover frequency (TOF) increases with the Pd particle size. The reaction orders in phenol and H2 indicate that the surface coverages by phenol, H2 and their derived intermediates are higher on 0.5% Pd/ZrO2 than on other samples. The activation energy was the lowest on the least active sample (0.5% Pd/ZrO2), while being identical on 1% and 2% Pd/ZrO2 catalysts. Thus, the significantly lower activity of the small Pd particles (1-2 nm on average) in 0.5%Pd/ZrO2 is explained by the unfavorable activation entropies for the strongly bound species. The presence of Ag increases considerably the TOF of the reaction by decreasing the Ea and increasing the coverages of phenol and H2.

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

International Nuclear Information System (INIS)

Megahed, M.M.

1992-01-01

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

Fluid-structure coupled dynamic response of PWR core barrel during LOCA

International Nuclear Information System (INIS)

Lu, M.W.; Zhang, Y.G.; Shi, F.

1991-01-01

This paper is engaged in the Fluid-Structure Interaction LOCA analysis of the core barrel of PWR. The analysis is performed by a multipurpose computer code SANES. The FSI inside the pressure vessel is treated by a FEM code including some structural and acoustic elements. The transient in the primary loop is solved by a two-phase flow code. Both codes are coupled one another. Some interesting conclusions are drawn. (author)

SU-E-T-19: A New End-To-End Test Method for ExacTrac for Radiation and Plan Isocenter Congruence

Energy Technology Data Exchange (ETDEWEB)

Lee, S; Nguyen, N; Liu, F; Huang, Y [Rhode Island Hospital / Warren Alpert Medical, Providence, RI (United States); Sio, T [Mayo Clinic, Rochester, MN (United States); Jung, J [East Carolina University, Greenville, North Carolina (United States); Pyakuryal, A [UniversityIllinois at Chicago, Chicago, IL (United States); Jang, S [Princeton Radiation Oncology Ctr., Jamesburg, NJ (United States)

2014-06-01

Purpose: To combine and integrate quality assurance (QA) of target localization and radiation isocenter End to End (E2E) test of BrainLAB ExacTrac system, a new QA approach was devised using anthropomorphic head and neck phantom. This test insures the target localization as well as radiation isocenter congruence which is one step ahead the current ExacTrac QA procedures. Methods: The head and neck phantom typically used for CyberKnife E2E test was irradiated to the sphere target that was visible in CT-sim images. The CT-sim was performed using 1 mm thickness slice with helical scanning technique. The size of the sphere was 3-cm diameter and contoured as a target volume using iPlan V.4.5.2. A conformal arc plan was generated using MLC-based with 7 fields, and five of them were include couch rotations. The prescription dose was 5 Gy and 95% coverage to the target volume. For the irradiation, two Gafchromic films were perpendicularly inserted into the cube that hold sphere inside. The linac used for the irradiation was TrueBeam STx equipped with HD120 MLC. In order to use ExacTrac, infra-red head–array was used to correlate orthogonal X-ray images. Results: Using orthogonal X-rays of ExacTrac the phantom was positioned. For each field, phantom was check again with X-rays and re-positioned if necessary. After each setup using ExacTrac, the target was irradiated. The films were analyzed to determine the deviation of the radiation isocenter in all three dimensions: superior-inferior, left-right and anterior-posterior. The total combining error was found to be 0.76 mm ± 0.05 mm which was within sub-millimeter accuracy. Conclusion: Until now, E2E test for ExacTrac was separately implemented to test image localization and radiation isocenter. This new method can be used for periodic QA procedures.

TOOKUIL: A case study in user interface development for safety code application

Energy Technology Data Exchange (ETDEWEB)

Gray, D.L.; Harkins, C.K.; Hoole, J.G. [and others

1997-07-01

Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today`s safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interface named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL.

Fuel management and core design code systems for pressurized water reactor neutronic calculations

International Nuclear Information System (INIS)

Ahnert, C.; Arayones, J.M.

1985-01-01

A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

POST: a postprocessor computer code for producing three-dimensional movies of two-phase flow in a reactor vessel

International Nuclear Information System (INIS)

Taggart, K.A.; Liles, D.R.

1977-08-01

The development of the TRAC computer code for analysis of LOCAs in light-water reactors involves the use of a three-dimensional (r-theta-z), two-fluid hydrodynamics model to describe the two-phase flow of steam and water through the reactor vessel. One of the major problems involved in interpreting results from this code is the presentation of three-dimensional flow patterns. The purpose of the report is to present a partial solution to this data display problem. A first version of a code which produces three-dimensional movies of flow in the reactor vessel has been written and debugged. This code (POST) is used as a postprocessor in conjunction with a stand alone three-dimensional two-phase hydrodynamics code (CYLTF) which is a test bed for the three-dimensional algorithms to be used in TRAC

CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

Energy Technology Data Exchange (ETDEWEB)

Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

2011-07-01

A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has

CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

International Nuclear Information System (INIS)

Pena, C.; Pellacani, F.; Macian Juan, R.; Chiva, S.; Barrachina, T.; Miro, R.

2011-01-01

A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has been

Fine numerical modelling of thermohydraulic phenomena in EDF PWR reactors

International Nuclear Information System (INIS)

Boulot, F.

1993-01-01

Over the last 20 years, EDF has developed a family of 2D and 3D industrial thermohydraulics software to solve problems encountered in existing PWR power plants and to design new reactors for the future. The equations used in the models are the averaged Navier-Stokes and energy equations. A brief description is given of the four main codes developed for single-phase and two-phase water-steam flows, some of which use finite differences or finite volumes methods, while others make use of finite elements methods. An example of application is given for each code. (author). 4 figs., 4 refs

Calculation of drop course of control rod assembly in PWR

International Nuclear Information System (INIS)

Zhou Xiaojia; Mao Fei; Min Peng; Lin Shaoxuan

2013-01-01

The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

Two-phase flow dynamics in ECC

International Nuclear Information System (INIS)

Albraaten, P.J.

1981-07-01

The present report summarizes the achievements within the project ''Two-phase Systems and ECC''. The results during 1978 - 1980 are accounted for in brief as they have been documented in earlier reports. The results during the first half of 1981 are accounted for in greater detail. They contain a new model for the Basset force and test runs with this model using the test code RISQUE. Furthermore, test runs have been performed with TRAC-PD2 MOD 1. This code was implemented on Edwards Pipe Blowdown experiment (a standard test case) and UC-Berkeley Reflooding experiment (a non-standard test case.) (Auth.)

NESTLE: A nodal kinetics code

International Nuclear Information System (INIS)

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

1993-01-01

The NESTLE nodal kinetics code has been developed for utilization as a stand-alone code for steady-state and transient reactor neutronic analysis and for incorporation into system transient codes, such as TRAC and RELAP. The latter is desirable to increase the simulation fidelity over that obtained from currently employed zero- and one-dimensional neutronic models and now feasible due to advances in computer performance and efficiency of nodal methods. As a stand-alone code, requirements are that it operate on a range of computing platforms from memory-limited personal computers (PCs) to supercomputers with vector processors. This paper summarizes the features of NESTLE that reflect the utilization and requirements just noted

Heat Transfer treatment in computer codes for safety analysis

International Nuclear Information System (INIS)

Jerele, A.; Gregoric, M.

1984-01-01

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

Influence of boron reduction strategies on PWR accident management flexibility

International Nuclear Information System (INIS)

Papukchiev, Angel Aleksandrov; Liu, Yubo; Schaefer, Anselm

2007-01-01

In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. Design changes to reduce boron concentration in the reactor coolant are of general interest regarding three aspects - improved reactivity feedback properties, lower impact of boron dilution scenarios on PWR safety and eventually more flexible accident management procedures. In order to assess the potential advantages through the introduction of boron reduction strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 ppm and 805 ppm. For the assessment of the potential safety advantages of these cores a hypothetical beyond design basis accident has been simulated with the system code ATHLET. The analyses showed improved inherent safety and increased accident management flexibility of the low boron cores in comparison with the standard PWR. (author)

Analysis, by Relap5 code, of boron dilution phenomena in a Small Break Loca Transient, performed in PKL III E 2.2 test

International Nuclear Information System (INIS)

Rizzo, G.; Vella, G.

2007-01-01

The present work is finalized to investigate the E2.2 thermal-hydraulics transient of the PKL III facility, which is a scaled reproduction of a typical German PWR, operated by FRAMATOME-ANP in Erlangen, Germany, within the framework of an international cooperation (OECD/SETH project). The main purpose of the project is to study boron dilution events in Pressurized Water Reactors and to contribute to the assessment of thermal-hydraulic system codes like Relap5. The experimental test PKL III E2.2 investigates the behavior of a typical PWR after a Small Break Loss Of Coolant Accident (SB-LOCA) in a cold leg and an immediate injection of borated water in two cold legs. The main purpose of this work is to simulate the PKL III test facility and particularly its experimental transient by Relap5 system code. The adopted nodalization, already available at Department of Nuclear Engineering (DIN), has been reviewed and applied with an accurate analysis of the experimental test parameters. The main result relies in a good agreement of calculated data with experimental measures for a number of main important variables. (author)

Some thermalhydraulics of closure head adapters in a 3 loops PWR

Energy Technology Data Exchange (ETDEWEB)

Hoffmann, F.; Daubert, O.; Hecker, M. [EDF/DER/National Hydraulics Laboratory, Chatou (France)] [and others

1995-09-01

In 1993 a R&D action, based on numerical simulations and experiments on PWR`s upper head was initiated. This paper presents the test facility TRAVERSIN (a scale model of a 900 MW PWR adapter) and the calculations performed on the geometry of different upper head sections with the Thermalhydraulic Finite Element Code N3S used for 2D and 3D computations. The paper presents the method followed to bring the adapter and upper head study to a successful conclusion. Two complementary approaches are performed to obtain global results on complete fluid flow in the upper head and local results on the flow around the adapters of closure head. A validation test case of these experimental and numerical tools is also presented.

Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code

International Nuclear Information System (INIS)

Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.

2002-01-01

The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)

Sensitivity of risk parameters to human errors for a PWR

International Nuclear Information System (INIS)

Samanta, P.; Hall, R.E.; Kerr, W.

1980-01-01

Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study

Best estimate LB LOCA approach based on advanced thermal-hydraulic codes

International Nuclear Information System (INIS)

Sauvage, J.Y.; Gandrille, J.L.; Gaurrand, M.; Rochwerger, D.; Thibaudeau, J.; Viloteau, E.

2004-01-01

Improvements achieved in thermal-hydraulics with development of Best Estimate computer codes, have led number of Safety Authorities to preconize realistic analyses instead of conservative calculations. The potentiality of a Best Estimate approach for the analysis of LOCAs urged FRAMATOME to early enter into the development with CEA and EDF of the 2nd generation code CATHARE, then of a LBLOCA BE methodology with BWNT following the Code Scaling Applicability and Uncertainty (CSAU) proceeding. CATHARE and TRAC are the basic tools for LOCA studies which will be performed by FRAMATOME according to either a deterministic better estimate (dbe) methodology or a Statistical Best Estimate (SBE) methodology. (author)

Modelling of pellet-cladding interaction in PWR's

International Nuclear Information System (INIS)

Esteves, A.M.; Silva, A.T. e.

1992-01-01

The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

Pressurizer and steam-generator behavior under PWR transient conditions

International Nuclear Information System (INIS)

Wahba, A.B.; Berta, V.T.; Pointner, W.

1983-01-01

Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

SU-E-J-47: Comparison of Online Image Registrations of Varian TrueBeam Cone-Beam CT and BrainLab ExacTrac Imaging Systems

International Nuclear Information System (INIS)

Li, J; Shi, W; Andrews, D; Werner-Wasik, M; Yu, Y; Liu, H

2015-01-01

Purpose To compare online image registrations of TrueBeam cone-beam CT (CBCT) and BrainLab ExacTrac imaging systems. Methods Tests were performed on a Varian TrueBeam STx linear accelerator (Version 2.0), which is integrated with a BrainLab ExacTrac imaging system (Version 6.0.5). The study was focused on comparing the online image registrations for translational shifts. A Rando head phantom was placed on treatment couch and immobilized with a BrainLab mask. The phantom was shifted by moving the couch translationally for 8 mm with a step size of 1 mm, in vertical, longitudinal, and lateral directions, respectively. At each location, the phantom was imaged with CBCT and ExacTrac x-ray. CBCT images were registered with TrueBeam and ExacTrac online registration algorithms, respectively. And ExacTrac x-ray image registrations were performed. Shifts calculated from different registrations were compared with nominal couch shifts. Results The averages and ranges of absolute differences between couch shifts and calculated phantom shifts obtained from ExacTrac x-ray registration, ExacTrac CBCT registration with default window, ExaxTrac CBCT registration with adjusted window (bone), Truebeam CBCT registration with bone window, and Truebeam CBCT registration with soft tissue window, were: 0.07 (0.02–0.14), 0.14 (0.01–0.35), 0.12 (0.02–0.28), 0.09 (0–0.20), and 0.06 (0–0.10) mm, in vertical direction; 0.06 (0.01–0.12), 0.27 (0.07–0.57), 0.23 (0.02–0.48), 0.04 (0–0.10), and 0.08 (0– 0.20) mm, in longitudinal direction; 0.05 (0.01–0.21), 0.35 (0.14–0.80), 0.25 (0.01–0.56), 0.19 (0–0.40), and 0.20 (0–0.40) mm, in lateral direction. Conclusion The shifts calculated from ExacTrac x-ray and TrueBeam CBCT registrations were close to each other (the differences between were less than 0.40 mm in any direction), and had better agreements with couch shifts than those from ExacTrac CBCT registrations. There were no significant differences between True

TRAC-PF1/MOD 1 independent assessment: Semiscale MOD-2A feedwater-line break (S-SF-3) and steam-line break (S-SF-5) tests

International Nuclear Information System (INIS)

Dobranich, D.

1985-11-01

The TRAC-PF1/MOD1 independent assessment project at Sandia is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. As part of this effort, calculations for Semiscale Mod-2A test S-SF-3, a feedwater-line break test, and S-SF-5, a steam-line break test, were performed with TRAC-PF1/MOD1. Most aspects of both the S-SF-3 and S-SF-5 steady-state calculations were found to be in good agreement with data. However, the need for a better steam separator model was identified from the S-SF-3 calculation. Overall, the qualitative behavior of both transients was calculated reasonably well; however, there were some discrepancies in the prediction of the quantitative behavior. The results for the S-SF-3 transient calculation indicate that the primary-to-secondary heat transfer degradation began too early. This was possibly due to overprediction of entrainment in the steam generator boiler, leading to an incorrect calculation of the secondary-side fluid distribution during the steady state. However, there was insufficient data to verify this. Results for the S-SF-5 transient calculation indicate that the primary-side fluid temperature response to a steam-line break was in reasonable agreement with data but the pressure response did not coincide with the data. Uncertainties in the data are sufficient to prevent us from determining the exact cause of this discrepancy, but there is indirect evidence that the calculated rate of phase change in the pressurizer was incorrect. 16 refs., 73 figs., 13 tabs

Characteristics of the aerosols released to the environment after a severe PWR accident

International Nuclear Information System (INIS)

Lhiaubet, G.; Manesse, D.

1988-05-01

In the event of a postulated severe accident on a pressurized water reactor (PWR) involving fuel degradation, gases and aerosols containing radioactive products could be released, with short, medium and long term consequences for the population and the environment. Under such accident conditions, the ESCADRE code system, developed at IPSN (Institute for Nuclear Safety and Protection) can be used to calculate the properties of the substances released and, especially with the AEROSOLS/B2 code, the main characteristics of the aerosols (concentration, size distribution, composition). For conditions representative of severe PWR accidents, by varying different main parameters (structural material aerosols, steam condensation in the containment, etc...), indications are given on the range of characteristics of the aerosols (containing notably Cs, Te, Sr, Ru, etc...) released to the atmosphere. Information is also given on how more accurate data (especially on the chemical forms) will be obtainable in the framework of current or planned experimental programs (HEVA, PITEAS, PHEBUS PF, etc...) [fr

TRAC, a collaborative computer tool for tracer-test interpretation

Directory of Open Access Journals (Sweden)

Fécamp C.

2013-05-01

Full Text Available Artificial tracer tests are widely used by consulting engineers for demonstrating water circulation, proving the existence of leakage, or estimating groundwater velocity. However, the interpretation of such tests is often very basic, with the result that decision makers and professionals commonly face unreliable results through hasty and empirical interpretation. There is thus an increasing need for a reliable interpretation tool, compatible with the latest operating systems and available in several languages. BRGM, the French Geological Survey, has developed a project together with hydrogeologists from various other organizations to build software assembling several analytical solutions in order to comply with various field contexts. This computer program, called TRAC, is very light and simple, allowing the user to add his own analytical solution if the formula is not yet included. It aims at collaborative improvement by sharing the tool and the solutions. TRAC can be used for interpreting data recovered from a tracer test as well as for simulating the transport of a tracer in the saturated zone (for the time being. Calibration of a site operation is based on considering the hydrodynamic and hydrodispersive features of groundwater flow as well as the amount, nature and injection mode of the artificial tracer. The software is available in French, English and Spanish, and the latest version can be downloaded from the web site http://trac.brgm.fr.

Cylindrization of a PWR core for neutronic calculations

International Nuclear Information System (INIS)

Santos, Rubens Souza dos

2005-01-01

In this work we propose a core cylindrization, starting from a PWR core configuration, through the use of an algorithm that becomes the process automated in the program, independent of the discretization. This approach overcomes the problem stemmed from the use of the neutron transport theory on the core boundary, in addition with the singularities associated with the presence of corners on the outer fuel element core of, existents in the light water reactors (LWR). The algorithm was implemented in a computational program used to identification of the control rod drop accident in a typical PWR core. The results showed that the algorithm presented consistent results comparing with an production code, for a problem with uniform properties. In our conclusions, we suggest, for future works, for analyzing the effect on mesh sizes for the Cylindrical geometry, and to compare the transport theory calculations versus diffusion theory, for the boundary conditions with corners, for typical PWR cores. (author)

VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 3: programmer's manual (Revision 2)

International Nuclear Information System (INIS)

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

1985-07-01

The VIPRE thermal-hydraulic computer code for PWR and BWR core analysis has undergone a detailed design review by a committee of experts. A new version of the code, incorporating the committee's recommendations, has been submitted for NRC review and issuance of a safety evaluation report. The changes in the programmers's manual are given

Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient

Energy Technology Data Exchange (ETDEWEB)

Chen, Che-Hao; Shih, Chunkuan [National Tsing Hua Univ., Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Lin, Hao-Tzu [Atomic Energy Council, Taiwan (China). Inst. of Nuclear Energy Research

2013-07-01

Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP

Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient

International Nuclear Information System (INIS)

Chen, Che-Hao; Shih, Chunkuan; Wang, Jong-Rong; Lin, Hao-Tzu

2013-01-01

Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP

Transient performance of flow in circuits of PWR type reactors

International Nuclear Information System (INIS)

Hirdes, V.R.; Carajilescov, P.

1988-09-01

Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

International Nuclear Information System (INIS)

Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

1984-01-01

This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

Energy Technology Data Exchange (ETDEWEB)

Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

2016-05-15

Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

Conceptual study on advanced PWR system

International Nuclear Information System (INIS)

Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

1997-07-01

In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. 1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. 2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. 3) Control rod drive mechanism for fine control : type and function were surveyed. 4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. 5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. 6) Steam injector concepts: analysis and experiment were conducted. 7) Fluidic diode concepts : analysis and experiment were conducted. 8) Wet thermal insulator : tests for thin steel layers and assessment of materials. 9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs

Conceptual study on advanced PWR system

Energy Technology Data Exchange (ETDEWEB)

Bae, Yoon Young; Chang, M H; Yu, K J; Lee, D J; Cho, B H; Kim, H Y; Yoon, J H; Lee, Y J; Kim, J P; Park, C T; Seo, J K; Kang, H S; Kim, J I; Kim, Y W; Kim, Y H

1997-07-01

In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

SAFARI 2000 FPAR TRAC Data for Mongu, Zambia, 1999-2002

Data.gov (United States)

National Aeronautics and Space Administration — Data from the Tracing Architecture and Radiation of Canopies (TRAC) instrument were processed to determine the fraction of intercepted photosynthetically active...

Comparison of PWR-IMF and FR fuel cycles

International Nuclear Information System (INIS)

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

2007-01-01

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

Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

Energy Technology Data Exchange (ETDEWEB)

Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

2015-07-01

During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

Reliability analysis of 2 types of auxiliary feedwater system for PWR

International Nuclear Information System (INIS)

Ekariansyah, Andi Sofrany

2002-01-01

This paper will explain the application of Fault Three Method for analyzing the system reliability of Auxiliary Feedwater System with 2 different configurations taken from PWR type nuclear power plant (NPP) in the USA. The first configuration of Braidwood NPP (design A) basically consists of 1 motor driven pump and 1 diesel driven pump. The second configuration of Haddam Neck NPP (Design B) consists of 2 turbine driven pumps. Based on the P and ID and success criteria the fault trees are constructed to estimate the system failure probabilities quantified from software code PIRAS 1.0. The result shows the second configuration (Design B) with 2 turbine driven pumps have the higher failure probability of 1,06 x 10 - 2 compared with design A of 1,09 x 10 - 3 . The modification of both systems are also tried to analyze its effect to the end result. Qualitatively, the common cause failures of 2 turbine driven pumps contribute to the highest risk of system failure probability. Combination with 1 turbine driven pump and 1 motor driven pump or 1 diesel driven pump will increase the system reliability about 80% and 50% without considering if this configuration is possible to realize in a real plant

Evaluation of leak rate by EPRI code

International Nuclear Information System (INIS)

Isozaki, Toshikuni; Hashiguchi, Issei; Kato, Kiyoshi; Miyazono, Shohachiro

1987-08-01

From 1987, a research on the leak rate from a cracked pipe under BWR or PWR operating condition is going to be carried out at the authors' laboratory. This report describes the computed results by EPRI's leak rate code which was mounted on JAERI FACOM-M380 machine. Henry's critical flow model is used in this program. For the planning of an experimental research, the leak rate from a crack under BWR or PWR operating condition is computed, varying a crack length 2c, crack opening diameter COD and pipe diameter. The COD value under which the minimum detectable leak rate of 5 gpm is given is 0.22 mm or 0.21 mm under the BWR or PWR condition with 2c = 100 mm and 16B pipe geometry. The entire lists are shown in the appendix. (author)

Investigation of palladium-103 production and IR07-103Pd brachytherapy seed preparation

International Nuclear Information System (INIS)

Saidi, Pooneh; Sadeghi, Mahdi; Enferadi, Milad; Aslani, Gholamreza

2011-01-01

Highlights: → We report the cyclotron production of 103-palladium via 103 Rh(p,n) 103 Pd reaction. → 103 Pd was absorbed on resin beads for brachytherapy seed preparation. → The optimum absorption of 103 Pd in resin was achieved at 0.5 M HCl. → Version 5 of MCNP code was employed to model a new 103 Pd brachytherapy seed. - Abstract: In this study, design and fabrication of 103 Pd brachytherapy seed was investigated. The excitation functions of 103 Rh(p,n) 103 Pd and 103 Rh(d,2n) 103 Pd reactions were calculated using EMPIRE (version 3.1 Rivoli), ALICE/ASH and TALYS-1.2 codes, the TENDL-2010 database and compared with the published data. Production of 103 Pd was done via 103 Rh(p,n) 103 Pd nuclear reaction. The target was bombarded with 18 MeV protons at 200 μA beam current for 15 h. After irradiation and radiochemical separation of the electroplated rhodium target, the optimum condition for absorption of 103 Pd into Amberlite (registered) IR-93 resin was achieved at 0.5 M HCl. Version 5 of the (MCNP) Monte Carlo radiation transport code was employed to calculate the dosimetric parameters around the 103 Pd brachytherapy seed. Finally the calculated results were compared with published results for other commercial sources.

Detailed resonance absorption calculations with the Monte Carlo code MCNP and collision probability version of the slowing down code ROLAIDS

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Janssen, A.J.

1994-01-01

Very accurate Mote Carlo calculations with Monte Carlo Code have been performed to serve as reference for benchmark calculations on resonance absorption by U 238 in a typical PWR pin-cell geometry. Calculations with the energy-pointwise slowing down code calculates the resonance absorption accurately. Calculations with the multigroup discrete ordinates code XSDRN show that accurate results can only be achieved with a very fine energy mesh. (authors). 9 refs., 5 figs., 2 tabs

Implantation of multigroup diffusion code 2DB in the IEAv CDC CYBER 170/750 system, and its preliminary evaluation

International Nuclear Information System (INIS)

Prati, A.; Anaf, J.

1988-09-01

The IBM version of the multigroup diffusion code 2DB was implemented in the IEAv CDC CYBER 170/750 system. It was optimized relative to the use of the central memory, limited to 132 K-words, through the memory manager CMM and its partition into three source codes: rectangular and cylindrical geometries, triangular geometry and hexagonal geometry. The reactangular, triangular and hexagonal geometry nodal options were revised and optimized. A fast reactor and a PWR type thermal reactor sample cases were studied. The results are presented and analized. An updated 2DB code user's manual was written in Portugueses and published separately. (author) [pt

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

International Nuclear Information System (INIS)

Chung, Bud-Dong; Kim, Hho-Jung

1990-04-01

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

Performance of high burned PWR fuel during transient

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio

1992-01-01

In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

Monte Carlo based radial shield design of typical PWR reactor

Energy Technology Data Exchange (ETDEWEB)

Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

2017-04-15

This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

Influence of Pd-precursor on the acetoxylation activity of Pd-Sb/TiO{sub 2} catalysts

Energy Technology Data Exchange (ETDEWEB)

Ben-homeid, A. [Benghazi Univ. (Libyan Arab Jamahiriya). Chemistry Dept.; Kalevaru, V.N.; Radnik, J.; Luecke, B.; Martin, A. [Leibniz-Institut fuer Katalyse e.V. an der Universitaet Rostock (Germany)

2013-11-01

The impact of palladium precursors (e.g. chloride-PdCl{sub 2}; acetate-Pd(OAc){sub 2}; nitrate-Pd(NO{sub 3}){sub 2}) on the catalytic properties of Pd-Sb/TiO{sub 2} catalysts used for acetoxylation of toluene has been investigated. The catalysts were characterized by different techniques such as N{sub 2}-adsorption (BET-surface area and pore volume), XRD, TEM, CO-Chemisorption and XPS for better understanding of the catalytic properties of the catalysts. The acetate and nitrate-type precursors exhibited higher surface areas, pore volumes and higher dispersion of Pd, but displayed poor performance compared to chloride precursor. TEM analysis indicated that the size of Pd particles depended upon the nature of Pd-precursor. Among the three, chloride precursor exhibited bigger Pd particles. XPS results revealed that all the fresh catalysts irrespective of Pd-precursor contained Pd in oxidized state (i.e. Pd{sup +2}), while in the spent catalysts such oxidized Pd species were reduced. The catalytic performance was found to depend strongly on the nature of precursor used. Among the three, the catalysts prepared from chloride-type precursor showed much higher overall catalytic activity (68%) than those of nitrate and/or acetate type precursors. Moreover, these two precursors (acetate and nitrate) gave higher total oxidation products due to oxidative decomposition of mainly acetic acid. Furthermore, the catalyst prepared from Cl-precursor revealed easy deactivation due to coke deposition and also due to loss of Pd in the near-surface-region. (orig.)

Modeling study of droplet behavior during blowdown period of large break LOCA based on experimental data

International Nuclear Information System (INIS)

Sakaba, Hiroshi; Umezawa, Shigemitsu; Teramae, Tetsuya; Furukawa, Yuji

2004-01-01

During LOCA (Loss Of Coolant Accident) in PWR, droplets behavior during blowdown period is one of the important phenomena. For example, the spattering from falling liquid film that flows from upper plenum generates those droplets in core region. The behavior of droplets in such flow has strong effect for cladding temperature behavior because these droplets are able to remove heat from a reactor core by its direct contact on fuel rods and its evaporation at the surface. For safety analysis of LOCA in PWR, it is necessary to evaluate droplet diameter precisely in order to predict fuel cladding temperature changing by the calculation code. Based on the test results, a new droplet behavior model was developed for the MCOBRA/TRC code that predicts the droplet behavior during such LOCA events. Furthermore, the verification calculations that simulated some blowdown tests were performed using by the MCOBRA/TRAC code. These results indicated the validity of this droplet model during blow down cooling period. The experiment was focused on investigating the Weber number of steady droplet in the blow down phenomenon of large break LOCA. (author)

75 FR 24323 - American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases

Science.gov (United States)

2010-05-04

...The NRC proposes to amend its regulations to incorporate by reference the 2005 Addenda through 2008 Addenda of Section III, Division 1, and the 2005 Addenda through 2008 Addenda of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code (ASME B&PV Code); and the 2005 Addenda and 2006 Addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code). The NRC also proposes to incorporate by reference ASME Code Case N-722-1, ``Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials Section XI, Division 1,'' and Code Case N-770, ``Alternative Examination Requirements and Acceptance Standards for Class 1 PWR [Pressurized- Water Reactor] Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material with or without Application of Listed Mitigation Activities.''

Monte Carlo based radial shield design of typical PWR reactor

Energy Technology Data Exchange (ETDEWEB)

Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

2016-11-15

Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

Method of characteristics - Based sensitivity calculations for international PWR benchmark

International Nuclear Information System (INIS)

Suslov, I. R.; Tormyshev, I. V.; Komlev, O. G.

2013-01-01

Method to calculate sensitivity of fractional-linear neutron flux functionals to transport equation coefficients is proposed. Implementation of the method on the basis of MOC code MCCG3D is developed. Sensitivity calculations for fission intensity for international PWR benchmark are performed. (authors)

Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Generino, G.

1984-07-01

Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

Timing analysis of PWR fuel pin failures

International Nuclear Information System (INIS)

Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J.; Straka, M.

1992-09-01

Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report

Resfria - a computational routine for thermal-hydraulic analysis of a cooldown in the PWR

International Nuclear Information System (INIS)

Silva Neto, A.J. da; Maciel Filho, L.A.

1989-01-01

This paper presents the computer code RESFRIA, designed to calculate the process parameters in a PWR nuclear power plant during a cooldown normal procedure. The procedure is described and some of the models developed to the simulation of systems and equipments are presented. A simplified flowchart of the computational routine and the results in the form of a diagram, for a typical PWR nuclear power plant, are also presented. (author)

Analyses of PWR boron dilution consequences with the Arrotta code

International Nuclear Information System (INIS)

Johanson, E.; Cheng, H.W.; Sehgal, B.R.

1998-03-01

During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced

Sample problem calculations related to two-phase flow transients in a PWR relief-piping network

International Nuclear Information System (INIS)

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

1981-03-01

Two sample problems related with the fast transients of water/steam flow in the relief line of a PWR pressurizer were calculated with a network-flow analysis computer code STAC (System Transient-Flow Analysis Code). The sample problems were supplied by EPRI and are designed to test computer codes or computational methods to determine whether they have the basic capability to handle the important flow features present in a typical relief line of a PWR pressurizer. It was found necessary to implement into the STAC code a number of additional boundary conditions in order to calculate the sample problems. This includes the dynamics of the fluid interface that is treated as a moving boundary. This report describes the methodologies adopted for handling the newly implemented boundary conditions and the computational results of the two sample problems. In order to demonstrate the accuracies achieved in the STAC code results, analytical solutions are also obtained and used as a basis for comparison

Methanol steam reforming over Pd/ZnO and Pd/CeO{sub 2} catalysts

Energy Technology Data Exchange (ETDEWEB)

Ranganathan, Easwar S.; Bej, Shyamal K.; Thompson, Levi T. [University of Michigan, Department of Chemical Engineering, 3026 H.H. Dow Building, 2300 Hayward Avenue, Ann Arbor, MI 48109-2136 (United States)

2005-08-10

The goal of work described in this paper was to better understand the methanol steam reforming (MSR) activity and selectivity patterns of ZnO and CeO{sub 2} supported Pd catalysts. This reaction is being used to produce H{sub 2}-rich gas for a number of applications including hydrogen fuel cells. The Pd/ZnO catalysts had lower MSR rates but were more selective for the production of CO{sub 2} than the Pd/CeO{sub 2} catalysts. The CH{sub 3}OH conversion rates were proportional to the H{sub 2} chemisorption uptake suggesting that the rate determining step was catalyzed by Pd. The corresponding turnover frequencies averaged 0.8+/-0.3s{sup -1} and 0.4+/-0.2s{sup -1} at 230{sup o}C for the Pd/ZnO and Pd/CeO{sub 2} catalysts, respectively. The selectivities are explained based on the reaction pathways, and characteristics of the support. The key surface intermediate appeared to be a formate. The ZnO supported catalysts had a higher density of acidic sites and favored pathways where the intermediate was converted to CO{sub 2} while the CeO{sub 2} supported catalysts had a higher density of basic sites and favored the production of CO.

The RNase PD2 gene of almond (Prunus dulcis) represents an evolutionarily distinct class of S-like RNase genes.

Science.gov (United States)

Ma, R C; Oliveira, M M

2000-07-01

A cDNA for an S-like RNase (RNase PD2) has been isolated from a pistil cDNA library of Prunus dulcis cv. Ferragnés. The cDNA encodes an acidic protein of 226 amino acid residues with a molecular weight of 25 kDa. A potential N-glycosylation site is present at the N-terminus in RNase PD2. A signal peptide of 23 amino acid residues and a transmembrane domain are predicted. The two active-site histidines present in enzymes of the T2/S RNase superfamily were detected in RNase PD2. Its amino acid sequence shows 71.2% similarity to RNSI of Arabidopsis and RNase T2 of chickpea, respectively. Northern blotting and RT-PCR analyses indicate that PD2 is expressed predominantly in petals, pistils of open flowers and leaves of the almond tree. Analyses of shoots cultured in vitro suggested that the expression of RNase PD2 is associated with phosphate starvation. Southern analysis detected two sequences related to RNase PD2 in the P. dulcis genome. RFLP analysis showed that S-like RNase genes are polymorphic in different almond cultivars. The PD2 gene sequence was amplified by PCR and two introns were shown to interrupt the coding region. Based on sequence analysis, we have defined three classes of S-like RNase genes, with the PD2 RNase gene representing a distinct class. The significance of the structural divergence of S-like RNase genes is further discussed.

ACTRAN: a code for depletion calculations in PWR cores aiming the production of minor actinide for using in ADS

International Nuclear Information System (INIS)

Santos, Rubens Souza dos

2009-01-01

Despite of the renewed willing to accept nuclear power as a mean of mitigate the climate changing, to deal with the long lived waste still cause some concerning in relation to maintain in safety condition, during so many years. A technological solution to overcome this leg of time is to use a facility that burn these waste, besides to generate electricity. This is the idea built in the accelerator driven systems (ADS). This technology is being though to use some minor actinides (MAs) as fuel. This work presents a program to assess actinide concentrations, aiming a fertile-free fuel to be used in the future ADS technology. For that, use was made of a numerical code to solve the steady-state multigroup diffusion equation 3D to calculate the neutron fluxes, coupled it with a new code to solve, also numerically, depletion equations, named ACTRAN code. This paper shows the simulation of a PWR core during the residence time of the nuclear fuel, for three years, and after, for almost four hundred years, to assess the MAs production. The results show some insight in the best management to get a minimum amount of some MAs to use in the future generations of ADS. (author)

Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Cho, Dong Keun

2010-11-01

In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

A code for structural analysis of fuel assemblies

International Nuclear Information System (INIS)

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

1988-08-01

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

Implementation of a dry process fuel cycle model into the DYMOND code

International Nuclear Information System (INIS)

Park, Joo Hwan; Jeong, Chang Joon; Choi, Hang Bok

2004-01-01

For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada Deuterium Uranium (CANDU) reactor, direct use of spent Pressurized Water Reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-through and DUPIC fuel cycles

Application of burnup credit for PWR spent fuel storage pool

International Nuclear Information System (INIS)

Shin, Hee Sung; Ro, Seung-Gy; Bae, Kang Mok; Kim, Ik Soo; Shin, Young Joon

1999-01-01

A study on the application of burnup credit for a PWR spent fuel storage pool has been investigated using a computer code system such as CSAS6 module of SCALE 4.3 in association with 44-group SCALE cross-section library. The calculation bias of the code system at a 95% probability with a 95% confidence level seems to be 0.00951 by benchmarking the system for forty six experimental data. With the aid of this computer code system, criticality analysis has been performed for the PWR spent fuel storage pool. Uncertainties due to postulated abnormal and accidental conditions, and manufacturing tolerance such as stainless steel thickness of storage rack, fuel enrichment, fuel density and box size have statistically been combined and resulted in 0.00674. Also, isotopic correction factor which was based on the calculated and measured concentration of 43 isotopes for both selected actinides and fission products important in burnup credit application has been taken into account in the criticality analysis. It is revealed that the minimum burnup with the corrected isotopic concentrations as required for the safe storage is 5,730 MWd/tU in enriched fuel of 5.0 wt%. (author)

Impact of 2D/3D-project on LOCA-licensing analysis and reactor safety of PWRs

International Nuclear Information System (INIS)

Winkler, F.; Krebs, W.D.

1989-01-01

In the past LOCA-licensing analysis has included large conservatisms to compensate for the lack of detailed two phase flow and full scale experimental data. The 2D/3D-project was established to improve the data base in order to minimize the conservatisms required. The significant results and findings of the full scale Upper Plenum Test Facility (UPTF) and from the electrically heated Slab Core Test Facility (SCTF) were particularly useful for understanding the multidimensional phenomena in the primary system and in the core of a PWR. UPTF results were used to verify the TRAC-PF1 analysis of a PWR with combined ECC-Injection during the reflood phase of a large break-LOCA. Comparison of these results with results from classic licensing calculations quantifies the large safety margin in earlier licensing procedures and in reactor systems. (orig.)

Generalized perturbation theory error control within PWR core-loading pattern optimization

International Nuclear Information System (INIS)

Imbriani, J.S.; Turinsky, P.J.; Kropaczek, D.J.

1995-01-01

The fuel management optimization code FORMOSA-P has been developed to determine the family of near-optimum loading patterns for PWR reactors. The code couples the optimization technique of simulated annealing (SA) with a generalized perturbation theory (GPT) model for evaluating core physics characteristics. To ensure the accuracy of the GPT predictions, as well as to maximize the efficient of the SA search, a GPT error control method has been developed

Validation of the probabilistic approach for the analysis of PWR transients

International Nuclear Information System (INIS)

Amesz, J.; Francocci, G.F.; Clarotti, C.

1978-01-01

This paper reviews the pilot study at present being carried out on the validation of probabilistic methodology with real data coming from the operational records of the PWR power station at Obrigheim (KWO, Germany) operating since 1969. The aim of this analysis is to validate the a priori predictions of reactor transients performed by a probabilistic methodology, with the posteriori analysis of transients that actually occurred at a power station. Two levels of validation have been distinguished: (a) validation of the rate of occurrence of initiating events; (b) validation of the transient-parameter amplitude (i.e., overpressure) caused by the above mentioned initiating events. The paper describes the a priori calculations performed using a fault-tree analysis by means of a probabilistic code (SALP 3) and event-trees coupled with a PWR system deterministic computer code (LOOP 7). Finally the principle results of these analyses are presented and critically reviewed

A computer code (MONA) for pH and chemistry evaluation in the secondary system water of PWR

International Nuclear Information System (INIS)

Nordmann, F.

1983-01-01

Many corrosion phenomena of the PWR secondary system materials, strongly depend on the pH of the fluid. On operating plants, only room temperature pH of the bulk water can be measured. The knowledge of the pH at the operating temperature and its relationship with the measured value is therefore particularly interesting. In addition, an evaluation of the local chemistry in flow-restricted areas of the steam generator (SG) where drastic corrosion generally occurs, is of utmost concern. The MONA code has been developed to compute the secondary water pH in the following cases: at temperatures ranging from 0 to 320 deg C; at any concentration of ammonia; at any amount of pollutants such as sea water, river water (from condenser leak), and/or sodium, chloride, sulfate (from demineralization resins); with possible addition of calcium hydroxide or boric acid in order to inhibit denting or intergranular attack. (author)

Investigation of electronic structure and chemical bonding of intermetallic Pd2HfIn: An ab-initio study

Science.gov (United States)

Bano, Amreen; Gaur, N. K.

2018-05-01

Ab-initio calculations are carried out to study the electronic and chemical bonding properties of Intermetallic full Heusler compound Pd2HfIn which crystallizes in F-43m structure. All calculations are performed by using density functional theory (DFT) based code Quantum Espresso. Generalized gradient approximations (GGA) of Perdew- Burke- Ernzerhof (PBE) have been adopted for exchange-correlation potential. Calculated electronic band structure reveals the metallic character of the compound. From partial density of states (PDoS), we found the presence of relatively high intensity electronic states of 4d-Pd atom at Fermi level. We have found a pseudo-gap just abouve the Fermi level and N(E) at Fermi level is observed to be 0.8 states/eV, these finding indicates the existence of superconducting character in Pd2HfIn.

Evaluation of the RELAP4/MOD6 thermal-hydraulic code

International Nuclear Information System (INIS)

Haigh, W.S.; Margolis, S.G.; Rice, R.E.

1978-01-01

The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA

Synthesis of PdO/MCM-41 nanocomposites using trans-[PdCl2(PEt3)2] as the source of metal

International Nuclear Information System (INIS)

Hernandez-Pineda, Jessica; Rio, Jose Manuel del; Carreto, Enrique; Terres, Eduardo; Montoya, J. Asencion; Zuniga-Gonzalez, Maria de Jesus; Morgado, Jorge

2009-01-01

Three PdO/MCM-41 nanocomposites containing PdO species that are dispersed on an MCM-41 support have been prepared using trans-[PdCl 2 (PEt 3 ) 2 ] as the source of metal. Thermal gravimetric analysis of the decomposition of this complex in air reveals a remarkable mass loss of all nonmetallic elements centered at 300 deg. C. The materials were characterized with X-ray powder diffraction (XRPD), nitrogen sorption and TEM, which showed that the PdO nanoparticles were present inside the pores of MCM-41. XRPD analysis confirms formation of PdO (palladinite) nanocrystals, as the sole crystalline phase present in the nanocomposites.

Advanced codes and methods supporting improved fuel cycle economics - 5493

International Nuclear Information System (INIS)

Curca-Tivig, F.; Maupin, K.; Thareau, S.

2015-01-01

AREVA's code development program was practically completed in 2014. The basic codes supporting a new generation of advanced methods are the followings. GALILEO is a state-of-the-art fuel rod performance code for PWR and BWR applications. Development is completed, implementation started in France and the U.S.A. ARCADIA-1 is a state-of-the-art neutronics/ thermal-hydraulics/ thermal-mechanics code system for PWR applications. Development is completed, implementation started in Europe and in the U.S.A. The system thermal-hydraulic codes S-RELAP5 and CATHARE-2 are not really new but still state-of-the-art in the domain. S-RELAP5 was completely restructured and re-coded such that its life cycle increases by further decades. CATHARE-2 will be replaced in the future by the new CATHARE-3. The new AREVA codes and methods are largely based on first principles modeling with an extremely broad international verification and validation data base. This enables AREVA and its customers to access more predictable licensing processes in a fast evolving regulatory environment (new safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation...). In this context, the advanced codes and methods and the associated verification and validation represent the key to avoiding penalties on products, on operational limits, or on methodologies themselves

The application of modern nodal methods to PWR reactor physics analysis