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

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)

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)

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)

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

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

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

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

Assessment of predictive capability of REFLA/TRAC code for large break LOCA transient in PWR using LOFT L2-5 test data

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

1994-03-01

The REFLA/TRAC code is a best estimate code developed at Japan Atomic Energy Research Institute (JAERI) to provide advanced predictions of thermal hydraulic transient in light water reactors (LWRs). The REFLA/TRAC code uses the TRAC-PF1/MOD1 code as the framework of the code. The REFLA/TRAC code is expected to be used for the calibration of licensing codes, accident analysis, accident simulation of LWRs, and design of advanced LWRs. Several models have been implemented to the TRAC-PF1/MOD1 code at JAERI including reflood model, condensation model, interfacial and wall friction models, etc. These models have been verified using data from various separate effect tests. This report describes an assessment result of the REFLA/TRAC code, which was performed to assess the predictive capability for integral system behavior under large break loss of coolant accident (LBLOCA) using data from the LOFT L2-5 test. The assessment calculation confirmed that the REFLA/TRAC code can predict break mass flow rate, emergency core cooling water bypass and clad temperature excellently in the LOFT L2-5 test. The CPU time of the REFLA/TRAC code was about 1/3 of the TRAC-PF1/MOD1 code. The REFLA/TRAC code can perform stable and fast simulation of thermal hydraulic behavior in PWR LBLOCA with enough accuracy for practical use. (author)

Assessment of one dimensional reflood model in REFLA/TRAC code

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

1993-12-01

Post-test calculations for twelve selected SSRTF, SCTF and CCTF tests were performed to assess the predictive capability of the one-dimensional reflood model in the REFLA/TRAC code for core thermal behavior during the reflood in a PWR LOCA. Both core void fraction profile and clad temperature transients were predicted excellently by the REFLA/TRAC code including parameter effect of core inlet subcooling, core flooding rate, core configuration, core power, system pressure, initial clad temperature and so on. The peak clad temperature was predicted within an error of 50 K. Based on these assessment results, it is verified that the core thermal hydraulic behaviors during the reflood can be predicted excellently with the REFLA/TRAC code under various conditions where the reflood may occur in a PWR LOCA. (author)

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

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

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

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

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

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

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)

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

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

Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

International Nuclear Information System (INIS)

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

1993-03-01

As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

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)

TRAC-PD2 analysis of FLECHT experiments

International Nuclear Information System (INIS)

Bott, T.F.; Mandell, D.A.

1981-01-01

This report describes TRAC-PD2 calculations of FLECHT (Full Length Emergency Cooling Heat Transfer) tests 4831 and 17201. The calculations were performed as part of the TRAC-PD2 developmental assessment where the objective was to assess TRAC-PD2 reflood modeling under forced flooding conditions. Calculated and experimental values for peak fuel-rod clad temperature, clad quenching time, and rod bundle effluent rates are compared; and calculations with an approximate radiation heat-transfer model added to the basic TRAC-PD2 code are performed. Findings demonstrate the potential importance of surface-to-surface radiation heat transfer in these tests

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

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

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-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

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.

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

Application of the TRAC-PD2 code to the simulation of the CANON experiment

International Nuclear Information System (INIS)

Neves Conti, T. das; Freitas, R.L.

1985-01-01

A comparison between the TRAC -PD2 code calculations and results from the CANON experiment is presented. The CANON experiment simulates the loss of coolant accident through the depressurization of a horizontal tube containing water at different temperatures. The experiment consist of the instantaneous rupture at one end of the tubing and the corresponding pressure and void fraction measurements during the transient. The comparison shows that the TRAC-PD2 code predicts satisfactorily the pressure and void fraction evolution in the CANON experiment. (F.C.) [pt

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)

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

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)

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)

TRAC-PF1/MOD1 computer code

International Nuclear Information System (INIS)

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

1983-01-01

TRAC-PF1 was designed to improve the ability of TRAC-PD2 to handle small-break LOCAs and other transients. TRAC-PF1 has all of the major improvements of TRAC-PD2 but, in addition, uses a full two-fluid model with two-step numerics in the one-dimensional components. The two-fluid model, in conjunction with a stratified-flow regime, handles countercurrent flow better than the drift-flux model previously used. The two-step numerics allow large time steps to be taken for slow transients. TRAC-PF1/MOD1 was designed to provide full balance-of-plant modeling capabilities. This required addition of a general capability for modeling plant control systems. The steam generator model was replaced to allow a wider variety of feedwater connections and better modeling of steam tube ruptures. A special turbine component also has been added, but new components were not required for adequate modeling of condensors, heaters, and pumps in the secondary system

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

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-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

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

Independent assessment of TRAC-PD2 and RELAP5/MOD1 codes at BNL in FY 1981

International Nuclear Information System (INIS)

Saha, P.; Jo, J.H.; Neymotin, L.; Rohatgi, U.S.; Slovik, G.

1982-12-01

This report documents the independent assessment calculations performed with the TRAC-PD2 and RELAP/MOD1 codes at Brookhaven National Laboratory (BNL) during Fiscal Year 1981. A large variety of separate-effects experiments dealing with (1) steady-state and transient critical flow, (2) level swell, (3) flooding and entrainment, (4) steady-state flow boiling, (5) integral economizer once-through steam generator (IEOTSG) performance, (6) bottom reflood, and (7) two-dimensional phase separation of two-phase mixtures were simulated with TRAC-PD2. In addition, the early part of an overcooling transient which occurred at the Rancho Seco nuclear power plant on March 20, 1978 was also computed with an updated version of TRAC-PD2. Three separate-effects tests dealing with (1) transient critical flow, (2) steady-state flow boiling, and (3) IEOTSG performance were also simulated with RELAP5/MOD1 code. Comparisons between the code predictions and the test data are presented

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.

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

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)

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)

TRAC-PD2 modeling of LOFT and PWR small cold-leg breaks

International Nuclear Information System (INIS)

Knight, T.D.; Willcutt, G.J.E. Jr.; Lime, J.F.

1981-01-01

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. TRAC-PD2, the latest publicly released version of the code, is currently being tested against small-break and other transients in experimental facilities; it is also being used to analyze postulated accidents in commercial power reactors. Calculated results for LOFT small-break experiments are compared to data, and the results from two small-break calculations for two different reactor systems are presented. It is concluded that TRAC-PD2 is useful for the analysis of cold-leg small-break accidents

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

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

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

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

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

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

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

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

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

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)

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.

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

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

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/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)

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)

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.)

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)

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

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

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

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.

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

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.

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

PWR plant transient analyses using TRAC-PF1

International Nuclear Information System (INIS)

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

1984-01-01

This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

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

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

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

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.

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...

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

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

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)

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

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

Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

International Nuclear Information System (INIS)

Ireland, J.R.

1982-01-01

A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950 0 K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated

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

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

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

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

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

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

An HMS/TRAC analysis of a high-level radioactive waste tank

International Nuclear Information System (INIS)

Travis, J.R.; Nichols, B.D.; Spore, J.W.; Wilson, T.L.

1991-01-01

It has been observed that a high-level radioactive waste tank generates quantities of hydrogen and nitrous oxide mixtures that are potentially well within flammability limits. These gases are produced from chemical and nuclear decay reactions in a slurry of radioactive waste material. The slurry is covered by a thick crust composed of sodium nitrate and nitrite salts. Significant amounts of the combustible and reactant gases are produced over a 3- to 4-month period before the crust ruptures and the gases are vented into the air cover gas space above the crust. Postulating an ignition of the hydrogen/nitrous oxide/air mixture after this venting into the cover gas, we have calculated the pressure and temperature loading on the double-walled waste tank with the three-dimensional, time-dependent fluid dynamics coupled with chemical kinetics HMS (Hydrogen Mixing Studies) computer code. The waste tank has a ventilation system designed to maintain a slight negative gage pressure during steady-state operation. We have modeled the ventilation system with TRAC (the Transient Reactor Analysis Code), and we have coupled these two best-estimate accident analysis tools to provide the ventilation response to pressure and temperatures generated by the hydrogen burn. Significant pressures are produced by this event, and the threat to the tank's integrity currently is being evaluated. 3 refs., 4 figs

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 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

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.

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)

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)

Comparisons of TRAC-PD2 calculations with Semiscale Mod-3 small-break tests

International Nuclear Information System (INIS)

Gilbert, J.S.; Sahota, M.S.; Boyack, B.E.; Booker, C.P.; Meier, J.K.

1981-01-01

Five experiments conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory (INEL) were calculated using the latest released version of the Transient Reactor Analysis Code (TRAC-PD2). The results were used to assess TRAC-PD2 predictions of thermal-hydraulic phenomena and the effects of pump operation on system response during slow transients. Tests S-SB-P1, S-SB-P2, and S-SB-P7 simulated equivalent 2.5% communicative cold-leg breaks for early pump-trip (pumps-off), intermediate pump-trip (pumps-on), and late pump-trip (pumps-on) operation, respectively. Tests S-SB-P3 and S-SB-P4 simulated equivalent 2.5% communicative hot-leg breaks for pumps-off and pumps-on operation, respectively. Parameters examined in the study included primary system mass distribution, mass inventory, and void fraction distribution

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

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

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.)

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

Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes

International Nuclear Information System (INIS)

Hwang, Moon Kyu; Park, Jee Won; Chung, Bub Dong; Kim, Soo Hyung; Kim, See Dal

2007-07-01

The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE

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 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

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.

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)

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

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

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

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

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

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.

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

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

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

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

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)

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

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

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

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

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

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

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)

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

Safety analysis code input automation using the Nuclear Plant Data Bank

International Nuclear Information System (INIS)

Kopp, H.; Leung, J.; Tajbakhsh, A.; Viles, F.

1985-01-01

The Nuclear Plant Data Bank (NPDB) is a computer-based system that organizes a nuclear power plant's technical data, providing mechanisms for data storage, retrieval, and computer-aided engineering analysis. It has the specific objective to describe thermohydraulic systems in order to support: rapid information retrieval and display, and thermohydraulic analysis modeling. The Nuclear Plant Data Bank (NPBD) system fully automates the storage and analysis based on this data. The system combines the benefits of a structured data base system and computer-aided modeling with links to large scale codes for engineering analysis. Emphasis on a friendly and very graphically oriented user interface facilitates both initial use and longer term efficiency. Specific features are: organization and storage of thermohydraulic data items, ease in locating specific data items, graphical and tabular display capabilities, interactive model construction, organization and display of model input parameters, input deck construction for TRAC and RELAP analysis programs, and traceability of plant data, user model assumptions, and codes used in the input deck construction process. The major accomplishments of this past year were the development of a RELAP model generation capability and the development of a CRAY version of the code

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

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

International Nuclear Information System (INIS)

Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

2005-01-01

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

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

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

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)

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).

Analysis of an ADS spurious opening event at a BWR/6 by means of the TRACE code

International Nuclear Information System (INIS)

Nikitin, Konstantin; Manera, Annalisa

2011-01-01

Highlights: → The spurious opening of 8 relief valves of the ADS system in a BWR/6 has been simulated. → The valves opening results in a fast depressurization and significant loads on the RPV internals. → This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the available plant data. - Abstract: The paper presents the results of a post-event analysis of a spurious opening of 8 relief valves of the automatic depressurization system (ADS) occurred in a BWR/6. The opening of the relief valves results in a fast depressurization (pressure blow down) of the primary system which might lead to significant dynamic loads on the RPV and associated internals. In addition, the RPV level swelling caused by the fast depressurization might lead to undesired water carry-over into the steam line and through the safety relief valves (SRVs). Therefore, the transient needs to be characterized in terms of evolution of pressure, temperature and fluid distribution in the system. This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the plant data.

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

Posttest TRAC-PD2/MOD1 predictions for FLECHT SEASET test 31504

International Nuclear Information System (INIS)

Booker, C.P.

1982-01-01

TRAC-PD2/MOD1 is a publicly released version of TRAC that is used primarily to analyze large-break loss-of-coolant accidents in pressurized-water reactors (PWRs). TRAC-PD2 can calculate, among other things, reflood phenomena. TRAC posttest predictions are compared with test 31504 reflood data from the Full-Length Emergency Core Heat Transfer (FLECHT) System Effects and Separate Effects Tests (SEASET) facility. A false top-down quench is predicted near the top of the core and the subcooling is underpredicted at the bottom of the core. However, the overall TRAC predictions are good, especially near the center of the core

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

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

Statistical safety evaluation of BWR turbine trip scenario using coupled neutron kinetics and thermal hydraulics analysis code SKETCH-INS/TRACE5.0

International Nuclear Information System (INIS)

Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

2012-01-01

The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)

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

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)

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

Assessment of MSIV full closure for Santa Maria de Garona Nuclear Power Plant using TRAC-BF1 (G1J1)

International Nuclear Information System (INIS)

Crespo, J.L.

1993-06-01

This document presents a spurious Main Steam Isolation Value (MSIV) closure analysis for Santa Maria de Garorta Nuclear Power Plan describing the problems found when comparing calculated and real data. The plant is a General Electric Boiling Water Reactor 3, containment type Mark 1. It is operated by NUCLENOR, S.A. and was connected to the grid in 1971. The analysis has been performed by the Apphed Physics Department from the University of Cantabria and the Analysis and Operation Section from NUCLENOR, S.A. as a part of an agreement for developing an engineering simulator of operational transients and accidents for Santa Maria de Gamma Power Plant. The analysis was performed using the frozen version of TRAC-BFI (GlJl) code and is the second of two NUCLENOR contributions to the International Code Applications and Assessment Program (ICAP). The code was run in a Cyber 932 with operating system NOS/VE, property of NUCLENOR, S.A.. A programming effort was carried out in order to provide suitable graphics from the output file

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

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

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)

A friend man-machine interface for thermo-hydraulic simulation codes of nuclear installations

International Nuclear Information System (INIS)

Araujo Filho, F. de; Belchior Junior, A.; Barroso, A.C.O.; Gebrim, A.

1994-01-01

This work presents the development of a Man-Machine Interface to the TRAC-PF1 code, a computer program to perform best estimate analysis of transients and accidents at nuclear power plants. The results were considered satisfactory and a considerable productivity gain was achieved in the activity of preparing and analyzing simulations. (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

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

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

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

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

First intermediate break test 6IB1 data comparison with a TRAC-BD1/MOD1 blind calculation

International Nuclear Information System (INIS)

Wheatley, P.D.

1985-04-01

TRAC-BD1/MOD1 has been used to calculate the behavior in the FIST (Full Integral Test Facility) facility during an intermediate break in one of the recirculation loops. Results of the calculation are compared with the data from the experiment, and the analysis is discussed in this report. The calculation was blind with only the initial and boundary conditions available prior to performance of the calculation. The calculation has been previously documented without reference to the experimental data (i.e., prior to release of the data). This report extends the prior report by discussing the analysis of the data to code comparisons. This work was performed as part of the Nuclear Regulatory Commission's support to the FIST program which is being provided at the Idaho National Engineering Laboratory

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)

Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

International Nuclear Information System (INIS)

Ireland, J.R.

1982-01-01

A 200% double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops and cold-leg emergency-core-cooling systems (ECCS). The calculated peak cladding temperature of 950 K occurred during blowdown and the cladding temperature excursion was terminated at 175 s when complete core quenching occurred. Accumulator flows were initiated at 10 s when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated. Top quenching was caused by entrainment from the lower plenum and lower core regions. The entrained liquid was sufficient to form a small, saturated pool (0.3 m deep) above the upper core support plate. Also, some of the entrained liquid was carried out the hot legs and vaporized in the steam generators. Strong multidimensional effects were calculated in the reactor vessel, particularly with respect to rod quenching

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)

Perspectives on the development of next generation reactor systems safety analysis codes

International Nuclear Information System (INIS)

Zhang, H.

2015-01-01

'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

Perspectives on the development of next generation reactor systems safety analysis codes

Energy Technology Data Exchange (ETDEWEB)

Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

2015-07-01

'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

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

Investigation of Two-Phase Flow Regime Maps for Development of Thermal-Hydraulic Analysis Codes

International Nuclear Information System (INIS)

Kim, Kyung Doo; Kim, Byoung Jae; Lee, Seong Wook

2010-04-01

This reports is a literature survey on models and correlations for determining flow pattern that are used to simulate thermal-hydraulics in nuclear reactors. Determination of flow patterns are a basis for obtaining physical values of wall/interfacial friction, wall/interfacial heat transfer, and droplet entrainment/de-entrainment. Not only existing system codes, such as RELAP5-3D, TRAC-M, MARS, TRACE, CATHARE) but also up-to-date researches were reviewed to find models and correlations

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

Science.gov (United States)

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...

75 FR 11226 - 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 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...

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

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....

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

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

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

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.

Comparison of Interfacial and Wall Friction Models in Thermal-Hydraulic System Analysis Codes (Rev1.0)

International Nuclear Information System (INIS)

Hwang, Moon Kyu; Kim, Soo Hyung; Kim, Byung Jae; Chung, Bub Dong; Kim, Hee Cheol

2010-04-01

This reports is a literature survey on models and correlations for interfacial and wall friction models that are used to simulate thermal-hydraulics in nuclear reactors. The interfacial and wall frictions are needed to solve the momentum equations of gas, continuous liquid and droplet. Not only existing system codes, such as RELAP5-3D, TRAC-M, MARS, TRACE, CATHARE) but also up-to-date researches were reviewed. This report is a revised version of the previous technical report(KAERI/TR-3437/2007)

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...

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

Evaluation of accuracy in the ExacTrac 6D image induced radiotherapy using CBCT

Energy Technology Data Exchange (ETDEWEB)

Park, Ho Chun; Kim, Hyo Jung; Kim, Jong Deok; Ji, Dong Hwa; Song, Ju Young [Dept. of Radiation Oncology, Chonnam National University School, Gwangju (Korea, Republic of)

2016-12-15

To verify the accuracy of the image guided radiotherapy using ExacTrac 6D couch, the error values in six directions are randomly assigned and corrected and then the corrected values were compared with CBCT image to check the accurateness of ExacTrac. The therapy coordination values in the Rando head Phantom were moved in the directions of X, Y and Z as the translation group and they were moved in the directions of pitch, roll and yaw as the rotation group. The corrected values were moved in 6 directions with the combined and mutual reactions. The Z corrected value ranges from 1mm to 23mm. In the analysis of errors between CBCT image of the phantom which is corrected with therapy coordinate and 3D/3D matching error value, the rotation group showed higher error value than the translation group. In the distribution of dose for the error value of the therapy coordinate corrected with CBCT, the restricted value of dosage for the normal organs in two groups meet the prescription dose. In terms of PHI and PCI values which are the dose homogeneity of the cancerous tissue, the rotation group showed a little higher in the low dose distribution range. This study is designed to verify the accuracy of ExacTrac 6D couch using CBCT. It showed that in terms of the error value in the simple movement, it showed the comparatively accurate correction capability but in the movement when the angle is put in the couch, it showed the inaccurate correction values. So, if the body of the patient is likely to have a lot of changes in the direction of rotation or there is a lot of errors in the pitch, roll and yaw in ExacTrac correction, it is better to conduct the CBCT guided image to correct the therapy coordinate in order to minimize any side effects.

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

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

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.

Preliminary Study of 1D Thermal-Hydraulic System Analysis Code Using the Higher-Order Numerical Scheme

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Woong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

The existing nuclear system analysis codes such as RELAP5, TRAC, MARS and SPACE use the first-order numerical scheme in both space and time discretization. However, the first-order scheme is highly diffusive and less accurate due to the first order of truncation error. So, the numerical diffusion problem which makes the gradients to be smooth in the regions where the gradients should be steep can occur during the analysis, which often predicts less conservatively than the reality. Therefore, the first-order scheme is not always useful in many applications such as boron solute transport. RELAP7 which is an advanced nuclear reactor system safety analysis code using the second-order numerical scheme in temporal and spatial discretization is being developed by INL (Idaho National Laboratory) since 2011. Therefore, for better predictive performance of the safety of nuclear reactor systems, more accurate nuclear reactor system analysis code is needed for Korea too to follow the global trend of nuclear safety analysis. Thus, this study will evaluate the feasibility of applying the higher-order numerical scheme to the next generation nuclear system analysis code to provide the basis for the better nuclear system analysis code development. The accuracy is enhanced in the spatial second-order scheme and the numerical diffusion problem is alleviated while indicates significantly lower maximum Courant limit and the numerical dispersion issue which produces spurious oscillation and non-physical results in the higher-order scheme. If the spatial scheme is the first order scheme then the temporal second-order scheme provides almost the same result with the temporal firstorder scheme. However, when the temporal second order scheme and the spatial second-order scheme are applied together, the numerical dispersion can occur more severely. For the more in-depth study, the verification and validation of the NTS code built in MATLAB will be conducted further and expanded to handle two

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.

Peach Bottom transient analysis with BWR TRACB02

International Nuclear Information System (INIS)

Alamgir, M.; Sutherland, W.A.

1984-01-01

TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRACRT

International Nuclear Information System (INIS)

Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto

2011-01-01

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 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 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 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)

Comparison of TRAC-PF1/MOD1 to a no-failure UPI test in the Cylindrical Core Test Facility

International Nuclear Information System (INIS)

Cappiello, M.; Spore, J.

1986-01-01

TRAC-PF1/MOD1 is compared to a no-failure upper plenum injection reflood test in the Cylindrical Core Test Facility. The results show that TRAC can accurately predict the asymmetric channeling of fluid from upper plenum into the core and that a multidimensional modeling capability is required to do so. The rod temperature behavior is accurately predicted for both the peak cladding temperature and the quench time in the high- and low-power zones. Excessive downflow of liquid at the tie plate is predicted as a result of the interfacial drag model used in TRAC. 10 figs

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

Vectorization, parallelization and porting of nuclear codes on the VPP500 system (vectorization). Progress report fiscal 1997

International Nuclear Information System (INIS)

Kawasaki, Nobuo; Ogasawara, Shinobu; Adachi, Masaaki; Kume, Etsuo; Ishizuki, Shigeru; Tanabe, Hidenobu; Nemoto, Toshiyuki; Kawai, Wataru; Watanabe, Hideo

1999-05-01

Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system and/or the AP3000 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 14 codes in fiscal 1997. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the vectorization. In this vectorization part, the vectorization of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations, statistical decay code SD and three-dimensional thermal analysis code for in-core test section (T2) of HENDEL SSPHEAT are described. In the parallelization part, the parallelization of cylindrical direct numerical simulation code CYLDNS44N, worldwide version of system for prediction of environmental emergency dose information code WSPEEDI, extension of quantum molecular dynamics code EQMD and three-dimensional non-steady compressible fluid dynamics code STREAM are described. In the porting part, the porting of transient reactor analysis code TRAC-BF1 and Monte Carlo radiation transport code MCNP4A on the AP3000 are described. In addition, a modification of program libraries for command-driven interactive data analysis plotting program IPLOT is described. (author)

Vectorization, parallelization and porting of nuclear codes on the VPP500 system (porting). Progress report fiscal 1997

International Nuclear Information System (INIS)

Ishizuki, Shigeru; Nemoto, Toshiyuki; Kawai, Wataru; Watanabe, Hideo; Tanabe, Hidenobu; Kawasaki, Nobuo; Adachi, Masaaki; Ogasawara, Shinobu; Kume, Etsuo

1999-05-01

Several computer codes in the nuclear field have been vectorized, parallelized and transported on the FUJITSU VPP500 system and/or the AP3000 system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 14 codes in fiscal 1997. These results are reported in 3 parts, i.e., the vectorization part, the parallelization part and the porting part. In this report, we describe the porting. In this porting part, the porting of transient reactor analysis code TRAC-BF1 and Monte Carlo radiation transport code MCNP4A on the AP3000 are described. In addition, a modification of program libraries for command-driven interactive data analysis plotting program IPLOT is described. In the vectorization part, the vectorization of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations, statistical decay code SD and three-dimensional thermal analysis code for in-core test section (T2) of HENDEL SSPHEAT are described. In the parallelization part, the parallelization of cylindrical direct numerical simulation code CYLDNS44N, worldwide version of system for prediction of environmental emergency dose information code WSPEEDI, extension of quantum molecular dynamics code EQMD and three-dimensional non-steady compressible fluid dynamics code STREAM are described. (author)

Clinical accuracy of ExacTrac intracranial frameless stereotactic system

International Nuclear Information System (INIS)

Ackerly, T.; Lancaster, C. M.; Geso, M.; Roxby, K. J.

2011-01-01

Purpose: In this paper, the authors assess the accuracy of the Brainlab ExacTrac system for frameless intracranial stereotactic treatments in clinical practice. Methods: They recorded couch angle and image fusion results (comprising lateral, longitudinal, and vertical shifts, and rotation corrections about these axes) for 109 stereotactic radiosurgery and 166 stereotactic radiotherapy patient treatments. Frameless stereotactic treatments involve iterative 6D image fusion corrections applied until the results conform to customizable pass criteria, theirs being 0.7 mm and 0.5 deg. for each axis. The planning CT slice thickness was 1.25 mm. It has been reported in the literature that the CT slices' thickness impacts the accuracy of localization to bony anatomy. The principle of invariance with respect to patient orientation was used to determine spatial accuracy. Results: The data for radiosurgery comprised 927 image pairs, of which 532 passed (pass ratio of 57.4%). The data for radiotherapy comprised 15983 image pairs, of which 10 050 passed (pass ratio of 62.9%). For stereotactic radiotherapy, the combined uncertainty of ExacTrac calibration, image fusion, and intrafraction motion was (95% confidence interval) 0.290-0.302 and 0.306-0.319 mm in the longitudinal and lateral axes, respectively. The combined uncertainty of image fusion and intrafraction motion in the anterior-posterior coordinates was 0.174-0.182 mm. For stereotactic radiosurgery, the equivalent ranges are 0.323-0.393, 0.337-0.409, and 0.231-0.281 mm. The overall spatial accuracy was 1.24 mm for stereotactic radiotherapy (SRT) and 1.35 mm for stereotactic radiosurgery (SRS). Conclusions: The ExacTrac intracranial frameless stereotactic system spatial accuracy is adequate for clinical practice, and with the same pass criteria, SRT is more accurate than SRS. They now use frameless stereotaxy exclusively at their center.

Financial Health of the Higher Education Sector: Financial Results and TRAC Outcomes 2013-14. Issues Paper 2015/07

Science.gov (United States)

Higher Education Funding Council for England, 2015

2015-01-01

This report provides an overview of the financial health of the Higher Education Funding Council for England (HEFCE)-funded higher education sector in England. The analysis covers financial results for the academic year 2013-14, as submitted to HEFCE in December 2014, as well as the outcomes from the sector's Transparent Approach to Costing (TRAC)…

TechTracS: NASA's commercial technology management system

Science.gov (United States)

Barquinero, Kevin; Cannon, Douglas

1996-03-01

The Commercial Technology Mission is a primary NASA mission, comparable in importance to those in aeronautics and space. This paper will discuss TechTracS, NASA Commercial Technology Management System that has been put into place in FY 1995 to implement this mission. This system is designed to identify and capture the NASA technologies which have commercial potential into an off-the-shelf database application, and then track the technologies' progress in realizing the commercial potential through collaborations with industry. The management system consists of four stages. The first is to develop an inventory database of the agency's entire technology portfolio and assess it for relevance to the commercial marketplace. Those technologies that are identified as having commercial potential will then be actively marketed to appropriate industries—this is the second stage. The third stage is when a NASA-industry partnership is entered into for the purposes of commercializing the technology. The final stage is to track the technology's success or failure in the marketplace. The collection of this information in TechTracS enables metrics evaluation and can accelerate the establishment on direct contacts between and NASA technologist and an industry technologist. This connection is the beginning of the technology commercialization process.

The improvement of the heat transfer coupling between the flow field and the conduction slab for TRAC-PF1

International Nuclear Information System (INIS)

Wang, Lang Chen.

1990-01-01

A semi-implicit coupling technique (treating only the fluid temperature in the energy source term implicitly) of fluid heat-transfer to conduction slabs is presently used in TRAC-PF1. In this study, a fully implicit coupling scheme between the flow field and conduction slab for the one-dimensional capability of TRAC-PF1 has been developed. The new methods treat the heat-transfer coefficient and wall temperature in the energy source term of both the convection and the conduction equation implicitly. In order to test the accuracy of the standard TRAC coupling method and the new methods used in TRAC-PF1, a series of simple tube experiments were modeled with TRAC-PF1 version 3.9B. Additional studies showed the current TRAC-PF1 convergence variables (pressure fraction change and temperature change) and the current convergence criteria are not appropriate for obtaining an accurate result. Use of these values would produce non-physical double values or would adversely affect results in the reflood calculation. Another study also had been done to determine the importance of each HTC variable on computational accuracy and speed by using the non-linear oterative method. It is found that the heat-transfer correlations have a strong dependency on the local void fraction rather than on fluid velocity, fluid temperature, and wall surface temperature in the film boiling heat-transfer regime. At the beginning of a cooldown transient, the old time void fraction is always higher than the new time void fraction in the same time step. The old time void fraction evaluates a lower heat-transfer coefficient for that cell and initiates cooldown later and more slowly. This then leads to a longer quench time and lowers the fluid temperature and the wall temperature plateau for the downstream control volumes

Feasibility study for objective oriented design of system thermal hydraulic analysis program

International Nuclear Information System (INIS)

Chung, Bub Dong; Jeong, Jae Jun; Hwang, Moon Kyu

2008-01-01

The system safety analysis code, such as RELAP5, TRAC, CATHARE etc. have been developed based on Fortran language during the past few decades. Refactoring of conventional codes has been also performed to improve code readability and maintenance. However the programming paradigm in software technology has been changed to use objects oriented programming (OOP), which is based on several techniques, including encapsulation, modularity, polymorphism, and inheritance. In this work, objective oriented program for system safety analysis code has been tried utilizing modernized C language. The analysis, design, implementation and verification steps for OOP system code development are described with some implementation examples. The system code SYSTF based on three-fluid thermal hydraulic solver has been developed by OOP design. The verifications of feasibility are performed with simple fundamental problems and plant models. (author)

XTV users guide

International Nuclear Information System (INIS)

Dearing, J.F.; Johns, R.C.

1996-09-01

XTV is an X-Windows based Graphical User Interface for viewing results of Transient Reactor Analysis Code (TRAC) calculations. It provides static and animated color mapped visualizations of both thermal-hydraulic and heat conduction components in a TRAC model of a nuclear power plant, as well as both on-screen and hard copy two-dimensional plot capabilities. XTV is the successor to TRAP, the former TRAC postprocessor using the proprietary DISSPLA graphics library. This manual describes Version 2.0, which requires TRAC version 5.4.20 or later for full visualization capabilities

Advanced LOCA code uncertainty assessment

International Nuclear Information System (INIS)

Wickett, A.J.; Neill, A.P.

1990-11-01

This report describes a pilot study that identified, quantified and combined uncertainties for the LOBI BL-02 3% small break test. A ''dials'' version of TRAC-PF1/MOD1, called TRAC-F, was used. (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.

Analysis results from the Los Alamos 2D/3D program

International Nuclear Information System (INIS)

Boyack, B.E.; Cappiello, M.W.; Harmony, S.C.; Shire, P.R.; Siebe, D.A.

1987-01-01

Los Alamos National Laboratory is a participant in the 2D/3D program. Activities conducted at Los Alamos National Laboratory in support of 2D/3D program goals include analysis support of facility design, construction, and operation; provision of boundary and initial conditions for test-facility operations based on analysis of pressurized water reactors; performance of pretest and posttest predictions and analyses; and use of experimental results to validate and assess the single- and multi-dimensional, nonequilibrium features in the Transient Reactor Analysis Code (TRAC). During fiscal year 1987, Los Alamos conducted analytical assessment activities using data from the Slab Core Test Facility, The Cylindrical Core Test Facility, and the Upper Plenum Test Facility. Finally, Los Alamos continued work to provide TRAC improvements. In this paper, Los Alamos activities during fiscal year 1987 will be summarized; several significant accomplishments will be described in more detail to illustrate the work activities at Los Alamos

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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)

TRACE/VALKIN: a neutronics-thermohydraulics coupled code to analyze strong 3D transients

Energy Technology Data Exchange (ETDEWEB)

Rafael Miro; Gumersindo Verdu; Ana Maria Sanchez [Chemical and Nuclear Engineering Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain); Damian Ginestar [Applied Mathematics Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain)

2005-07-01

Full text of publication follows: A nuclear reactor simulator consists mainly of two different blocks, which solve the models used for the basic physical phenomena taking place in the reactor. In this way, there is a neutronic module which simulates the neutron balance in the reactor core, and a thermal-hydraulics module, which simulates the heat transfer in the fuel, the heat transfer from the fuel to the water, and the different condensation and evaporation processes taking place in the reactor core and in the condenser systems. TRACE is a two-phase, two-fluid thermal-hydraulic reactor systems analysis code. The TRACE acronym stands for TRAC/RELAP Advanced Computational Engine, reflecting its ability to run both RELAP5 and TRAC legacy input models. It includes a three-dimensional kinetics module called PARCS for performing advanced analysis of coupled core thermal-hydraulic/kinetics problems. TRACE-VALKIN code is a new time domain analysis code to study transients in LWR reactors. This code uses the best estimate code TRACE to give account of the heat transfer and thermal-hydraulic processes, and a 3D neutronics module. This module has two options, the MODKIN option that makes use of a modal method based on the assumption that the neutronic flux can be approximately expanded in terms of the dominant lambda modes associated with a static configuration of the reactor core, and the NOKIN option that uses a one-step backward discretization of the neutron diffusion equation. The lambda modes are obtained using the Implicit Restarted Arnoldi approach or the Jacob-Davidson algorithm. To check the performance of the coupled code TRACE-VALKIN against complex 3D neutronic transients, using the cross-sections tables generated with the translator SIMTAB from SIMULATE to TRACE/VALKIN, the Cofrentes NPP SCRAM-61 transient is simulated. Cofrentes NPP is a General Electric BWR-6 design located in Valencia-land (Spain). It is in operation since 1985 and currently in its fifteenth

SU-F-J-42: Comparison of Varian TrueBeam Cone-Beam CT and BrainLab ExacTrac X-Ray for Cranial Radiotherapy

Energy Technology Data Exchange (ETDEWEB)

Li, J; Shi, W; Andrews, D; Werner-Wasik, M; Lu, B; Yu, Y; Dicker, A; Liu, H [Thomas Jefferson University Hospital, Philadelphia, PA (United States)

2016-06-15

Purpose: To compare online image registrations of TrueBeam cone-beam CT (CBCT) and BrainLab ExacTrac x-ray imaging systems for cranial radiotherapy. Method: Phantom and patient studies were performed on a Varian TrueBeam STx linear accelerator (Version 2.5), which is integrated with a BrainLab ExacTrac imaging system (Version 6.1.1). The phantom study was based on a Rando head phantom, which was designed to evaluate isocenter-location dependence of the image registrations. Ten isocenters were selected at various locations in the phantom, which represented clinical treatment sites. CBCT and ExacTrac x-ray images were taken when the phantom was located at each isocenter. The patient study included thirteen patients. CBCT and ExacTrac x-ray images were taken at each patient’s treatment position. Six-dimensional image registrations were performed on CBCT and ExacTrac, and residual errors calculated from CBCT and ExacTrac were compared. Results: In the phantom study, the average residual-error differences between CBCT and ExacTrac image registrations were: 0.16±0.10 mm, 0.35±0.20 mm, and 0.21±0.15 mm, in the vertical, longitudinal, and lateral directions, respectively. The average residual-error differences in the rotation, roll, and pitch were: 0.36±0.11 degree, 0.14±0.10 degree, and 0.12±0.10 degree, respectively. In the patient study, the average residual-error differences in the vertical, longitudinal, and lateral directions were: 0.13±0.13 mm, 0.37±0.21 mm, 0.22±0.17 mm, respectively. The average residual-error differences in the rotation, roll, and pitch were: 0.30±0.10 degree, 0.18±0.11 degree, and 0.22±0.13 degree, respectively. Larger residual-error differences (up to 0.79 mm) were observed in the longitudinal direction in the phantom and patient studies where isocenters were located in or close to frontal lobes, i.e., located superficially. Conclusion: Overall, the average residual-error differences were within 0.4 mm in the translational

GPS-based microenvironment tracker (MicroTrac) model to estimate time-location of individuals for air pollution exposure assessments: model evaluation in central North Carolina.

Science.gov (United States)

Breen, Michael S; Long, Thomas C; Schultz, Bradley D; Crooks, James; Breen, Miyuki; Langstaff, John E; Isaacs, Kristin K; Tan, Yu-Mei; Williams, Ronald W; Cao, Ye; Geller, Andrew M; Devlin, Robert B; Batterman, Stuart A; Buckley, Timothy J

2014-07-01

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 with 24-h diary data from nine participants, 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.

Development of the Multi-Phase/Multi-Dimensional Code BUBBLEX

International Nuclear Information System (INIS)

Lee, Sang Yong; Kim, Shin Whan; Kim, Eun Kee

2005-01-01

A test version of the two-fluid program has been developed by extending the PISO algorithm. Unlike the conventional industry two-fluid codes, such as, RELAP5 and TRAC, this scheme does not need to develop a pressure matrix. Instead, it adopts the iterative procedure to implement the implicitness of the pressure. In this paper, a brief introduction to the numerical scheme will be presented. Then, its application to bubble column simulation will be described. Some concluding remarks will be followed

Assessment of TRAC-PF1/MOD3 Mark-22 assembly model using SRL ''A'' tank single-assembly flow experiments

International Nuclear Information System (INIS)

Fischer, S.R.; Lam, K.; Lin, J.C.

1991-01-01

This paper summarizes the results of an assessment of our TRAC-PF1/MOD3 Mark-22 prototype fuel assembly model against single-assembly data obtained from the ''A'' Tank single-assembly tests that were performed at the Savannah River Laboratory. We felt the data characterize prototypic assembly behavior over a range of air-water flow conditions of interest for loss-of-coolant accident (LOCA) calculations. This study was part of a benchmarking effort performed to evaluate and validate a multiple-assembly, full-plant model that is being developed by Los Alamos National Laboratory to study various aspects of the Savannah River plant operating conditions, including LOCA transients, using TRAC-PF1/MOD3 Version 1.10. The results of this benchmarking effort demonstrate that TRAC-PF1/MOD3 is capable pf calculating plenum conditions and assembly flows during conditions thought to be typical of the Emergency Cooling System (ECS) phase of a LOCA. 10 refs., 12 fig

Analysis results from the Los Alamos 2D/3D program

International Nuclear Information System (INIS)

Boyack, B.E.; Cappiello, M.W.; Stumpf, H.; Shire, P.; Gilbert, J.; Hedstrom, J.

1986-01-01

Los Alamos National Laboratory is a participant in the 2D/3D program. Activities conducted at Los Alamos National Laboratory in support of 2D/3D program goals include analysis support of facility design, construction, and operation; provision of boundary and initial conditions for test facility operations based on analysis of pressurized water reactors; performance of pretest and posttest predictions and analyses; and use of experimental results to validate and assess the single- and multidimensional nonequilibrium features in the Transient Reactor Analysis Code (TRAC). During Fiscal Year 1986, Los Alamos conducted analytical assessment activities using data from the Cylindrical Core Test Facility and the Slab Core Test Facility. Los Alamos also continued to provide support analysis for the planning of Upper Plenum Test Facility experiments. Finally, Los Alamos either completed or is currently working on three areas of TRAC modeling improvement. In this paper, Los Alamos activities during Fiscal Year 1986 are summarized; several significant accomplishments are described in more detail to illustrate the work activities at Los Alamos

The use of a heat transfer coefficient for describing the radiative exchange between water vapour and the bounding walls

International Nuclear Information System (INIS)

Eifler, W.; Shepherd, I.M.

1983-01-01

During the ''severe-fuel-damage'' experiments of the SUPER SARA test program radiation heat transfer will play an important part. For the analysis of these experiments it should be modelled therefore in a particularly appropriate manner. Based on the same engineering type principles which are used in the radiation model of the TRAC code version for boiling water reactors a new model has been developed. This model is less computer time consuming than the TRAC model and particularly appropriate for the use in the subchannel - type bundle computer code which is planned to be developed for the analysis of the ''severe-fuel-damage'' experiments. Sample calculations for the ''severe-fuel-damage'' test array show that the difference between the results obtained with the new model and those obtained with the TRAC model is in general not significant

FTK: the hardware Fast TracKer of the ATLAS experiment at CERN

CERN Document Server

Maznas, Ioannis; The ATLAS collaboration

2016-01-01

FTK: the hardware Fast TracKer of the ATLAS experiment at CERN In the ever increasing pile-up of the Large Hadron Collider environment, the trigger systems of the experiments have to be exceedingly sophisticated and fast at the same time, in order to select the relevant physics processes against the background processes. The Fast TracKer (FTK) is a track finding implementation at hardware level that is designed to deliver full-scan tracks with $p_{T}$ above 1 GeV to the ATLAS trigger system for every L1 accept (at a maximum rate of 100kHz). To accomplish this, FTK is a highly parallel system which is currently under installation in ATLAS. It will first provide the trigger system with tracks in the central region of the ATLAS detector, and next year it is expected to cover the whole detector. The system is based on pattern matching between hits coming from the silicon trackers of the ATLAS detector and 1 billion simulated patterns stored in specially designed ASIC chips (Associative memory – AM06). In a firs...

Implementation and Performance of FPGA based track fitting for the Atlas Fast TracKer

CERN Document Server

Zou, Rui; The ATLAS collaboration

2018-01-01

The Fast TracKer (FTK) within the ATLAS trigger system provides global track reconstruction for all events passing the ATLAS Level 1 trigger by dividing the detector into parallel processing pipelines that implement pattern matching in custom integrated circuits and data routing, reduction, and parameter extraction in FPGAs. In this presentation we will describe the implementation of a critical component of the system which does partial track fitting using a method based on a principal component analysis at a rate of greater than 1 fit per 10 ps, system-wide, to reduce the output of the pattern matching. Firmware design, timing performance and preliminary results will be discussed.

FIST small break accident analysis with BWR TRACBO2-pretest predictions

International Nuclear Information System (INIS)

Alamgir, M.; Sutherland, W.A.

1984-01-01

The BWR Full Integral Simulation Test (FIST) program includes experimental simulation and analytical evaluation of BWR thermal-hydraulic phenomena during transient events. One such event is a small size break in the suction line of one of the recirculation pumps. The results from a test simulating this transient in the FIST facility are compared with a system analysis using the Transient Reactor Analysis Code (TRACB02). This comparison demonstrates BWR-TRAC capability for small break analyses and provides detailed understanding of the phenomena

EYE-TRAC: monitoring attention and utility for mTBI

Science.gov (United States)

Maruta, Jun; Tong, Jianliang; Lee, Stephanie W.; Iqbal, Zarah; Schonberger, Alison; Ghajar, Jamshid

2012-06-01

Attention is a core function in cognition and also the most prevalent cognitive deficit in mild traumatic brain injury (mTBI). Predictive timing is an essential element of attention functioning because sensory processing and execution of goal-oriented behavior are facilitated by temporally accurate prediction. It is hypothesized that impaired synchronization between prediction and external events accounts for the attention deficit in mTBI. Other cognitive and somatic or affective symptoms associated with mTBI may be explained as secondary consequences of impaired predictive timing. Eye-Tracking Rapid Attention Computation (EYE-TRAC) is the quantification of predictive timing with indices of dynamic visuo-motor synchronization (DVS) between the gaze and the target during continuous predictive visual tracking. Such quantification allows for cognitive performance monitoring in comparison to the overall population as well as within individuals over time. We report preliminary results of normative data and data collected from subjects with a history of mTBI within 2 weeks of injury and post-concussive symptoms at the time of recruitment. A substantial proportion of mTBI subjects demonstrated DVS scores worse than 95% of normal subjects. In addition, longitudinal monitoring of acute mTBI subjects showed that initially abnormal DVS scores were followed by improvement toward the normal range. In summary, EYE-TRAC provides fast and objective indices of DVS that allow comparison of attention performance to a normative standard and monitoring of within-individual changes.

Modeling of High pressure core spray system (HPCS) with TRAC-BF1 code.; Modelacion del HPCS con el codigo TRAC-BF1.

Energy Technology Data Exchange (ETDEWEB)

Angel M, E Del

1994-12-31

In this work we present a model of the HPCS system of Laguna Verde Nuclear Power Plant (CNLV) which consist of a condensate storage tank (TAC) a vertical surge pipe (TOS), implemented by the Comision Federal de Electricidad (CFE), and the suppression pool (PSP), as well as the associated piping and valves for the HPCS pump suction, to study the system under transient state conditions. The analysis results show that the implemented surge pipe, allows a normal HPCS pump start without automatic inadvertented transfer of the HPCS pump suction from the condensate storage tank to the suppression pool. We briefly mention some problems found during the stimulation and their solution, further we point out some deficiencies of the code for this type of studies. The present model can be used to stimulate other transients with only minor modifications of it. (Author).

GPS-based microenvironment tracker (MicroTrac) model to estimate time–location of individuals for air pollution exposure assessments: Model evaluation in central North Carolina

Science.gov (United States)

Breen, Michael S.; Long, Thomas C.; Schultz, Bradley D.; Crooks, James; Breen, Miyuki; Langstaff, John E.; Isaacs, Kristin K.; Tan, Yu-Mei; Williams, Ronald W.; Cao, Ye; Geller, Andrew M.; Devlin, Robert B.; Batterman, Stuart A.; Buckley, Timothy J.

2014-01-01

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 with 24-h diary data from nine participants, 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. PMID:24619294

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)

BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

1985-10-01

Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

PBF severe fuel damage program: results and comparison to analysis

International Nuclear Information System (INIS)

McDonald, P.E.; Buescher, B.J.; Gruen, G.E.; Hobbins, R.R.; McCardell, R.K.

1983-01-01

The United States Nuclear Regulatory Commission has initiated a severe fuel damage research program in the Power Burst Facility (PBF) to investigate fuel rod and core response, and fission product and hydrogen release and transport under degraded core cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

PBF Severe Fuel-Damage Program: results and comparison to analysis

International Nuclear Information System (INIS)

MacDonald, P.E.; Buescher, B.J.; Hobbins, R.R.; McCardell, R.K.; Gruen, G.E.

1983-01-01

The United States Nuclear Regulatory Commission has initiated a severe fuel-damage research program in the Power Burst Facility (PBF) to investigate fuel-rod and core response, and fission-product and hydrogen release and transport under degraded-core-cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

Modeling of High pressure core spray system (HPCS) with TRAC-BF1 code

International Nuclear Information System (INIS)

Angel M, E. Del.

1993-01-01

In this work we present a model of the HPCS system of Laguna Verde Nuclear Power Plant (CNLV) which consist of a condensate storage tank (TAC) a vertical surge pipe (TOS), implemented by the Comision Federal de Electricidad (CFE), and the suppression pool (PSP), as well as the associated piping and valves for the HPCS pump suction, to study the system under transient state conditions. The analysis results show that the implemented surge pipe, allows a normal HPCS pump start without automatic inadvertented transfer of the HPCS pump suction from the condensate storage tank to the suppression pool. We briefly mention some problems found during the stimulation and their solution, further we point out some deficiencies of the code for this type of studies. The present model can be used to stimulate other transients with only minor modifications of it. (Author)

Fuel performance analysis code 'FAIR'

International Nuclear Information System (INIS)

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

1994-01-01

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

Objective Oriented Design of System Thermal Hydraulic Analysis Program and Verification of Feasibility

International Nuclear Information System (INIS)

Chung, Bub Dong; Jeong, Jae Jun; Hwang, Moon Kyu

2008-01-01

The system safety analysis code, such as RELAP5, TRAC, CATHARE etc. have been developed based on Fortran language during the past few decades. Refactoring of conventional codes has been also performed to improve code readability and maintenance. TRACE, RELAP5-3D and MARS codes are examples of these activities. The codes were redesigned to have modular structures utilizing Fortran 90 features. However the programming paradigm in software technology has been changed to use objects oriented programming (OOP), which is based on several techniques, including encapsulation, modularity, polymorphism, and inheritance. It was not commonly used in mainstream software application development until the early 1990s. Many modern programming languages now support OOP. Although the recent Fortran language also support the OOP, it is considered to have limited functions compared to the modern software features. In this work, objective oriented program for system safety analysis code has been tried utilizing modern C language feature. The advantage of OOP has been discussed after verification of design feasibility

Comparison of Online 6 Degree-of-Freedom Image Registration of Varian TrueBeam Cone-Beam CT and BrainLab ExacTrac X-Ray for Intracranial Radiosurgery.

Science.gov (United States)

Li, Jun; Shi, Wenyin; Andrews, David; Werner-Wasik, Maria; Lu, Bo; Yu, Yan; Dicker, Adam; Liu, Haisong

2017-06-01

The study was aimed to compare online 6 degree-of-freedom image registrations of TrueBeam cone-beam computed tomography and BrainLab ExacTrac X-ray imaging systems for intracranial radiosurgery. Phantom and patient studies were performed on a Varian TrueBeam STx linear accelerator (version 2.5), which is integrated with a BrainLab ExacTrac imaging system (version 6.1.1). The phantom study was based on a Rando head phantom and was designed to evaluate isocenter location dependence of the image registrations. Ten isocenters at various locations representing clinical treatment sites were selected in the phantom. Cone-beam computed tomography and ExacTrac X-ray images were taken when the phantom was located at each isocenter. The patient study included 34 patients. Cone-beam computed tomography and ExacTrac X-ray images were taken at each patient's treatment position. The 6 degree-of-freedom image registrations were performed on cone-beam computed tomography and ExacTrac, and residual errors calculated from cone-beam computed tomography and ExacTrac were compared. In the phantom study, the average residual error differences (absolute values) between cone-beam computed tomography and ExacTrac image registrations were 0.17 ± 0.11 mm, 0.36 ± 0.20 mm, and 0.25 ± 0.11 mm in the vertical, longitudinal, and lateral directions, respectively. The average residual error differences in the rotation, roll, and pitch were 0.34° ± 0.08°, 0.13° ± 0.09°, and 0.12° ± 0.10°, respectively. In the patient study, the average residual error differences in the vertical, longitudinal, and lateral directions were 0.20 ± 0.16 mm, 0.30 ± 0.18 mm, 0.21 ± 0.18 mm, respectively. The average residual error differences in the rotation, roll, and pitch were 0.40°± 0.16°, 0.17° ± 0.13°, and 0.20° ± 0.14°, respectively. Overall, the average residual error differences were cone-beam computed tomography image registration in intracranial treatments.

A GFR benchmark comparison of transient analysis codes based on the ETDR concept

International Nuclear Information System (INIS)

Bubelis, E.; Coddington, P.; Castelliti, D.; Dor, I.; Fouillet, C.; Geus, E. de; Marshall, T.D.; Van Rooijen, W.; Schikorr, M.; Stainsby, R.

2007-01-01

A GFR (Gas-cooled Fast Reactor) transient benchmark study was performed to investigate the ability of different code systems to calculate the transition in the core heat removal from the main circuit forced flow to natural circulation cooling using the Decay Heat Removal (DHR) system. This benchmark is based on a main blower failure in the Experimental Technology Demonstration Reactor (ETDR) with reactor scram. The codes taking part into the benchmark are: RELAP5, TRAC/AAA, CATHARE, SIM-ADS, MANTA and SPECTRA. For comparison purposes the benchmark was divided into several stages: the initial steady-state solution, the main blower flow run-down, the opening of the DHR loop and the transition to natural circulation and finally the 'quasi' steady heat removal from the core by the DHR system. The results submitted by the participants showed that all the codes gave consistent results for all four stages of the benchmark. In the steady-state the calculations revealed some differences in the clad and fuel temperatures, the core and main loop pressure drops and in the total Helium mass inventory. Also some disagreements were observed in the Helium and water flow rates in the DHR loop during the final natural circulation stage. Good agreement was observed for the total main blower flow rate and Helium temperature rise in the core, as well as for the Helium inlet temperature into the core. In order to understand the reason for the differences in the initial 'blind' calculations a second round of calculations was performed using a more precise set of boundary conditions

FERRET data analysis code

International Nuclear Information System (INIS)

Schmittroth, F.

1979-09-01

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

Post-test analysis of PANDA test P4

International Nuclear Information System (INIS)

Hart, J.; Woudstra, A.; Koning, H.

1999-01-01

The results of a post-test analysis of the integral system test P4, which has been executed in the PANDA facility at PSI in Switzerland within the framework of Work Package 2 of the TEPSS project are presented. The post-test analysis comprises an evaluation of the PANDA test P4 and a comparison of the test results with the results of simulations using the RELAPS/MOD3.2, TRAC-BF1, and MELCOR 1.8.4 codes. The PANDA test P4 has provided data about how trapped air released from the drywell later in the transient affects PCCS performance in an adequate manner. The well-defined measurements can serve as an important database for the assessment of thermal hydraulic system analysis codes, especially for conditions that could be met in passively operated advanced reactors, i.e. low pressure and small driving forces. Based on the analysis of the test data, the test acceptance criteria have been met. The test P4 has been successfully completed and the instrument readings were with the permitted ranges. The PCCs showed a favorable and robust performance and a wide margin for decay heat removal from the containment. The PANDA P4 test demonstrated that trapped air, released from the drywell later in the transient, only temporarily and only slightly affected the performance of the passive containment cooling system. The analysis of the results of the RELAPS code showed that the overall behaviour of the test has been calculated quite well with regards to pressure, mass flow rates, and pool boil-down. This accounts both for the pre-test and the post-test simulations. However, due to the one-dimensional, stacked-volume modeling of the PANDA DW, WW, and GDCS vessels, 3D-effects such as in-vessel mixing and recirculation could not be calculated. The post-test MELCOR simulation showed an overall behaviour that is comparable to RELAPS. However, MELCOR calculated almost no air trapping in the PCC tubes that could hinder the steam condensation rate. This resulted in lower calculated

Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

International Nuclear Information System (INIS)

Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

1991-01-01

The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

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

Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

International Nuclear Information System (INIS)

Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

1989-01-01

The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

Evaluations of the setup discrepancy between BrainLAB 6D ExacTrac and cone-beam computed tomography used with the imaging guidance system Novalis-Tx for intracranial stereotactic radiosurgery.

Science.gov (United States)

Oh, Se An; Park, Jae Won; Yea, Ji Woon; Kim, Sung Kyu

2017-01-01

The objective of this study was to evaluate the setup discrepancy between BrainLAB 6 degree-of-freedom (6D) ExacTrac and cone-beam computed tomography (CBCT) used with the imaging guidance system Novalis Tx for intracranial stereotactic radiosurgery. We included 107 consecutive patients for whom white stereotactic head frame masks (R408; Clarity Medical Products, Newark, OH) were used to fix the head during intracranial stereotactic radiosurgery, between August 2012 and July 2016. The patients were immobilized in the same state for both the verification image using 6D ExacTrac and online 3D CBCT. In addition, after radiation treatment, registration between the computed tomography simulation images and the CBCT images was performed with offline 6D fusion in an offline review. The root-mean-square of the difference in the translational dimensions between the ExacTrac system and CBCT was <1.01 mm for online matching and <1.10 mm for offline matching. Furthermore, the root-mean-square of the difference in the rotational dimensions between the ExacTrac system and the CBCT were <0.82° for online matching and <0.95° for offline matching. It was concluded that while the discrepancies in residual setup errors between the ExacTrac 6D X-ray and the CBCT were minor, they should not be ignored.

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

Atarse a estacas como trepar a mástiles: El trac piñeriano

Directory of Open Access Journals (Sweden)

Gallardo Saborido, Emilio J.

2007-12-01

Full Text Available This essay thoroughly studies the last stage of Virgilio Piñera Llera’s dramatic production and the biographical and political context in which it was written. This is to be done by means of the study of El trac (1974. It is highlighted the image of a dynamic Piñera in his work as playwright. So, he was able to adapt and melt the main worries that had been the core of his work till that very moment.

A través del análisis de El trac (1974 se repasa la fase final de la producción teatral de Virgilio Piñera, así como las circunstancias vitales y políticas que la rodearon. Se enfatiza la visión de un Piñera dinámico en su quehacer como dramaturgo, capaz de aprovechar las últimas corrientes teatrales y fundirlas con las preocupaciones vertebrales que hasta entonces habían determinado sus piezas.

Summary of papers on current and anticipated uses of thermal-hydraulic codes

Energy Technology Data Exchange (ETDEWEB)

Caruso, R.

1997-07-01

The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especially faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).

Summary of papers on current and anticipated uses of thermal-hydraulic codes

International Nuclear Information System (INIS)

Caruso, R.

1997-01-01

The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especially faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the 'user effect' is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices)

Big Data Challenges in High Energy Physics Experiments: The ATLAS (CERN) Fast TracKer Approach

CERN Document Server

Sotiropoulou, Calliope Louisa; The ATLAS collaboration

2016-01-01

We live in the era of “Big Data” problems. Massive amounts of data are produced and captured, data that require significant amounts of filtering to be processed in a realistically useful form. An excellent example of a “Big Data” problem is the data processing flow in High Energy Physics experiments, in our case the ATLAS detector in CERN. In the Large Hadron Collider (LHC) 40 million collisions of bunches of protons take place every second, which is about 15 trillion collisions per year. For the ATLAS detector alone 1 Mbyte of data is produced for every collision or 2000 Tbytes of data per year. Therefore what is needed is a very efficient real-time trigger system to filter the collisions (events) and identify the ones that contain “interesting” physics for processing. One of the upgrades of the ATLAS Trigger system is the Fast TracKer system. The Fast TracKer is a real-time pattern matching machine able to reconstruct the tracks of the particles in the inner silicon detector of the ATLAS experim...

Computer codes for safety analysis

International Nuclear Information System (INIS)

Holland, D.F.

1986-11-01

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

Light-water reactor safety analysis codes

International Nuclear Information System (INIS)

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

1980-01-01

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Validation of thermal hydraulic computer codes for advanced light water reactor

International Nuclear Information System (INIS)

Macek, J.

2001-01-01

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

R-matrix analysis code (RAC)

International Nuclear Information System (INIS)

Chen Zhenpeng; Qi Huiquan

1990-01-01

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

Development of realistic thermal hydraulic system analysis code

Energy Technology Data Exchange (ETDEWEB)

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

2002-05-01

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

Development of realistic thermal hydraulic system analysis code

International Nuclear Information System (INIS)

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

2002-05-01

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

Nuclear plant analyzer development and analysis applications

International Nuclear Information System (INIS)

Laats, E.T.

1984-01-01

The Nuclear Plant Analyzer (NPA) is being developed as the U.S. Nuclear Regulatory Commission's (NRC's) state of the art safety analysis and engineering tool to address key nuclear plant safety issues. The NPA integrates the NRC's computerized reactor behavior simulation codes such as RELAP5 and TRAC-BWR, both of which are well-developed computer graphics programs, and large repositories of reactor design and experimental data. Utilizing the complex reactor behavior codes as well as the experiment data repositories enables simulation applications of the NPA that are generally not possible with more simplistic, less mechanistic reactor behavior codes. These latter codes are used in training simulators or with other NPA-type software packages and are limited to displaying calculated data only. This paper describes four applications of the NPA in assisting reactor safety analyses. Two analyses evaluated reactor operating procedures, during off-normal operation, for a pressurized water reactor (PWR) and a boiling water reactor (BWR), respectively. The third analysis was performed in support of a reactor safety experiment conducted in the Semiscale facility. The final application demonstrated the usefulness of atmospheric dispersion computer codes for site emergency planning purposes. An overview of the NPA and how it supported these analyses are the topics of this paper

Parallel processing of structural integrity analysis codes

International Nuclear Information System (INIS)

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

1996-01-01

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

OPR1000 RCP Flow Coastdown Analysis using SPACE Code

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Use of computer codes for system reliability analysis

International Nuclear Information System (INIS)

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

1988-01-01

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

SU-F-J-24: Setup Uncertainty and Margin of the ExacTrac 6D Image Guide System for Patients with Brain Tumors

Energy Technology Data Exchange (ETDEWEB)

Kim, S; Oh, S; Yea, J; Park, J [Yeungnam University Medical Center, Daegu, Daegu (Korea, Republic of)

2016-06-15

Purpose: This study evaluated the setup uncertainties for brain sites when using BrainLAB’s ExacTrac X-ray 6D system for daily pretreatment to determine the optimal planning target volume (PTV) margin. Methods: Between August 2012 and April 2015, 28 patients with brain tumors were treated by daily image-guided radiotherapy using the BrainLAB ExacTrac 6D image guidance system of the Novalis-Tx linear accelerator. DUONTM (Orfit Industries, Wijnegem, Belgium) masks were used to fix the head. The radiotherapy was fractionated into 27–33 treatments. In total, 844 image verifications were performed for 28 patients and used for the analysis. The setup corrections along with the systematic and random errors were analyzed for six degrees of freedom in the translational (lateral, longitudinal, and vertical) and rotational (pitch, roll, and yaw) dimensions. Results: Optimal PTV margins were calculated based on van Herk et al.’s [margin recipe = 2.5∑ + 0.7σ − 3 mm] and Stroom et al.’s [margin recipe = 2∑ + 0.7σ] formulas. The systematic errors (∑) were 0.72, 1.57, and 0.97 mm in the lateral, longitudinal, and vertical translational dimensions, respectively, and 0.72°, 0.87°, and 0.83° in the pitch, roll, and yaw rotational dimensions, respectively. The random errors (σ) were 0.31, 0.46, and 0.54 mm in the lateral, longitudinal, and vertical rotational dimensions, respectively, and 0.28°, 0.24°, and 0.31° in the pitch, roll, and yaw rotational dimensions, respectively. According to van Herk et al.’s and Stroom et al.’s recipes, the recommended lateral PTV margins were 0.97 and 1.66 mm, respectively; the longitudinal margins were 1.26 and 3.47 mm, respectively; and the vertical margins were 0.21 and 2.31 mm, respectively. Conclusion: Therefore, daily setup verifications using the BrainLAB ExacTrac 6D image guide system are very useful for evaluating the setup uncertainties and determining the setup margin.∑σ.

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

CERN Document Server

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

2017-01-01

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

Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

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

How good are thermal-hydraulics codes for analyses of plant transients

International Nuclear Information System (INIS)

Fabic, S.

1996-01-01

In the early seventies, all thermal-hydraulics codes were based on the Homogeneous Equilibrium Model (HEM), represented by three conservation equations: mixture mass, momentum and energy. Various means were utilized to solve the resulting system of equations: finite differences in FLASH, SATAN, RELAP3 and RELAP4, method of characteristics in BLOWDWN2, loop momentum method in RAMONA and NORCOOL, and others. As the result the world came to regard HEM as too restrictive and the Two-Fluid model came into fashion, first featuring a six and later, a seven-equation model. New codes like KACHINA, TRAC and RELAP5 were developed also. Experience and comparisons with test data have recently forced us to wonder whether the ability to 'compute' while considering great many complexities, ran ahead of the ability to competently define various interactions between fluid phases and components that such complex codes require. The long running times are also a problem that needs to be resolved. More recent trends in the treatment of thermal-hydraulics in Power Plant Simulators and in Plant Analyzers will also be discussed

The use of best estimate codes to improve the simulation in real time

International Nuclear Information System (INIS)

Rivero, N.; Esteban, J. A.; Lenhardt, G.

2007-01-01

Best estimate codes are assumed to be the technology solution providing the most realistic and accurate response. Best estimate technology provides a complementary solution to the conservative simulation technology usually applied to determine plant safety margins and perform security related studies. Tecnatom in the early 90's, within the MAS project, pioneered the initiative to implement best estimate code in its training simulators. Result of this project was the implementation of the first six-equations thermal hydraulic code worldwide (TRAC R T), running in a training environment. To meet real time and other specific training requirements, it was necessary to overcome important difficulties. Tecnatom has just adapted the Global Nuclear Fuel core Design code: PANAC 11, and is about to complete the General Electric TRACG04 thermal hydraulic code adaptation. This technology features a unique solution for nuclear plants aiming at providing the highest fidelity in simulation, enabling to consider the simulator as a multipurpose: engineering and training, simulation platform. Besides, a visual environment designed to optimize the models life cycle, covering both pre and post-processing activities, is in its late development phase. (Author)

Static Code Analysis with Gitlab-CI

CERN Document Server

Datko, Szymon Tomasz

2016-01-01

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

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

International Nuclear Information System (INIS)

Sakai, Takaaki

2002-11-01

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

76 FR 64931 - Building Energy Codes Cost Analysis

Science.gov (United States)

2011-10-19

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

Two-dimensional disruption thermal analysis code DREAM

International Nuclear Information System (INIS)

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

1988-08-01

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

The Associative Memory Serial Link Processor for the Fast TracKer (FTK) at ATLAS

International Nuclear Information System (INIS)

Andreani, A; Citterio, M; Liberali, V; Annovi, A; Beretta, M; Beccherle, R; Crescioli, F; Biesuz, N; Billereau, W; Combe, J M; Cipriani, R; Citraro, S; Donati, S; Giannetti, P; Luciano, P; Colombo, A; Dimas, D; Gentsos, C; Kordas, K; Lanza, A

2014-01-01

The Fast TracKer (FTK) is an extremely powerful and very compact processing unit, essential for efficient Level 2 trigger selection in future high-energy physics experiments at the LHC. FTK employs Associative Memories (AM) to perform pattern recognition; input and output data are transmitted over serial links at 2 Gbit/s to reduce routing congestion at the board level. Prototypes of the AM chip and of the AM board have been manufactured and tested, in preparation of the imminent design of the final version

Subchannel analysis code development for CANDU fuel channel

International Nuclear Information System (INIS)

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

1998-07-01

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

Nuclear reactor safety. Quarterly progress report, October 1--December 31, 1977

International Nuclear Information System (INIS)

Jackson, J.F.; Stevenson, M.G.

1978-02-01

Progress in reactor safety research is summarized. LWR studies include TRAC code development for thermal-hydraulic analysis of accidents, containment systems evaluation, and safety experiments. LMFBR studies include SIMMER code development and applications, modeling of core disruptive accidents, and safety test facilities studies. HTGR safety studies cover fission product release and transport, structural evaluation, phenomena modeling, systems analysis, and accident delineation. GCFR studies are focussed on core disruptive testing

Experimental transport analysis code system in JT-60

International Nuclear Information System (INIS)

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

1988-03-01

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

A simple in-surge pressure analysis using the SPACE code

International Nuclear Information System (INIS)

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

2010-01-01

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

Automated uncertainty analysis methods in the FRAP computer codes

International Nuclear Information System (INIS)

Peck, S.O.

1980-01-01

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

Accelerator-driven transmutation reactor analysis code system (ATRAS)

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

Proposal for the transmittal of data to LASL and the reporting of TRAC analyses for the multinational reflood experimental program

International Nuclear Information System (INIS)

Bleiweis, P.B.; Kirchner, W.L.; Sicilian, J.M.

1979-04-01

The proposed form of the digital tape containing the reduced experimental data from any of the 2D/3D facilities (CCTF, SCTF, UPTF, and possibly PKL Core-II) and the procedures which LASL will use in performing TRAC calculations and reporting results are described in this document

Impact testing and analysis for structural code benchmarking

International Nuclear Information System (INIS)

Glass, R.E.

1989-01-01

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

Quantifying reactor safety margins: Part 1: An overview of the code scaling, applicability, and uncertainty evaluation methodology

International Nuclear Information System (INIS)

Boyack, B.E.; Duffey, R.B.; Griffith, P.

1988-01-01

In August 1988, the Nuclear Regulatory Commission (NRC) approved the final version of a revised rule on the acceptance of emergency core cooling systems (ECCS) entitled ''Emergency Core Cooling System; Revisions to Acceptance Criteria.'' The revised rule states an alternate ECCS performance analysis, based on best-estimate methods, may be used to provide more realistic estimates of plant safety margins, provided the licensee quantifies the uncertainty of the estimates and included that uncertainty when comparing the calculated results with prescribed acceptance limits. To support the revised ECCS rule, the NRC and its contractors and consultants have developed and demonstrated a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology. It is an auditable, traceable, and practical method for combining quantitative analyses and expert opinions to arrive at computed values of uncertainty. This paper provides an overview of the CSAU evaluation methodology and its application to a postulated cold-leg, large-break loss-of-coolant accident in a Westinghouse four-loop pressurized water reactor with 17 /times/ 17 fuel. The code selected for this demonstration of the CSAU methodology was TRAC-PF1/MOD1, Version 14.3. 23 refs., 5 figs., 1 tab

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

Science.gov (United States)

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

2018-01-01

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

Challenges on innovations of newly-developed safety analysis codes

International Nuclear Information System (INIS)

Yang, Yanhua; Zhang, Hao

2016-01-01

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

Challenges on innovations of newly-developed safety analysis codes

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Code Coupling for Multi-Dimensional Core Transient Analysis

International Nuclear Information System (INIS)

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

2015-01-01

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

Code Coupling for Multi-Dimensional Core Transient Analysis

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

Is ExacTrac x-ray system an alternative to CBCT for positioning patients with head and neck cancers?

International Nuclear Information System (INIS)

Clemente, Stefania; Chiumento, Costanza; Fiorentino, Alba; Cozzolino, Mariella; Oliviero, Caterina; Califano, Giorgia; Caivano, Rocchina; Fusco, Vincenzo; Simeon, Vittorio

2013-01-01

Purpose: To evaluate the usefulness of a six-degrees-of freedom (6D) correction using ExacTrac robotics system in patients with head-and-neck (HN) cancer receiving radiation therapy.Methods: Local setup accuracy was analyzed for 12 patients undergoing intensity-modulated radiation therapy (IMRT). Patient position was imaged daily upon two different protocols, cone-beam computed tomography (CBCT), and ExacTrac (ET) images correction. Setup data from either approach were compared in terms of both residual errors after correction and punctual displacement of selected regions of interest (Mandible, C2, and C6 vertebral bodies).Results: On average, both protocols achieved reasonably low residual errors after initial correction. The observed differences in shift vectors between the two protocols showed that CBCT tends to weight more C2 and C6 at the expense of the mandible, while ET tends to average more differences among the different ROIs.Conclusions: CBCT, even without 6D correction capabilities, seems preferable to ET for better consistent alignment and the capability to see soft tissues. Therefore, in our experience, CBCT represents a benchmark for positioning head and neck cancer patients

CARTOTOOL: un utilitaire de tracé de cartes sur stations de travail et micro-ordinateurs

Directory of Open Access Journals (Sweden)

Xavier Bemardet

1990-05-01

Full Text Available CARTOTOOL, logiciel interactif de tracé de cartes à différentes échelles, fonctionne, aussi bien sur écran monochrome que couleur, dans un environnement multifenêtrage SUNTOOLS. Développé pour utiliser WORLD DATA BASE II, il peut aussi tracer n'importe quel type de fichier de contours ayant un format compatible avec celui de la base. Outre ses fonctions de dessin, CARTOTOOL permet la recherche manuelle de coordonnées de points et le positionnement de relevés.

Development of the next generation reactor analysis code system, MARBLE

International Nuclear Information System (INIS)

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

2011-03-01

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

Uncertainty analysis of the FRAP code

International Nuclear Information System (INIS)

Peck, S.O.

1978-01-01

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

Performance testing of thermal analysis codes for nuclear fuel casks

International Nuclear Information System (INIS)

Sanchez, L.C.

1987-01-01

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

CATHENA 4. A thermalhydraulics network analysis code

International Nuclear Information System (INIS)

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

2009-01-01

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

Two-phase wall friction model for the trace computer code

International Nuclear Information System (INIS)

Wang Weidong

2005-01-01

The wall drag model in the TRAC/RELAP5 Advanced Computational Engine computer code (TRACE) has certain known deficiencies. For example, in an annular flow regime, the code predicts an unphysical high liquid velocity compared to the experimental data. To address those deficiencies, a new wall frictional drag package has been developed and implemented in the TRACE code to model the wall drag for two-phase flow system code. The modeled flow regimes are (1) annular/mist, (2) bubbly/slug, and (3) bubbly/slug with wall nucleation. The new models use void fraction (instead of flow quality) as the correlating variable to minimize the calculation oscillation. In addition, the models allow for transitions between the three regimes. The annular/mist regime is subdivided into three separate regimes for pure annular flow, annular flow with entrainment, and film breakdown. For adiabatic two-phase bubbly/slug flows, the vapor phase primarily exists outside of the boundary layer, and the wall shear uses single-phase liquid velocity for friction calculation. The vapor phase wall friction drag is set to zero for bubbly/slug flows. For bubbly/slug flows with wall nucleation, the bubbles are presented within the hydrodynamic boundary layer, and the two-phase wall friction drag is significantly higher with a pronounced mass flux effect. An empirical correlation has been studied and applied to account for nucleate boiling. Verification and validation tests have been performed, and the test results showed a significant code improvement. (authors)

Transitioning from interpretive to predictive in thermal hydraulic codes

International Nuclear Information System (INIS)

Mousseau, V.A.

2004-01-01

accuracy of thermal hydraulic codes. The first step was a detailed modified equation analysis of a simplified form of the gas momentum equation. This analysis indicated the source of some of the errors in the current thermal hydraulic codes such as RELAP and TRAC. This analysis also showed the benefit of employing an implicitly balanced solution method and particularly a second order in time implicitly balanced method. Even though the analysis was highly simplified, the simulation results on the coupled system of seven equations demonstrated many of the features of the simplified analysis. In particular, the under-relaxation did not seem to consistently have any positive effect on the simulations. The under-relaxation only effected the simulation when the time step was large and the solution was of poor quality. Once the time step was made small enough to provide a quality solution, the under-relaxation had very little effect. The first order in time implicitly balanced solution, apparently because of its smaller truncation error, provided a more accurate solution than the traditional RELAP solution. The second order in time implicitly balanced solution was clearly the most accurate and deserves future study. For the more violent transient, the CFL number for the RELAP solution to obtain the same level of accuracy as the second order in time implicitly balanced was approximately one hundred times smaller. For the gentler transient this ratio goes up to about one thousand. This indicates that the second order in time implicitly balanced solution may be of great value to the long slow transients of modern passive reactor designs. The implicitly balanced method should also be considered for steady state calculations

Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

International Nuclear Information System (INIS)

Aksan, S.N.; Stierli, F.; Analytis, G.T.

1992-03-01

The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended

2D/3D-project - handling of tasks within the framework of the continuation of the 2D/3D-program

International Nuclear Information System (INIS)

Sawitzki, M.; Emmerling, R.; Hertlein, R.; Strobel, R.

1987-01-01

Definition and performance of diverse complex tasks. 1. Comment on the results of calibration tests performed and flow algorithms developed in the USA. Development and recommendations for expansion of calibration testing in the FRG. 2. Compilation and review of input data for GPWR-TRAC and UPTF-TRAC code calculations. 3. Analysis of the results of the GPWR-TRAC code calculation shows that oscillating water plugs form in the hot loops near the location of emergency core coolant injection and travel into the steam generators. In order to reproduce this phenomenon in the UPTF some components and pipes have to be modified and the experimental procedure and mode of operation of the facility have to be investigated. 4. In keeping with the results of Pt. 3, the theoretical basis for a steam generator feedback control system was defined in collaboration with TU Muenchen and 5. the process parameter approach for simulation of actual core behavior was formulated. 6. Collaboration on the preparation of test matrices for UPTF and SCTF III. 7. Provision of eight fuel assembly top end pieces for the Japanese test facility SCTF III. With 13 refs., 52 figs [de

Turbo Pascal Computer Code for PIXE Analysis

International Nuclear Information System (INIS)

Darsono

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

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

International Nuclear Information System (INIS)

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

2013-10-01

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

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

International Nuclear Information System (INIS)

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

1997-09-01

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

A Semantic Analysis Method for Scientific and Engineering Code

Science.gov (United States)

Stewart, Mark E. M.

1998-01-01

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

Content Analysis Coding Schemes for Online Asynchronous Discussion

Science.gov (United States)

Weltzer-Ward, Lisa

2011-01-01

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

Production of analysis code for 'JOYO' dosimetry experiment

International Nuclear Information System (INIS)

Sasaki, Makoto; Nakazawa, Masaharu.

1981-01-01

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

Sensitivity analysis of the RESRAD, a dose assessment code

International Nuclear Information System (INIS)

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

1991-01-01

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

Three-field modeling with droplet entrainment and de-entrainment models for TRAC-M

International Nuclear Information System (INIS)

Lee, Sang Ik

2005-02-01

A three-field modeling has been developed and implemented to the basic one-dimensional components of TRAC-M/F90 (Modernized Transient Reactor Analysis Code, Fortran90) to improve the estimation of the behavior of droplet entrainment. the divide and conquer algorithmic technique is adjusted for the implementation after considering the verifications. The governing equations are composed of the conservation equations of each field with thermal-equilibrium assumption between the two liquid fields. As a result, three momentum equations, four mass equations including noncondensable gas mass equation and two energy equations are used. In the development of the momentum equation for an entrained droplet, two different kinds of approaches are tested: the simple force balance on a single droplet and the field-type equation based on the continuum assumption. The first showed unstable nature and the latter was chosen. All of the necessary empirical correlations were obtained from literatures, in particular, the physical models of COBRA-TF. The sensitivities of each empirical correlation were investigated where the effect of droplet diameter and droplet drag was found to be negligible. The advantages of the SETS numerical scheme in comparison with the semi-implicit nimerical scheme were addressed in the simulation time and the maximum allowed time-step size. Finally, the simulations of Collier and Hewitt's experiment were performed and it is concluded that the three-field model developed in the present study is reasonable. The predictions by the present code agree with the measured values within the RMS errors of 16.5% and 9.6% in the entrained liquid mass flow rates and the pressure gradients, respectively. Assessment on the physical models is performed. The wall and interfacial drag models were determined at first. After that, the assessments of entrainment model were performed. Wurtz's entrainment model, which was used in COBRA-TF and Kataoka and Ishii's model, were mainly

Development of Monte Carlo input code for proton, alpha and heavy ion microdosimetric trac structure simulations

International Nuclear Information System (INIS)

Douglass, M.; Bezak, E.

2010-01-01

Full text: Radiobiology science is important for cancer treatment as it improves our understanding of radiation induced cell death. Monte Carlo simulations playa crucial role in developing improved knowledge of cellular processes. By model Ii ng the cell response to radiation damage and verifying with experimental data, understanding of cell death through direct radiation hits and bystander effects can be obtained. A Monte Carlo input code was developed using 'Geant4' to simulate cellular level radiation interactions. A physics list which enables physically accurate interactions of heavy ions to energies below 100 e V was implemented. A simple biological cell model was also implemented. Each cell consists of three concentric spheres representing the nucleus, cytoplasm and the membrane. This will enable all critical cell death channels to be investigated (i.e. membrane damage, nucleus/DNA). The current simulation has the ability to predict the positions of ionization events within the individual cell components on I micron scale. We have developed a Geant4 simulation for investigation of radiation damage to cells on sub-cellular scale (∼I micron). This code currently allows the positions of the ionisation events within the individual components of the cell enabling a more complete picture of cell death to be developed. The next stage will include expansion of the code to utilise non-regular cell lattice. (author)

Analytical validation of the CACECO containment analysis code

International Nuclear Information System (INIS)

Peak, R.D.

1979-08-01

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

Development of a nuclear power plant system analysis code

International Nuclear Information System (INIS)

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

1997-07-01

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

Level tracking in detailed reactor simulations

Energy Technology Data Exchange (ETDEWEB)

Aktas, B.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

1995-09-01

We introduce a useful test problem for judging the performance of reactor safety codes in situations where moving two-phase mixture levels are present. The test problem tracks a two-phase liquid level as it rises and then falls back to its original position. Pure air exists above the level, and a low void air-water mixture is below the level. Conditions are subcooled and isothermal to remove complications resulting from failures of interfacial heat transfer packages to properly account for the level. Comparisons are made between the performance of current versions of CATHARE, RELAP5, TRAC-BF1, and TRAC-PF1. These system codes are based on finite-difference methods with a fixed, Eulerian staggered grid in space. When a partially filled cell with a mixture level discontinuity becomes the donor cell, the sharp changes in fluid properties across the interface results in numerical oscillations of various terms. Furthermore, the cell-to-cell convection of mass, momentum and energy are inaccurately predicted nearby a mixture level. To adequately model moving mixture levels, an efficient discontinuity tracking method for the finite-difference Eulerian approximations is described. This model had been implemented in the TRAC-BWR code for the two-phase mixture level tracking since the TRAC-BD1 Version (released April 1984). The result of the test problem run by the current version of TRAC-BF1/MOD1 with the mixture level tracking model shows some peculiar behavior of the variables such as velocities, pressures and interfacial terms. A systematic approach to improving performance of the tracking method is described. Implementing this approach in TRAC-BF1/MOD1 has shown a major improvement in the results.

Code requirements document: MODFLOW 2.1: A program for predicting moderator flow patterns

International Nuclear Information System (INIS)

Peterson, P.F.

1992-03-01

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in operation of the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW code discussed here provides transient moderator flow pattern information with stratification effects, and tracks the location of ink plumes in the reactor. The code, written in Fortran, is compiled for Macintosh II computers, and includes subroutines for interactive control and graphical output. Removing the graphics capabilities, the code can also be compiled on other computers. With graphics, in addition to the capability to perform safety related computations, MODFLOW also provides an easy tool for becoming familiar with flow distributions in SRS reactors

Dispersed flow film boiling: An investigation of the possibility to improve the models implemented in the NRC computer codes for the reflooding phase of the LOCA

International Nuclear Information System (INIS)

Andreani, M.; Yadigaroglu, G.; Paul Scherrer Inst.

1992-08-01

Dispersed Flow Film Boiling is the heat transfer regime that occurs at high void fractions in a heated channel. The way this heat transfer mode is modelled in the NRC computer codes (RELAP5 and TRAC) and the validity of the assumptions and empirical correlations used is discussed. An extensive review of the theoretical and experimental work related with heat transfer to highly dispersed mixtures reveals the basic deficiencies of these models: the investigation refers mostly to the typical conditions of low rate bottom reflooding, since the simulation of this physical situation by the computer codes has often showed poor results. The alternative models that are available in the literature are reviewed, and their merits and limits are highlighted. The modifications that could improve the physics of the models implemented in the codes are identified

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

Calculation of fuel pin failure timing under LOCA conditions

International Nuclear Information System (INIS)

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

1991-10-01

The objective of this research was 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) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B ampersand W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs

Code development for eigenvalue total sensitivity analysis and total uncertainty analysis

International Nuclear Information System (INIS)

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

2015-01-01

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

Development of the integrated system reliability analysis code MODULE

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

Makino, Tohru

2001-03-01

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

PACC information management code for common cause failures analysis

International Nuclear Information System (INIS)

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

1987-01-01

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

Fault tree analysis. Implementation of the WAM-codes

International Nuclear Information System (INIS)

Bento, J.P.; Poern, K.

1979-07-01

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

Developments of fuel performance analysis codes in KEPCO NF

International Nuclear Information System (INIS)

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

2012-01-01

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

A compendium of computer codes in fault tree analysis

International Nuclear Information System (INIS)

Lydell, B.

1981-03-01

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

Multi-dimensional Code Development for Safety Analysis of LMR

International Nuclear Information System (INIS)

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

2006-08-01

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

A code for structural analysis of fuel assemblies

International Nuclear Information System (INIS)

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

1988-08-01

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

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

Science.gov (United States)

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

2008-01-01

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

Best-estimate analysis development for BWR systems

International Nuclear Information System (INIS)

Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

1986-01-01

The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

Impact testing and analysis for structural code benchmarking

International Nuclear Information System (INIS)

Glass, R.E.

1989-01-01

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

NACOM - a code for sodium spray fire analysis

International Nuclear Information System (INIS)

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

2002-01-01

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

Use of computer codes for system reliability analysis

International Nuclear Information System (INIS)

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

1989-01-01

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

Use of computer codes for system reliability analysis

Energy Technology Data Exchange (ETDEWEB)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

Hanawa, Satoshi; Iyoku, Tatsuo

2001-08-01

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

Level-Swell Prediction With RETRAN-3D And Its Application To A BWR Steam-Line-Break Analysis

International Nuclear Information System (INIS)

Aounallah, Y.; Hofer, K.

2003-01-01

Level-swell experiments have often been simulated using system codes, such as TRAC and RELAP, but only cursory assessments have been performed with the operational-transient code RETRAN-3D, the main system code used within the STARS project. The present study, initiated in the framework of a BWR Steam-Line-Break (SLB) accident scenario, addresses this lacuna by performing RETRAN simulations of the General Electric Level-Swell experiments, and by investigating their implications on power plant accident analyses. Parameters to which the predicted level swell is sensitive have been identified, and recommendations on code options are made. The SLB analysis objective was to determine the amount of steam and liquid discharged through the break under specified boundary conditions, and to gauge the results against reference values. The impact of the nodalization of the upper part of the reactor pressure vessel was investigated and found to play an important role, whereas the level swell induced from flashing was found not to be the predominant factor for these simulations. (author)

Development of the versatile reactor analysis code system, MARBLE2

International Nuclear Information System (INIS)

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

2015-07-01

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

An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model

Directory of Open Access Journals (Sweden)

Cesare Frepoli

2008-01-01

Full Text Available Since the 1988 amendment of the 10 CFR 50.46 rule in 1988, Westinghouse has been developing and applying realistic or best-estimate methods to perform LOCA safety analyses. A realistic analysis requires the execution of various realistic LOCA transient simulations where the effect of both model and input uncertainties are ranged and propagated throughout the transients. The outcome is typically a range of results with associated probabilities. The thermal/hydraulic code is the engine of the methodology but a procedure is developed to assess the code and determine its biases and uncertainties. In addition, inputs to the simulation are also affected by uncertainty and these uncertainties are incorporated into the process. Several approaches have been proposed and applied in the industry in the framework of best-estimate methods. Most of the implementations, including Westinghouse, follow the Code Scaling, Applicability and Uncertainty (CSAU methodology. Westinghouse methodology is based on the use of the WCOBRA/TRAC thermal-hydraulic code. The paper starts with an overview of the regulations and its interpretation in the context of realistic analysis. The CSAU roadmap is reviewed in the context of its implementation in the Westinghouse evaluation model. An overview of the code (WCOBRA/TRAC and methodology is provided. Finally, the recent evolution to nonparametric statistics in the current edition of the W methodology is discussed. Sample results of a typical large break LOCA analysis for a PWR are provided.

Computer codes for the analysis of flask impact problems

International Nuclear Information System (INIS)

Neilson, A.J.

1984-09-01

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

Energy Savings Analysis of the Proposed NYStretch-Energy Code 2018

Energy Technology Data Exchange (ETDEWEB)

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

2018-01-20

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

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)

Web interface for plasma analysis codes

Energy Technology Data Exchange (ETDEWEB)

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

2008-04-15

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

Web interface for plasma analysis codes

International Nuclear Information System (INIS)

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

2008-01-01

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

76 FR 57982 - Building Energy Codes Cost Analysis

Science.gov (United States)

2011-09-19

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

Users' guide to CACECO containment analysis code. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Peak, R.D.

1979-06-01

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

CONTEMPT-DG containment analysis code

International Nuclear Information System (INIS)

Deem, R.E.; Rousseau, K.

1982-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-11-15

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

Adaptable Value-Set Analysis for Low-Level Code

OpenAIRE

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

2012-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Thermohydraulic analysis of nuclear power plant accidents by computer codes

International Nuclear Information System (INIS)

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

1982-01-01

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

Dominant accident sequences in Oconee-1 pressurized water reactor

International Nuclear Information System (INIS)

Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

1985-04-01

A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

The analysis of thermal-hydraulic models in MELCOR code

Energy Technology Data Exchange (ETDEWEB)

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

1996-07-15

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

Validation of containment thermal hydraulic computer codes for VVER reactor

Energy Technology Data Exchange (ETDEWEB)

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

2005-07-01

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

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

International Nuclear Information System (INIS)

Engelman, G.P.

1978-10-01

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

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

ZERBERUS - the code for reliability analysis of crack containing structures

International Nuclear Information System (INIS)

Cizelj, L.; Riesch-Oppermann, H.

1992-04-01

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

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

International Nuclear Information System (INIS)

1975-06-01

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

Development of three dimensional transient analysis code STTA for SCWR core

International Nuclear Information System (INIS)

Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping

2015-01-01

Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation

Evaluation of the DRAGON code for VHTR design analysis.

Energy Technology Data Exchange (ETDEWEB)

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

2006-01-12

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

Evaluation of the DRAGON code for VHTR design analysis

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

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

1997-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-01

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

Code system for fast reactor neutronics analysis

International Nuclear Information System (INIS)

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

1983-04-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Development of the containment transient analysis code for the passive reactor

Energy Technology Data Exchange (ETDEWEB)

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

1998-05-01

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

Two-phase flow characteristics analysis code: MINCS

International Nuclear Information System (INIS)

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

1992-03-01

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

Transient electromagnetic analysis in tokamaks using TYPHOON code

International Nuclear Information System (INIS)

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

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Jung, Jong Yeob

2006-11-15

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

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

International Nuclear Information System (INIS)

Park, Joo Hwan; Jung, Jong Yeob

2006-11-01

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

Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

International Nuclear Information System (INIS)

1997-08-01

In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

Analysis of pipe stress using CAESAR II code

International Nuclear Information System (INIS)

Sitandung, Y.B.; Bandriyana, B.

2002-01-01

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

Some computational challenges of developing efficient parallel algorithms for data-dependent computations in thermal-hydraulics supercomputer applications

International Nuclear Information System (INIS)

Woodruff, S.B.

1992-01-01

The Transient Reactor Analysis Code (TRAC), which features a two- fluid treatment of thermal-hydraulics, is designed to model transients in water reactors and related facilities. One of the major computational costs associated with TRAC and similar codes is calculating constitutive coefficients. Although the formulations for these coefficients are local the costs are flow-regime- or data-dependent; i.e., the computations needed for a given spatial node often vary widely as a function of time. Consequently, poor load balancing will degrade efficiency on either vector or data parallel architectures when the data are organized according to spatial location. Unfortunately, a general automatic solution to the load-balancing problem associated with data-dependent computations is not yet available for massively parallel architectures. This document discusses why developers algorithms, such as a neural net representation, that do not exhibit algorithms, such as a neural net representation, that do not exhibit load-balancing problems

Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Seo, K. W.

2006-01-01

A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Seo, K. W

2006-01-15

A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

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

International Nuclear Information System (INIS)

Asahi, Yoshiro; Hirano, Masashi; Sato, Kazuo

1986-12-01

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

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

Science.gov (United States)

Hendricks, Eric S.

2016-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Qualification of ARROTTA code for LWR accident analysis

International Nuclear Information System (INIS)

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

2004-01-01

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

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

OpenAIRE

Sun, Yizeng

2013-01-01

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

Development and application of best-estimate LWR safety analysis codes

International Nuclear Information System (INIS)

Reocreux, M.

1997-01-01

This paper is a review of the status and the future orientations of the development and application of best estimate LWR safety analysis codes. The present status of these codes exhibits a large success and almost a complete fulfillment of the objectives which were assigned in the 70s. The applications of Best Estimate codes are numerous and cover a large variety of safety questions. However these applications raised a number of problems. The first ones concern the need to have a better control of the quality of the results. This means requirements on code assessment and on uncertainties evaluation. The second ones concern needs for code development and specifically regarding physical models, numerics, coupling with other codes and programming. The analysis of the orientations for code developments and applications in the next years, shows that some developments should be made without delay in order to solve today questions whereas some others are more long term and should be tested for example in some pilot programmes before being eventually applied in main code development. Each of these development programmes are analyzed in the paper by detailing their main content and their possible interest. (author)

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

OpenAIRE

Wilson, Andrew M.

2012-01-01

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

Uncertainty analysis of the FRAP code

International Nuclear Information System (INIS)

Peck, S.O.

1978-01-01

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

CINETHICA - Core accident analysis code

International Nuclear Information System (INIS)

Nakata, H.

1989-10-01

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

OECD-LOFT large break LOCA experiments: phenomenology and computer code analyses

International Nuclear Information System (INIS)

Brittain, I.; Aksan, S.N.

1990-08-01

Large break LOCA data from LOFT are a very important part of the world database. This paper describes the two double-ended cold leg break tests LP-02-6 and LP-LB-1 carried out within the OECD-LOFT Programme. Tests in LOFT were the first to show the importance of both bottom-up and top-down quenching during blowdown in removing stored energy from the fuel. These phenomena are discussed in detail, together with the related topics of the thermal performance of nuclear fuel and its simulation by electric fuel rod simulators, and the accuracy of cladding external thermocouples. The LOFT data are particularly important in the validation of integral thermal-hydraulics codes such as TRAC and RELAP5. Several OECD partner countries contributed analyses of the large break tests. Results of these analyses are summarised and some conclusions drawn. 32 figs., 3 tabs., 45 refs

Validation of a Subchannel Analysis Code MATRA Version 1.0

Energy Technology Data Exchange (ETDEWEB)

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

2008-10-15

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

THYDE-NEU: Nuclear reactor system analysis code

International Nuclear Information System (INIS)

Asahi, Yoshiro

2002-03-01

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

Benchmark calculation of subchannel analysis codes

International Nuclear Information System (INIS)

1996-02-01

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

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

Directory of Open Access Journals (Sweden)

Hüseyin Uzunboylu

2017-11-01

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

Analysis of the KUCA MEU experiments using the ANL code system

Energy Technology Data Exchange (ETDEWEB)

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

1982-01-01

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

Analysis of the KUCA MEU experiments using the ANL code system

International Nuclear Information System (INIS)

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

1982-01-01

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

Development of regulatory technology for thermal-hydraulic safety analysis

International Nuclear Information System (INIS)

Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

2001-02-01

The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

Room Heat-Up Analysis with GOTHIC code

International Nuclear Information System (INIS)

Jimenez, G.; Olza, J. M.

2010-01-01

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

Code comparison for accelerator design and analysis

International Nuclear Information System (INIS)

Parsa, Z.

1988-01-01

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

NORTICA - a new code for cyclotron analysis

International Nuclear Information System (INIS)

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

2001-01-01

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

NORTICA—a new code for cyclotron analysis

Science.gov (United States)

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

2001-12-01

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

Establishment of computer code system for nuclear reactor design - analysis

International Nuclear Information System (INIS)

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

1996-01-01

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

Application of coupled codes for safety analysis and licensing issues

International Nuclear Information System (INIS)

Langenbuch, S.; Velkov, K.

2006-01-01

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Performance Comparison of Containment PT analysis between CAP and CONTEMPT Code

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

Development and improvement of safety analysis code for geological disposal

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Severe accident analysis code Sampson for impact project

International Nuclear Information System (INIS)

Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

2001-01-01

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

Current lead thermal analysis code 'CURRENT'

International Nuclear Information System (INIS)

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

1985-08-01

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

Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-15

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

Development of the criticality accident analysis code, AGNES

International Nuclear Information System (INIS)

Nakajima, Ken

1989-01-01

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