Pub Date : 1999-01-01DOI: 10.1016/0140-6701(00)94994-4
Zhou Weihong
{"title":"IMPROVING FIRE PROTECTION : GUANGDONG DEVISES ITS OWN PERFORMANCE INDICATORS","authors":"Zhou Weihong","doi":"10.1016/0140-6701(00)94994-4","DOIUrl":"https://doi.org/10.1016/0140-6701(00)94994-4","url":null,"abstract":"","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1999-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1016/0140-6701(00)94994-4","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"53496000","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Equations of the typical ID two-fluid model of two-phase flow, which is used in today’s computer codes for the simulations in nuclear thermal-hydraulics (RELAP5, TRAC), can be written in vectorial form as:∂ψ∂t+C∂ψ∂x=P Numerical scheme in RELAP5 is based on direct discretization of the equations using first-order accurate temporal and spatial discretizations. Staggered grid is used and upwind scheme for the spatial discretization of the convection terms in the equations. Second-order accurate central differences are used only for the discretization of the pressure gradients in the momentum equations. In this paper we have demonstrated the capability of the RELAP5/MOD3 code to trace the acoustic waves with second-order accuracy, if a very small time step is chosen for the simulations. This feature of the RELAP5 code is important especially for the simulations of the fast transients describing pressure waves, i.e. shocks, rarefaction waves, water hammer... Second-order accuracy cannot be achieved for the propagation of the temperature and void fraction waves.
{"title":"How to Get Second-Order Accurate Solutions From the First-Order Accurate RELAP5 Code","authors":"I. Tiselj, S. Petelin","doi":"10.1115/imece1998-1128","DOIUrl":"https://doi.org/10.1115/imece1998-1128","url":null,"abstract":"Equations of the typical ID two-fluid model of two-phase flow, which is used in today’s computer codes for the simulations in nuclear thermal-hydraulics (RELAP5, TRAC), can be written in vectorial form as:∂ψ∂t+C∂ψ∂x=P Numerical scheme in RELAP5 is based on direct discretization of the equations using first-order accurate temporal and spatial discretizations. Staggered grid is used and upwind scheme for the spatial discretization of the convection terms in the equations. Second-order accurate central differences are used only for the discretization of the pressure gradients in the momentum equations. In this paper we have demonstrated the capability of the RELAP5/MOD3 code to trace the acoustic waves with second-order accuracy, if a very small time step is chosen for the simulations. This feature of the RELAP5 code is important especially for the simulations of the fast transients describing pressure waves, i.e. shocks, rarefaction waves, water hammer... Second-order accuracy cannot be achieved for the propagation of the temperature and void fraction waves.","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1998-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86050159","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The results of a low-pressure subcooled flow boiling experiment simulated with the RELAP5/MOD3.2 thermal-hydraulic computer program are presented. The objective of the present study is to assess the suitability of application of the RELAP5 computer program for the thermal hydraulic safety analysis of research reactors. These reactors as the High Flux Australian reactor HIFAR operate at near-atmospheric pressures with upward forced convective flow. Subcooled boiling phenomenon plays an important role in the heat transfer aspects of this reactor. It is found that the void fraction profile predicted by the code for the various experimental conditions differ considerably from the experimental results.
{"title":"Simulation of a Subcooled Boiling Experiment Using RELAP5/MOD3.2 Computer Code","authors":"S. Hari, Y. Hassan, J. Tu","doi":"10.1115/imece1998-1133","DOIUrl":"https://doi.org/10.1115/imece1998-1133","url":null,"abstract":"\u0000 The results of a low-pressure subcooled flow boiling experiment simulated with the RELAP5/MOD3.2 thermal-hydraulic computer program are presented. The objective of the present study is to assess the suitability of application of the RELAP5 computer program for the thermal hydraulic safety analysis of research reactors. These reactors as the High Flux Australian reactor HIFAR operate at near-atmospheric pressures with upward forced convective flow. Subcooled boiling phenomenon plays an important role in the heat transfer aspects of this reactor. It is found that the void fraction profile predicted by the code for the various experimental conditions differ considerably from the experimental results.","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1998-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77331709","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Experimental data indicate that bubbly, turbulent, boundary layer has the same structure as its single-phase counterpart. The modified logarithmic law of the wall for the bubbly turbulent log layer is obtained. Eddy viscosity concept is used to obtain boundary layer equation. Total turbulent stress in the inner layer is assumed to be the sum of the local stress caused by bubbles in the log layer and the stress taking into account the inherent liquid turbulence and bubble-liquid interaction in the outer layer. The proposed two-phase law of the wall can be used as a boundary condition in multidimensional models of two-phase turbulent flows. It is applicable to the upward and downward flows with the value of void fractions in the log layer not more than 10%.
{"title":"Boundary Layer Two-Phase Bubbly Flow Equation","authors":"A. Troshko, Y. Hassan","doi":"10.1115/imece1998-1130","DOIUrl":"https://doi.org/10.1115/imece1998-1130","url":null,"abstract":"\u0000 Experimental data indicate that bubbly, turbulent, boundary layer has the same structure as its single-phase counterpart. The modified logarithmic law of the wall for the bubbly turbulent log layer is obtained. Eddy viscosity concept is used to obtain boundary layer equation. Total turbulent stress in the inner layer is assumed to be the sum of the local stress caused by bubbles in the log layer and the stress taking into account the inherent liquid turbulence and bubble-liquid interaction in the outer layer. The proposed two-phase law of the wall can be used as a boundary condition in multidimensional models of two-phase turbulent flows. It is applicable to the upward and downward flows with the value of void fractions in the log layer not more than 10%.","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1998-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86890177","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Computational Fluid Dynamics (CFD) was used to evaluate Reactor Vessel (RV) shell temperatures during a Direct Vessel Injection (DVI) transient for the Advanced Passive 600 mW Pressurized Water Reactor (AP600). The circumferential, axial and radial temperature distributions were calculated for the RV shell using CFD methods.
{"title":"CFD Analysis of a Direct Vessel Injection (DVI) Transient to Calculate AP600 Reactor Vessel Shell Temperatures","authors":"J. Matarazzo, R. Schwirian","doi":"10.1115/imece1998-1127","DOIUrl":"https://doi.org/10.1115/imece1998-1127","url":null,"abstract":"Computational Fluid Dynamics (CFD) was used to evaluate Reactor Vessel (RV) shell temperatures during a Direct Vessel Injection (DVI) transient for the Advanced Passive 600 mW Pressurized Water Reactor (AP600). The circumferential, axial and radial temperature distributions were calculated for the RV shell using CFD methods.","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1998-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76888626","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A strongly conservative finite volume formulation for complex geometries in three-dimensions using a complete transformation of the governing equations on a nonstaggered grid is presented. This method retains its conservative character at the scalar discretization level. The use of physical contravariant components as dependent variables eliminates the need for any transformation to calculate the cell face mass fluxes. A partially implicit treatment of the nonorthogonal diffusion terms is used to enhance the diagonal dominance of the scheme. This is an extension of the method proposed by Sharatchandra (1994). The method is then tested for two test problems for which analytical solutions are available and an error analysis is performed.
{"title":"A Three Dimensional Finite Volume Method in Curvilinear Coordinates for Complex Geometries: Formulation and Analysis","authors":"Sibashis S. Banerjee, Y. Hassan","doi":"10.1115/imece1998-1132","DOIUrl":"https://doi.org/10.1115/imece1998-1132","url":null,"abstract":"\u0000 A strongly conservative finite volume formulation for complex geometries in three-dimensions using a complete transformation of the governing equations on a nonstaggered grid is presented. This method retains its conservative character at the scalar discretization level. The use of physical contravariant components as dependent variables eliminates the need for any transformation to calculate the cell face mass fluxes. A partially implicit treatment of the nonorthogonal diffusion terms is used to enhance the diagonal dominance of the scheme. This is an extension of the method proposed by Sharatchandra (1994). The method is then tested for two test problems for which analytical solutions are available and an error analysis is performed.","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1998-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83694509","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
When performing best estimate calculations uncertainty needs to be quantified. An optimal statistical estimator algorithm was adapted, extended and used for response surface generation. The objective of the study was to demonstrate optimal statistical estimator for uncertainty evaluation of single value or time dependent parameters when Code Scaling, Applicability and Uncertainty (CSAU) method is used for uncertainty quantification. The scenario selected was small-break loss-of-coolant accident with break in cold leg of a two-loop pressurized water reactor Westinghouse type. The uncertainty was quantified for RELAP5/MOD3.2 thermalhydraulic computer code. The results showed that peak cladding temperature selected as one of primary safety criteria with added uncertainty does not exceed the safety limit. Uncertainty was additionally evaluated for nine time dependent parameters. This finding indicate that CSAU method can be applied to transients other than loss-of-coolant accident.
{"title":"Uncertainty Quantification of SB LOCA With CSAU Using Optimal Statistical Estimator","authors":"A. Prošek, B. Mavko","doi":"10.1115/imece1998-1129","DOIUrl":"https://doi.org/10.1115/imece1998-1129","url":null,"abstract":"\u0000 When performing best estimate calculations uncertainty needs to be quantified. An optimal statistical estimator algorithm was adapted, extended and used for response surface generation. The objective of the study was to demonstrate optimal statistical estimator for uncertainty evaluation of single value or time dependent parameters when Code Scaling, Applicability and Uncertainty (CSAU) method is used for uncertainty quantification. The scenario selected was small-break loss-of-coolant accident with break in cold leg of a two-loop pressurized water reactor Westinghouse type. The uncertainty was quantified for RELAP5/MOD3.2 thermalhydraulic computer code.\u0000 The results showed that peak cladding temperature selected as one of primary safety criteria with added uncertainty does not exceed the safety limit. Uncertainty was additionally evaluated for nine time dependent parameters. This finding indicate that CSAU method can be applied to transients other than loss-of-coolant accident.","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1998-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87949249","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In 1993 the International Atomic Energy Agency Standard Problem Exercise no.4 (IAEA-SPE-4), a small break LOCA scenario simulation with no high pressure injection system available, was performed on the PMK-2 integral test facility in Budapest. Later, in 1994, another SB LOCA experiment was performed. The analyses of the PMK-2 facility response, a model of WER-440 nuclear power plant, have been performed using the MOD3.2.1.2 version of the RELAP5 thermal-hydraulic computer code.
{"title":"Hydroaccumulator Influence on SB LOCA Transients at PMK-2 Test Facility","authors":"S. Petelin, I. Parzer","doi":"10.1115/imece1998-1131","DOIUrl":"https://doi.org/10.1115/imece1998-1131","url":null,"abstract":"\u0000 In 1993 the International Atomic Energy Agency Standard Problem Exercise no.4 (IAEA-SPE-4), a small break LOCA scenario simulation with no high pressure injection system available, was performed on the PMK-2 integral test facility in Budapest. Later, in 1994, another SB LOCA experiment was performed. The analyses of the PMK-2 facility response, a model of WER-440 nuclear power plant, have been performed using the MOD3.2.1.2 version of the RELAP5 thermal-hydraulic computer code.","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1998-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87468833","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A simulation of VVER1000/320 operational transient with CATHARE2 V1.3L was performed. This transient consisted of shutdown of primary side pump in one of the loops. Before the transient, reactor was at nominal operational condition with 72% power level. The results of comparison between calculated and measured data indicated that the code was able to reasonably reproduce main phenomena taking place in both primary and secondary sides.
{"title":"Analysis of VVER1000/320 Operational Transient With CATHARE Computer Program","authors":"A. Troshko, Y. Hassan","doi":"10.1115/imece1998-1134","DOIUrl":"https://doi.org/10.1115/imece1998-1134","url":null,"abstract":"\u0000 A simulation of VVER1000/320 operational transient with CATHARE2 V1.3L was performed. This transient consisted of shutdown of primary side pump in one of the loops. Before the transient, reactor was at nominal operational condition with 72% power level.\u0000 The results of comparison between calculated and measured data indicated that the code was able to reasonably reproduce main phenomena taking place in both primary and secondary sides.","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1998-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74324732","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
During the last few years the safety of the Hungarian Paks Nuclear Power Plant was reassessed in the framework of the AGNES project. Results of the program revealed that the safety of VVER-440/213 type reactors could be enhanced by modernizing a number of event oriented emergency operating procedures (EOPs) and the implementation of accident management (AM). Before the accomplishment of systematic AM all the possible thermal-hydraulic effects and consequences should be considered and experimentally verified. This paper summarizes the most important findings of the relevant tests performed in the PMK-2 facility, which is a full-pressure, integral-type model of the Paks NPP. The present analysis concentrates on a particular method, the bleed-and-feed, applied in the primary or secondary circuit and conclusions are drawn for the effectiveness of this AM measure. Modelling of the thermal-hydraulic processes is extended with computer simulations using the RELAP5/MOD3.2 system code. Finally, a short assessment is given for the code capabilities to represent some significant phenomena of the transients.
{"title":"Simulation of Some Emergency Operating Procedure for VVER-440 Reactors","authors":"J. Bánáti, G. Ézsöl","doi":"10.1115/imece1997-0607","DOIUrl":"https://doi.org/10.1115/imece1997-0607","url":null,"abstract":"\u0000 During the last few years the safety of the Hungarian Paks Nuclear Power Plant was reassessed in the framework of the AGNES project. Results of the program revealed that the safety of VVER-440/213 type reactors could be enhanced by modernizing a number of event oriented emergency operating procedures (EOPs) and the implementation of accident management (AM). Before the accomplishment of systematic AM all the possible thermal-hydraulic effects and consequences should be considered and experimentally verified.\u0000 This paper summarizes the most important findings of the relevant tests performed in the PMK-2 facility, which is a full-pressure, integral-type model of the Paks NPP. The present analysis concentrates on a particular method, the bleed-and-feed, applied in the primary or secondary circuit and conclusions are drawn for the effectiveness of this AM measure. Modelling of the thermal-hydraulic processes is extended with computer simulations using the RELAP5/MOD3.2 system code. Finally, a short assessment is given for the code capabilities to represent some significant phenomena of the transients.","PeriodicalId":49736,"journal":{"name":"Nuclear Engineering International","volume":null,"pages":null},"PeriodicalIF":0.6,"publicationDate":"1997-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83921880","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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