Test 6.9c OECD ISP-38 was performed on “BETHSY” facility, France on April 14. 1992 and simulated loss of RHR system during Mid-Loop operation at 0.5% of nominal value core power. Initial liquid level in RCS was at horizontal axis of the hot legs. Pressurizer and steam generator manways were opened 1 s after the transient was initiated. The paper presents the test observation to the physical phenomena comparing to the experimental data using RELAP5 NPA (Nuclear Plant Analyzer) graphical postprocessor. The most important and interesting turned out to be coolant distribution around the loops.
{"title":"Bethsy Test 6.9C Mid-Loop Operation Phenomena Identification Using RELAP5 NPA","authors":"S. Petelin, B. Mavko, M. Jurković","doi":"10.1115/imece1997-0605","DOIUrl":"https://doi.org/10.1115/imece1997-0605","url":null,"abstract":"\u0000 Test 6.9c OECD ISP-38 was performed on “BETHSY” facility, France on April 14. 1992 and simulated loss of RHR system during Mid-Loop operation at 0.5% of nominal value core power. Initial liquid level in RCS was at horizontal axis of the hot legs. Pressurizer and steam generator manways were opened 1 s after the transient was initiated. The paper presents the test observation to the physical phenomena comparing to the experimental data using RELAP5 NPA (Nuclear Plant Analyzer) graphical postprocessor. The most important and interesting turned out to be coolant distribution around the loops.","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":"78168839","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 shape of a rising air bubble in a pipe flow is investigated with the Particle Image Velocimetry (PIV) flow visualization technique. To do so, a test volume of bubbly flow is globally illuminated with a pulsed light from a continuous wave laser with an acoustic-optic beam chopper, and is observed with four CCD cameras connected to frame grabbers. A digital image of the rising bubble is acquired and analyzed to identify its shape. A reconstruction method, based on the Dynamic Generalized Hough Transform (DGHT), is described that can determine the two-dimensional shape of a bubble from a PIV image.
{"title":"Bubbly Flow Identification Using Particle Image Velocimetry","authors":"Y. Hassan, W. Schmidl, J. Ortíz-Villafuerte","doi":"10.1115/imece1997-0606","DOIUrl":"https://doi.org/10.1115/imece1997-0606","url":null,"abstract":"\u0000 The shape of a rising air bubble in a pipe flow is investigated with the Particle Image Velocimetry (PIV) flow visualization technique. To do so, a test volume of bubbly flow is globally illuminated with a pulsed light from a continuous wave laser with an acoustic-optic beam chopper, and is observed with four CCD cameras connected to frame grabbers. A digital image of the rising bubble is acquired and analyzed to identify its shape.\u0000 A reconstruction method, based on the Dynamic Generalized Hough Transform (DGHT), is described that can determine the two-dimensional shape of a bubble from a PIV image.","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":"89833178","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}
This paper provides the scaling-up methodology that was applied from the BETHSY integral test facility to the real Framatome NPP (Nuclear Power Plant). The ISP-27 (International Standard Problem) transient scenario was used, based on test 9.1b. The objectives were to evaluate the ideal scaling-up of BETHSY facility for real NPP and to compare and analyse the physical phenomena known from experimental background with the phenomena predicted by RELAP5/MOD3.2 simulation of real NPP transient. Further, in order to test phenomenological scaling-up basis two models for RELAP5/MOD3.2 code were constructed differing in scaling criteria for the primary cooling system. Special attention was concentrated on heat structures scaling. Data were analysed through plotting plots and NPA’s (Nuclear Plant Analyzer) graphical presentation.
{"title":"Data Transfer From BETHSY 9.1B Experiment to Real NPP","authors":"S. Petelin, B. Mavko, I. Ravnikar, Y. Hassan","doi":"10.1115/imece1997-0609","DOIUrl":"https://doi.org/10.1115/imece1997-0609","url":null,"abstract":"\u0000 This paper provides the scaling-up methodology that was applied from the BETHSY integral test facility to the real Framatome NPP (Nuclear Power Plant). The ISP-27 (International Standard Problem) transient scenario was used, based on test 9.1b. The objectives were to evaluate the ideal scaling-up of BETHSY facility for real NPP and to compare and analyse the physical phenomena known from experimental background with the phenomena predicted by RELAP5/MOD3.2 simulation of real NPP transient. Further, in order to test phenomenological scaling-up basis two models for RELAP5/MOD3.2 code were constructed differing in scaling criteria for the primary cooling system. Special attention was concentrated on heat structures scaling. Data were analysed through plotting plots and NPA’s (Nuclear Plant Analyzer) graphical presentation.","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":"76072967","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 ADI (Alternating Direction Implicit) method is used for solving second order partial differential equations. In this case it is applied for viscid incompressible fluid flow described by the Navier-Stokes equation. The method is tested on the simple case of the abrupt area change for the several Reynolds numbers up to Re = 800. A vortex is formed already at low Reynolds number. It affects the pressure field and contribute to the phenomena complicity. The influence of the boundary conditions and nodalization density is also examined.
{"title":"Features of the Calculational Method at Fluid Flow Modelling","authors":"G. Černe, S. Petelin","doi":"10.1115/imece1997-0604","DOIUrl":"https://doi.org/10.1115/imece1997-0604","url":null,"abstract":"\u0000 The ADI (Alternating Direction Implicit) method is used for solving second order partial differential equations. In this case it is applied for viscid incompressible fluid flow described by the Navier-Stokes equation. The method is tested on the simple case of the abrupt area change for the several Reynolds numbers up to Re = 800. A vortex is formed already at low Reynolds number. It affects the pressure field and contribute to the phenomena complicity. The influence of the boundary conditions and nodalization density is also examined.","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":"80796996","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 paper discusses studies conducted to determine the extent of thermal mixing in the upper head region of the AP600 pressurized water reactor (PWR). This information is ultimately useful in the assessment of thermally-induced stresses in the upper head itself and the components adjacent to it. Coolant temperature profiles in the upper head region are also of interest in the evaluation of transients in which buoyancy forces are significant, such as natural circulation (NC) cooldown.
{"title":"Computational Fluid Dynamics Analysis of Thermal Mixing in the AP600 Upper Head","authors":"R. Schwirian","doi":"10.1115/imece1997-0608","DOIUrl":"https://doi.org/10.1115/imece1997-0608","url":null,"abstract":"\u0000 The paper discusses studies conducted to determine the extent of thermal mixing in the upper head region of the AP600 pressurized water reactor (PWR). This information is ultimately useful in the assessment of thermally-induced stresses in the upper head itself and the components adjacent to it. Coolant temperature profiles in the upper head region are also of interest in the evaluation of transients in which buoyancy forces are significant, such as natural circulation (NC) cooldown.","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":"76960567","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}
Jinzhao Zhang, S. Bajorek, R. M. Kemper, M. Nissley, N. Petkov, L. Hochreiter
The AP600 is a simplified advanced pressurized water reactor (PWR) design incorporating passive safety systems that perform the same function as the active emergency core cooling systems (ECCSs) on the current reactors. In order to verify the effectiveness of the AP600 design features for mitigation of a postulated large-break loss-of-coolant accident (LOCA), the recently USNRC-approved best-estimate LOCA methodology (BELOCA) was applied to perform the AP600 standard safety analysis report large-break LOCA analysis. The applicability of the WCOBRA/TRAC code to model the AP600 unique features was validated against CCTF and UFTE downcomer injection tests, the blowdown and reflood cooling heat transfer uncertainties were re-assessed for the AP600 large-break LOCA conditions, and a conservative minimum film boiling temperature was applied as a bounded parameter for blowdown cooling. The BELOCA methodology was simplified to quantify the code uncertainties due to local and global models as well as the statistical approximation methods, with the other uncertainties being bounded by limiting assumptions on the initial and boundary conditions. The final 95 percentile peak cladding temperature (PCT95%) was 1186 K, which meets the 10CFR50.46 criteria with a considerable margin. It is therefore concluded that the AP600 design is effective in mitigation of a postulated large-break LOCA.
{"title":"Application of the WCOBRA/TRAC Best Estimate Methodology to the AP600 Large-Break LOCA Analysis","authors":"Jinzhao Zhang, S. Bajorek, R. M. Kemper, M. Nissley, N. Petkov, L. Hochreiter","doi":"10.1115/imece1997-0603","DOIUrl":"https://doi.org/10.1115/imece1997-0603","url":null,"abstract":"\u0000 The AP600 is a simplified advanced pressurized water reactor (PWR) design incorporating passive safety systems that perform the same function as the active emergency core cooling systems (ECCSs) on the current reactors. In order to verify the effectiveness of the AP600 design features for mitigation of a postulated large-break loss-of-coolant accident (LOCA), the recently USNRC-approved best-estimate LOCA methodology (BELOCA) was applied to perform the AP600 standard safety analysis report large-break LOCA analysis. The applicability of the WCOBRA/TRAC code to model the AP600 unique features was validated against CCTF and UFTE downcomer injection tests, the blowdown and reflood cooling heat transfer uncertainties were re-assessed for the AP600 large-break LOCA conditions, and a conservative minimum film boiling temperature was applied as a bounded parameter for blowdown cooling. The BELOCA methodology was simplified to quantify the code uncertainties due to local and global models as well as the statistical approximation methods, with the other uncertainties being bounded by limiting assumptions on the initial and boundary conditions. The final 95 percentile peak cladding temperature (PCT95%) was 1186 K, which meets the 10CFR50.46 criteria with a considerable margin. It is therefore concluded that the AP600 design is effective in mitigation of a postulated large-break LOCA.","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":"78860045","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}
Westinghouse’s AP600 thermohydraulic design, with passive safety features, poses new challenges to computer simulation and analyses, the design of experimental test facilities to represent it, and to the proper interpretation of the data from these facilities. The conventional approach of modeling the reactor thermohydraulic system as a closed, steady state, natural circulation loop from which non-dimensional groups of parameters can be derived and used in the design of integral tests, is limited and can not capture the abrupt time-varying open system nature of the new design. A rigorous and systematic, eight-step methodology has been developed to scale and interpret the results from three different integral test facilities, and to relate them to the full scale plant. In this paper, the aforementioned scaling methodology is applied to the analysis of the long term cooling phase of the AP600 behavior. This long term cooling phase, which appears independent of the initiating event, is divided for its analysis into two sub-phases. A first sub-phase dominated by the draining of the large In-Containment Refueling Water Storage Tank through the primary systems, and a second sub-phase characterized by the quasi-steady recirculation of coolant through the reactor vessel and the outside of the primary system. The analysis shows that with a few verifiable assumptions one can determine the key parameters and non-dimensional groups that govern the behavior in either of these sub-phases. One then uses these parameters and non-dimensional groups to evaluate the relevancy of existing test data.
{"title":"Scaling Analysis of AP600 Long Term Cooling Performance","authors":"M. G. Ortiz, Constance E. Nielson, Laura Teerlink","doi":"10.1115/imece1997-0611","DOIUrl":"https://doi.org/10.1115/imece1997-0611","url":null,"abstract":"\u0000 Westinghouse’s AP600 thermohydraulic design, with passive safety features, poses new challenges to computer simulation and analyses, the design of experimental test facilities to represent it, and to the proper interpretation of the data from these facilities. The conventional approach of modeling the reactor thermohydraulic system as a closed, steady state, natural circulation loop from which non-dimensional groups of parameters can be derived and used in the design of integral tests, is limited and can not capture the abrupt time-varying open system nature of the new design.\u0000 A rigorous and systematic, eight-step methodology has been developed to scale and interpret the results from three different integral test facilities, and to relate them to the full scale plant. In this paper, the aforementioned scaling methodology is applied to the analysis of the long term cooling phase of the AP600 behavior. This long term cooling phase, which appears independent of the initiating event, is divided for its analysis into two sub-phases. A first sub-phase dominated by the draining of the large In-Containment Refueling Water Storage Tank through the primary systems, and a second sub-phase characterized by the quasi-steady recirculation of coolant through the reactor vessel and the outside of the primary system. The analysis shows that with a few verifiable assumptions one can determine the key parameters and non-dimensional groups that govern the behavior in either of these sub-phases. One then uses these parameters and non-dimensional groups to evaluate the relevancy of existing test data.","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":"83989612","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}
This paper presents a numerical study on heat transfer by laminar flows in a two-dimensional channel. The channel is filled with a porous material which is assumed to be non-uniform across the channel, especially near the channel walls where the permeability of the porous material increases drastically to simulate the complexity at the interface between the porous media and the solid walls. Degrees of the non-uniformity of the porous material are described in this study by a parameter β. Heat sources, either constant temperature or constant heat flux, are located on the channel walls, so that heat transfer from the sources into the channel flow is greatly affected by the condition at the wall-porous media interfaces. The temperature distribution near the heat source region and the Nusselt numbers are calculated for different β values. The sensitivity of the non-uniformity of the porous media near the channel walls on the overall heat transfer is assessed. It is found that a higher non-uniformity of the porous material near the channel wall can actually enhance the heat transfer from the sources to the channel flow, as the flow velocity near the wall has been significantly increased due to the higher permeability of the porous material in the near-wall region.
{"title":"Heat Transfer in Laminar Channel Flow With a Non-Uniform Porous Media","authors":"X. Huang, W. Qin, C. Y. Liu, K. Toh","doi":"10.1115/imece1997-0615","DOIUrl":"https://doi.org/10.1115/imece1997-0615","url":null,"abstract":"\u0000 This paper presents a numerical study on heat transfer by laminar flows in a two-dimensional channel. The channel is filled with a porous material which is assumed to be non-uniform across the channel, especially near the channel walls where the permeability of the porous material increases drastically to simulate the complexity at the interface between the porous media and the solid walls. Degrees of the non-uniformity of the porous material are described in this study by a parameter β. Heat sources, either constant temperature or constant heat flux, are located on the channel walls, so that heat transfer from the sources into the channel flow is greatly affected by the condition at the wall-porous media interfaces. The temperature distribution near the heat source region and the Nusselt numbers are calculated for different β values. The sensitivity of the non-uniformity of the porous media near the channel walls on the overall heat transfer is assessed. It is found that a higher non-uniformity of the porous material near the channel wall can actually enhance the heat transfer from the sources to the channel flow, as the flow velocity near the wall has been significantly increased due to the higher permeability of the porous material in the near-wall region.","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":"84235520","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 thermal stratification in two auxiliary piping systems of the Advanced Passive 600 mW Pressurized Water Reactor (AP600).
{"title":"CFD Analysis of Thermal Mixing and Stratification in AP600 Auxiliary Lines","authors":"A. Harkness, R. Schwirian, P. Dymáček","doi":"10.1115/imece1997-0614","DOIUrl":"https://doi.org/10.1115/imece1997-0614","url":null,"abstract":"\u0000 Computational Fluid Dynamics (CFD) was used to evaluate thermal stratification in two auxiliary piping systems of the Advanced Passive 600 mW Pressurized Water Reactor (AP600).","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":"73651488","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}
G. McCreery, C. Kullberg, R. Schultz, T. Yonomoto, Y. Anoda
Data from a model Passive Residual Heat Removal (PRHR) system operated in the ROSA facility during simulated loss-of-coolant accidents were compared with heat transfer calculations using a one-dimensional steady-state computer model. The calculations agree reasonably well with the data. The PRHR consists of a “C” shaped tube bundle submerged in a large tank filled with water. The calculations show that subcooled nucleate boiling occurred near the heat exchanger inlet. The boiling region was followed by turbulent and then laminar natural convection regions for horizontal and vertical the tubes. Transition boiling, with intermittently occuring patches of steam attached to the tubes, may have occurred near the inlet for the higher heat transfer tests, but did not significantly affect the overall heat transfer process.
{"title":"Heat Transfer Modeling of the LSTF Passive Residual Heat Removal System","authors":"G. McCreery, C. Kullberg, R. Schultz, T. Yonomoto, Y. Anoda","doi":"10.1115/imece1997-0612","DOIUrl":"https://doi.org/10.1115/imece1997-0612","url":null,"abstract":"\u0000 Data from a model Passive Residual Heat Removal (PRHR) system operated in the ROSA facility during simulated loss-of-coolant accidents were compared with heat transfer calculations using a one-dimensional steady-state computer model. The calculations agree reasonably well with the data. The PRHR consists of a “C” shaped tube bundle submerged in a large tank filled with water. The calculations show that subcooled nucleate boiling occurred near the heat exchanger inlet. The boiling region was followed by turbulent and then laminar natural convection regions for horizontal and vertical the tubes. Transition boiling, with intermittently occuring patches of steam attached to the tubes, may have occurred near the inlet for the higher heat transfer tests, but did not significantly affect the overall heat transfer process.","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":"87622863","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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