For over 60 years the Advisory Committee on Reactor Safeguards (ACRS) has had a continuing statutory responsibility for providing independent reviews of, and advising on, the safety of proposed or existing reactor facilities and the adequacy of proposed reactor safety standards in the United States. This paper discusses the role of the Committee as it has evolved during its more than 60 years of history, noting some of its significant contributions to reactor safety.
{"title":"ACRS’ Enduring Legacy Contributing to Reactor Safety","authors":"Hossein P. Nourbakhsh","doi":"10.1115/ICONE26-82275","DOIUrl":"https://doi.org/10.1115/ICONE26-82275","url":null,"abstract":"For over 60 years the Advisory Committee on Reactor Safeguards (ACRS) has had a continuing statutory responsibility for providing independent reviews of, and advising on, the safety of proposed or existing reactor facilities and the adequacy of proposed reactor safety standards in the United States. This paper discusses the role of the Committee as it has evolved during its more than 60 years of history, noting some of its significant contributions to reactor safety.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"12 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125553960","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
T. Sugawara, Y. Eguchi, K. Tsujimoto, H. Obayashi, H. Iwamoto, H. Matsuda
Engineering feasibility of the beam window is one of the design issues in the accelerator-driven system (ADS). This study aims to perform the coupled analysis for the feasible beam window concept. To mitigate the design condition, namely to reduce the required proton beam current, subcriticality adjustment rod (SAR) was installed to the ADS core. The burnup analysis was performed for the ADS core with SAR and the results indicated that the maximum proton beam current during the burnup cycle was reduced from 20 to 13.5 mA. Based on the burnup analysis result, the coupled analysis; particle transport, thermal hydraulics and structural analyses, was performed. As the final result, the following design; the hemisphere shape, the outer radius = 180 mm, the thickness at the top of the beam window = 1.5 mm, and the factor of safety for the buckling = 3.8, was presented. The buckling pressure was almost same as the previous one and more feasible beam window concept was presented through this study.
{"title":"Design Study of Beam Window for Accelerator-Driven System With Subcriticality Adjustment Rod","authors":"T. Sugawara, Y. Eguchi, K. Tsujimoto, H. Obayashi, H. Iwamoto, H. Matsuda","doi":"10.1115/ICONE26-81233","DOIUrl":"https://doi.org/10.1115/ICONE26-81233","url":null,"abstract":"Engineering feasibility of the beam window is one of the design issues in the accelerator-driven system (ADS). This study aims to perform the coupled analysis for the feasible beam window concept. To mitigate the design condition, namely to reduce the required proton beam current, subcriticality adjustment rod (SAR) was installed to the ADS core. The burnup analysis was performed for the ADS core with SAR and the results indicated that the maximum proton beam current during the burnup cycle was reduced from 20 to 13.5 mA.\u0000 Based on the burnup analysis result, the coupled analysis; particle transport, thermal hydraulics and structural analyses, was performed. As the final result, the following design; the hemisphere shape, the outer radius = 180 mm, the thickness at the top of the beam window = 1.5 mm, and the factor of safety for the buckling = 3.8, was presented. The buckling pressure was almost same as the previous one and more feasible beam window concept was presented through this study.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"21 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123010122","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Based on ASME Boilers and Pressure Vessels Code the major fracture mechanic analysis is limited to protection of class 1 components to brittle fracture. All the Operators of future plants have to enlarge the scope of these analyses to different concepts, at design or operation stage: - brittle and ductile analysis of hypothetical large flaw - leak before break approach - break exclusion concept - incredibility of failure of high integrity components - end of fabrication acceptable defect - in-service inspection performance - acceptable standards in operation - Long Term Operation (LTO) All these requirements needs a procedure, an analysis method with material properties and criteria. After a short overview of each topic, the paper will present how RCC-M, RSE-M French Codes and ASME III and XI take care of all these new modern regulatory requirements.
{"title":"New Needs of Fracture Mechanic Analysis at Design and Operation Level: Status of French Nuclear Mechanical Codes","authors":"C. Faidy","doi":"10.1115/ICONE26-81096","DOIUrl":"https://doi.org/10.1115/ICONE26-81096","url":null,"abstract":"Based on ASME Boilers and Pressure Vessels Code the major fracture mechanic analysis is limited to protection of class 1 components to brittle fracture.\u0000 All the Operators of future plants have to enlarge the scope of these analyses to different concepts, at design or operation stage:\u0000 - brittle and ductile analysis of hypothetical large flaw\u0000 - leak before break approach\u0000 - break exclusion concept\u0000 - incredibility of failure of high integrity components\u0000 - end of fabrication acceptable defect\u0000 - in-service inspection performance\u0000 - acceptable standards in operation\u0000 - Long Term Operation (LTO)\u0000 All these requirements needs a procedure, an analysis method with material properties and criteria.\u0000 After a short overview of each topic, the paper will present how RCC-M, RSE-M French Codes and ASME III and XI take care of all these new modern regulatory requirements.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"269 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132951684","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Fire hazard is an important issue for the safety of nuclear power plants: the main internal hazard in terms of frequency, and probably one the most significant with regards to the design costs. AFCEN is publishing in 2018 a new code for fire protection of new built PWR nuclear plants, so-called RCC-F. This code is an evolution of the former ETC-F code which has been applied to different EPR plants under construction (Flamanville 3 (FA3, France), Hinkley Point C (HPC, United Kingdom), Taïshan (TSN, China)). The RCC-F code presents significant enhancement and evolutions resulting from eight years of work by the AFCEN dedicated sub-committee, involving a panel of contributors from the nuclear field. It is now opened to any type of PWR (Pressurized Water Reactor) type of nuclear power plants and not any longer limited to EPR (European Pressurized Reactor) plants. It can potentially be adapted to other light water concepts. Its objective is to help engineers design the fire prevention and protection scheme, systems and equipment with regards to the safety case and the defense in depth taking into account the French and European experience in the field. It deals also with the national regulations, with two appendices dedicated to French and British regulations respectively. The presentation gives an overview of the code specifications and focuses on the significant improvements.
{"title":"AFCEN RCC-F: A New Standard for the Fire Protection Design of New Built Light Water Nuclear Power Plants","authors":"B. Gautier, M. Cesbron, Richard Tulinski","doi":"10.1115/ICONE26-81893","DOIUrl":"https://doi.org/10.1115/ICONE26-81893","url":null,"abstract":"Fire hazard is an important issue for the safety of nuclear power plants: the main internal hazard in terms of frequency, and probably one the most significant with regards to the design costs. AFCEN is publishing in 2018 a new code for fire protection of new built PWR nuclear plants, so-called RCC-F. This code is an evolution of the former ETC-F code which has been applied to different EPR plants under construction (Flamanville 3 (FA3, France), Hinkley Point C (HPC, United Kingdom), Taïshan (TSN, China)). The RCC-F code presents significant enhancement and evolutions resulting from eight years of work by the AFCEN dedicated sub-committee, involving a panel of contributors from the nuclear field. It is now opened to any type of PWR (Pressurized Water Reactor) type of nuclear power plants and not any longer limited to EPR (European Pressurized Reactor) plants. It can potentially be adapted to other light water concepts. Its objective is to help engineers design the fire prevention and protection scheme, systems and equipment with regards to the safety case and the defense in depth taking into account the French and European experience in the field. It deals also with the national regulations, with two appendices dedicated to French and British regulations respectively. The presentation gives an overview of the code specifications and focuses on the significant improvements.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"263 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"117101419","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
E. Osigwe, A. Gad-Briggs, T. Nikolaidis, P. Pilidis, S. Sampath
With renewed interest in nuclear power to meet the world’s future energy demand, the Generation IV nuclear reactors are the next step in the deployment of nuclear power generation. However, for the potentials of these nuclear reactor designs to be fully realized, its suitability, when coupled with different configurations of closed-cycle gas turbine power conversion systems, have to be explored and performance compared for various possible working fluids over a range of operating pressures and temperatures. The purpose of this paper is to carry out performance analysis at the design and off-design conditions for a Generation IV nuclear-powered reactor in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as working fluid in the cycle. This analysis is demonstrated in GTACYSS; a performance and preliminary design code developed by the authors for closed-cycle gas turbine simulations. The results obtained shows that the choice of working fluid controls the range of cycle operating pressures, temperatures and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO2 due to the behavior CO2 below its critical conditions.
{"title":"Performance Analysis of Generation IV Nuclear Reactor Power Plant Using CO2 and N2: Case Study of a Recuperated Brayton Gas Turbine Cycle","authors":"E. Osigwe, A. Gad-Briggs, T. Nikolaidis, P. Pilidis, S. Sampath","doi":"10.1115/ICONE26-81337","DOIUrl":"https://doi.org/10.1115/ICONE26-81337","url":null,"abstract":"With renewed interest in nuclear power to meet the world’s future energy demand, the Generation IV nuclear reactors are the next step in the deployment of nuclear power generation. However, for the potentials of these nuclear reactor designs to be fully realized, its suitability, when coupled with different configurations of closed-cycle gas turbine power conversion systems, have to be explored and performance compared for various possible working fluids over a range of operating pressures and temperatures. The purpose of this paper is to carry out performance analysis at the design and off-design conditions for a Generation IV nuclear-powered reactor in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as working fluid in the cycle. This analysis is demonstrated in GTACYSS; a performance and preliminary design code developed by the authors for closed-cycle gas turbine simulations. The results obtained shows that the choice of working fluid controls the range of cycle operating pressures, temperatures and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO2 due to the behavior CO2 below its critical conditions.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"61 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124722099","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Songlin Liu, Xuebin Ma, K. Jiang, Min Li, Xiaokang Zhang
The Chinese Fusion Engineering Testing Reactor (CFETR) will be operated in two phases. Phase I focuses on fusion power Pfusion = 200 MW, fusion power gain Qplasma = 1 – 5, tritium breeding ratio TBR>1.0, neutron DPA requirement ∼10 dpa. Phase II emphasizes DEMO validation, which means Qplasma > 10, Pfusion > 1 GW, e.g. 1.5 GW. It is required that one blanket design can cover the operation of both phases of CFETR from the viewpoint of saving construction cost and reducing waste inventory. However, fusion power in Phase-II is 4–6.5 times larger than those in Phase-I, and this also causes the great challenge facing the thermal-hydraulics design of the blanket. A new version of water cooled ceramic breeder (WCCB) blanket for both phases is proposed for CFETR, based on a trade-off considering on TBR, release tritium temperature in breeder zone, and removal heat capability of coolant. This design continues to employ the mixed breeder of Li2TiO3 and Be12Ti as tritium breeder and primary neutron multiplier, and a few Be as supplement of multiplying neutrons, Reduced Activation Ferritic/Martensitic steel as structural material, tungsten as plasma facing material. Pressurized water of 15.5 MPa is chosen as coolant with 285 °C inlet/325 °C outlet temperature. Main design change is that it employs two independent coolant systems in the blanket cooling components. For Phase I, one coolant system is only used and hoped to improve the breeder zone temperature higher than tritium release temperature. For Phase II, all of two coolant systems are put into using to ensure the material temperature less than the allowable limit. In this paper, the WCCB blanket design work is presented and its feasibility is investigated from the aspect of neutronics and thermo-hydraulics.
{"title":"Conceptual Design of the Water Cooled Breeder Blanket for Both Phases of CFETR","authors":"Songlin Liu, Xuebin Ma, K. Jiang, Min Li, Xiaokang Zhang","doi":"10.1115/ICONE26-81816","DOIUrl":"https://doi.org/10.1115/ICONE26-81816","url":null,"abstract":"The Chinese Fusion Engineering Testing Reactor (CFETR) will be operated in two phases. Phase I focuses on fusion power Pfusion = 200 MW, fusion power gain Qplasma = 1 – 5, tritium breeding ratio TBR>1.0, neutron DPA requirement ∼10 dpa. Phase II emphasizes DEMO validation, which means Qplasma > 10, Pfusion > 1 GW, e.g. 1.5 GW. It is required that one blanket design can cover the operation of both phases of CFETR from the viewpoint of saving construction cost and reducing waste inventory. However, fusion power in Phase-II is 4–6.5 times larger than those in Phase-I, and this also causes the great challenge facing the thermal-hydraulics design of the blanket. A new version of water cooled ceramic breeder (WCCB) blanket for both phases is proposed for CFETR, based on a trade-off considering on TBR, release tritium temperature in breeder zone, and removal heat capability of coolant. This design continues to employ the mixed breeder of Li2TiO3 and Be12Ti as tritium breeder and primary neutron multiplier, and a few Be as supplement of multiplying neutrons, Reduced Activation Ferritic/Martensitic steel as structural material, tungsten as plasma facing material. Pressurized water of 15.5 MPa is chosen as coolant with 285 °C inlet/325 °C outlet temperature. Main design change is that it employs two independent coolant systems in the blanket cooling components. For Phase I, one coolant system is only used and hoped to improve the breeder zone temperature higher than tritium release temperature. For Phase II, all of two coolant systems are put into using to ensure the material temperature less than the allowable limit. In this paper, the WCCB blanket design work is presented and its feasibility is investigated from the aspect of neutronics and thermo-hydraulics.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"30 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129325454","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
AFCEN is a French Standard Development Organization which publishes codes for design, construction and in-service inspection rules for Pressurized Water Reactors. The fields covered by theses codes are mechanical components, in-service surveillance of mechanical components, electrical equipments, nuclear fuel, civil works and fire protection. AFCEN was initially founded by electric utility EDF and nuclear steam supply system manufacturer FRAMATOME. AFCEN has more than 60 institutional members, representing more than 650 experts who contribute to the development and continuous improvement of codes. The RCC-C code, which is dedicated to PWR fuel assemblies and associated core components, set forth generic requirements to be fulfilled by the suppliers and by the manufacturers for the design justifications and for the manufacturing and inspection operations of PWR fuel assemblies and rod cluster control assemblies. The RCC-C is intended to be used in the frame of contractual relations between a customer (nuclear operator) and a nuclear fuel supplier. The first edition was published in 1981. Over the years, many changes have been made to the original text but the structure hasn’t been much modified. Because of this, the text was becoming less coherent for the users and was lacking also minimal explanations. A redesign of the code was scheduled for the 2015 edition to address those problems. With the involvement of fuel vendors FRAMATOME, WESTINGHOUSE, and French nuclear operator EDF, the text was restructured and clarified. New requirements were implemented and the set of both design and manufacturing rules was strengthened to reflect fuel vendors’ practices and operator expectations. This article explains the main modifications that were implemented since the 2015 edition, and also outlines the prospects for future changes taking into account the latest regulatory requirements and evolutions of the industrial practices.
{"title":"Main Evolutions of the RCC-C Design and Construction Code for Fuel Assemblies Since 2015","authors":"M. Ton-That, C. Vauglin, G. Trillon","doi":"10.1115/ICONE26-81436","DOIUrl":"https://doi.org/10.1115/ICONE26-81436","url":null,"abstract":"AFCEN is a French Standard Development Organization which publishes codes for design, construction and in-service inspection rules for Pressurized Water Reactors. The fields covered by theses codes are mechanical components, in-service surveillance of mechanical components, electrical equipments, nuclear fuel, civil works and fire protection. AFCEN was initially founded by electric utility EDF and nuclear steam supply system manufacturer FRAMATOME. AFCEN has more than 60 institutional members, representing more than 650 experts who contribute to the development and continuous improvement of codes.\u0000 The RCC-C code, which is dedicated to PWR fuel assemblies and associated core components, set forth generic requirements to be fulfilled by the suppliers and by the manufacturers for the design justifications and for the manufacturing and inspection operations of PWR fuel assemblies and rod cluster control assemblies. The RCC-C is intended to be used in the frame of contractual relations between a customer (nuclear operator) and a nuclear fuel supplier.\u0000 The first edition was published in 1981. Over the years, many changes have been made to the original text but the structure hasn’t been much modified. Because of this, the text was becoming less coherent for the users and was lacking also minimal explanations. A redesign of the code was scheduled for the 2015 edition to address those problems. With the involvement of fuel vendors FRAMATOME, WESTINGHOUSE, and French nuclear operator EDF, the text was restructured and clarified. New requirements were implemented and the set of both design and manufacturing rules was strengthened to reflect fuel vendors’ practices and operator expectations.\u0000 This article explains the main modifications that were implemented since the 2015 edition, and also outlines the prospects for future changes taking into account the latest regulatory requirements and evolutions of the industrial practices.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"24 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127473217","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
M. Angelucci, I. Piazza, M. Tarantino, R. Marinari, G. Polazzi, V. Sermenghi
An experimental campaign was performed on a non-uniformly heated 19-pins wire-spaced fuel pin bundle simulator, cooled by Heavy Liquid Metal and installed in the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The experimental tests concerned mass flow rate transition of the primary coolant from forced to natural circulation, with fuel pin bundle simulator characterized by non-uniform power distribution. The main objective of the performed experimental campaign was to perform integral system and local thermal-hydraulic analysis, in particular to investigate the flow in different flow regimes and specifically the transition from forced to natural circulation flow and, more specifically, analyze the behavior of the 19-pins wire-spaced fuel pin simulator (FPS) during such transient. Indeed, the performed test were characterized by non-uniform heating of the bundle (i.e. just some pins switched on), so the effects of this non-uniformity on the local temperatures and on the overall system behavior was evaluated. A deep investigation on the local temperature distribution was performed thanks to the accurate instrumentation provided in the bundle (67 thermocouples). For instance, in some cases, the wall temperatures relative to pins switched off remained below the relative sub-channel temperature, depending on the heating distribution. The obtained experimental data provided useful information for the characterization of the bundle and the computation of the heat transfer coefficient. Moreover, the collected system data can be helpful for STH codes validation, whereas the local fuel bundle data, especially the ones from dissymmetric tests can be useful for the qualification and benchmarking of CFD codes and coupled STH/CFD methods for HLM systems.
{"title":"Experimental Tests With Non-Uniformly Heated 19-Pins Fuel Bundle Cooled by HLM","authors":"M. Angelucci, I. Piazza, M. Tarantino, R. Marinari, G. Polazzi, V. Sermenghi","doi":"10.1115/ICONE26-81216","DOIUrl":"https://doi.org/10.1115/ICONE26-81216","url":null,"abstract":"An experimental campaign was performed on a non-uniformly heated 19-pins wire-spaced fuel pin bundle simulator, cooled by Heavy Liquid Metal and installed in the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The experimental tests concerned mass flow rate transition of the primary coolant from forced to natural circulation, with fuel pin bundle simulator characterized by non-uniform power distribution. The main objective of the performed experimental campaign was to perform integral system and local thermal-hydraulic analysis, in particular to investigate the flow in different flow regimes and specifically the transition from forced to natural circulation flow and, more specifically, analyze the behavior of the 19-pins wire-spaced fuel pin simulator (FPS) during such transient. Indeed, the performed test were characterized by non-uniform heating of the bundle (i.e. just some pins switched on), so the effects of this non-uniformity on the local temperatures and on the overall system behavior was evaluated. A deep investigation on the local temperature distribution was performed thanks to the accurate instrumentation provided in the bundle (67 thermocouples). For instance, in some cases, the wall temperatures relative to pins switched off remained below the relative sub-channel temperature, depending on the heating distribution. The obtained experimental data provided useful information for the characterization of the bundle and the computation of the heat transfer coefficient. Moreover, the collected system data can be helpful for STH codes validation, whereas the local fuel bundle data, especially the ones from dissymmetric tests can be useful for the qualification and benchmarking of CFD codes and coupled STH/CFD methods for HLM systems.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"25 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129923640","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Molten salts were widely used in nuclear and solar power field due to the excellent heat transfer and storage. Molten fluoride salts were selected as primary and secondary coolants in the Molten Salt Reactor Experiment (MSRE) developed by Oak Ridge National Laboratory (ORNL). Therefore, it is dramatically important to study the physical and chemical properties of molten fluoride salts that impact on the design of reactor core and thermohydraulics. The molecular structure directly determines the physical and chemical properties of matter, so it is also essential to study the structure of molten salts. Spectroscopy has been proven to be a very useful tool for investigating molten salts structures. However, the standard instrument is inapplicable for measurement of the high temperature molten salts, especially for molten fluoride salts. To obtain the ultraviolet-visible (UV-Vis) absorption spectra of molten salts at high temperature, an instrument was designed to study the structures of molten salts in situ. The instrument is mainly composed of a vertical pit furnace connecting with a glovebox and an assembled cuvette which can operate from room temperature up to 800°C. The assembled cuvette is made of Hastelloy C/N as the main body with a reverse ‘T’ contour and diamond or crystalline CaF2 etc. as the window plates, so it can withstand the corrosion produced by the sample and allow the interest light passing through. The effective spectral range of this instrument is from 200 to 1000 nm. Performances of the instrument are testified by spectral studies on water under room temperature and molten salts under high temperature.
{"title":"An Instrument Established for the High Temperature Measurement of Ultraviolet-Visible Absorption Spectra of Molten Fluoride Salt Behaving As Coolant in the Molten Salt Reactor","authors":"Hongtao Liu, Yiyang Liu, Tao Su","doi":"10.1115/ICONE26-82013","DOIUrl":"https://doi.org/10.1115/ICONE26-82013","url":null,"abstract":"Molten salts were widely used in nuclear and solar power field due to the excellent heat transfer and storage. Molten fluoride salts were selected as primary and secondary coolants in the Molten Salt Reactor Experiment (MSRE) developed by Oak Ridge National Laboratory (ORNL). Therefore, it is dramatically important to study the physical and chemical properties of molten fluoride salts that impact on the design of reactor core and thermohydraulics. The molecular structure directly determines the physical and chemical properties of matter, so it is also essential to study the structure of molten salts. Spectroscopy has been proven to be a very useful tool for investigating molten salts structures. However, the standard instrument is inapplicable for measurement of the high temperature molten salts, especially for molten fluoride salts. To obtain the ultraviolet-visible (UV-Vis) absorption spectra of molten salts at high temperature, an instrument was designed to study the structures of molten salts in situ. The instrument is mainly composed of a vertical pit furnace connecting with a glovebox and an assembled cuvette which can operate from room temperature up to 800°C. The assembled cuvette is made of Hastelloy C/N as the main body with a reverse ‘T’ contour and diamond or crystalline CaF2 etc. as the window plates, so it can withstand the corrosion produced by the sample and allow the interest light passing through. The effective spectral range of this instrument is from 200 to 1000 nm. Performances of the instrument are testified by spectral studies on water under room temperature and molten salts under high temperature.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"14 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134007940","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
“Begging the question” describes a situation in which the statement under examination is assumed to be true (i.e., the statement is used to support itself). Examples of this can be found in analysis reports that were prepared by analysts who are not mindful (or maybe uninformed) of the analysis criteria they’re required to fulfill. This is generally seen in analyses of anticipated operational occurrences (AOOs). AOOs are defined in Appendix A of 10 CFR §50 [1], and in ANS-N18.2-1973 [2], where they’re also known as American Nuclear Society (ANS) Condition II events. This standard [2] also defines more serious, Condition III and IV events. Analyses of AOOs, or ANS Condition II events are required to show that: (1) reactor coolant system (RCS) pressure will not exceed its safety limit, and (2) no fuel damage will be incurred, and (3) a more serious accident will not develop, unless there is a simultaneous occurrence of another, independent fault. The three requirements are often demonstrated by three different analyses, each of which is designed to yield conservative results with respect to one of the requirements. Accident analyses that are performed to demonstrate compliance with the first two requirements are relatively straightforward. They rely mostly upon the design of safety valves and the timing of reactor trips. “Begging the question” is seen in analyses that are designed to demonstrate compliance with the third requirement. This paper will describe how this logical fallacy has been applied in licensees’ accident analyses, and accepted by the NRC staff.
{"title":"“Begging the Question” in Licensing Basis Accident Analyses","authors":"Samuel Miranda","doi":"10.1115/ICONE26-81902","DOIUrl":"https://doi.org/10.1115/ICONE26-81902","url":null,"abstract":"“Begging the question” describes a situation in which the statement under examination is assumed to be true (i.e., the statement is used to support itself). Examples of this can be found in analysis reports that were prepared by analysts who are not mindful (or maybe uninformed) of the analysis criteria they’re required to fulfill.\u0000 This is generally seen in analyses of anticipated operational occurrences (AOOs). AOOs are defined in Appendix A of 10 CFR §50 [1], and in ANS-N18.2-1973 [2], where they’re also known as American Nuclear Society (ANS) Condition II events. This standard [2] also defines more serious, Condition III and IV events. Analyses of AOOs, or ANS Condition II events are required to show that:\u0000 (1) reactor coolant system (RCS) pressure will not exceed its safety limit, and\u0000 (2) no fuel damage will be incurred, and\u0000 (3) a more serious accident will not develop, unless there is a simultaneous occurrence of another, independent fault.\u0000 The three requirements are often demonstrated by three different analyses, each of which is designed to yield conservative results with respect to one of the requirements. Accident analyses that are performed to demonstrate compliance with the first two requirements are relatively straightforward. They rely mostly upon the design of safety valves and the timing of reactor trips.\u0000 “Begging the question” is seen in analyses that are designed to demonstrate compliance with the third requirement. This paper will describe how this logical fallacy has been applied in licensees’ accident analyses, and accepted by the NRC staff.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130581915","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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