Venesa Watson, Edita Bajramovic, Xinxin Lou, K. Waedt
Working Group WGA9 of IEC SC45A (Nuclear I&C and ES), has recently completed a further working draft (WD) of the new IEC 63096 (unpublished) standard, aptly entitled Nuclear Power Plants – Instrumentation, Control and Electrical Systems – Security Controls. IEC 63096 specifically focuses on the selection and application of computer security controls for computer-based I&C and ES systems. This standard follows the commonly accepted ISO/IEC 27000 series security objectives of confidentiality, integrity and availability, and borrows and expands the objectives and implementation guidance from ISO/IEC 27002, while considering recommendations on sector-specific standards by ISO/IEC 27009. In addition, this guidance introduces a security grading, as well as lifecycle phase-specific controls. The grading aligns with the stringency of security controls, starting with Baseline Requirements (BR), Security Degree S3 and up to S1 (from lowest to highest degree). The lifecycle phase concerns the I&C development (D), project engineering (E) and operation and maintenance phases (O). This paper applies a sub-clause of IEC 63096 clause 15 (Supplier Relationships), to a programmable logic controller (PLC) that is typically used in power plants, to show the intended use of this standard and how it complements highest safety requirements in power plants. The Supplier Relationship clause concerns topics related to supply chain security, and is used to develop a use case example for the PLC. This example demonstrates how the controls and security degrees fits the implementation guidance from ISO/IEC 27002 and how they can be methodically applied to an I&C system.
{"title":"Example of Graded and Lifecycle Phase-Specific Security Controls for Nuclear I&C and EPS Use Cases","authors":"Venesa Watson, Edita Bajramovic, Xinxin Lou, K. Waedt","doi":"10.1115/ICONE26-81601","DOIUrl":"https://doi.org/10.1115/ICONE26-81601","url":null,"abstract":"Working Group WGA9 of IEC SC45A (Nuclear I&C and ES), has recently completed a further working draft (WD) of the new IEC 63096 (unpublished) standard, aptly entitled Nuclear Power Plants – Instrumentation, Control and Electrical Systems – Security Controls. IEC 63096 specifically focuses on the selection and application of computer security controls for computer-based I&C and ES systems. This standard follows the commonly accepted ISO/IEC 27000 series security objectives of confidentiality, integrity and availability, and borrows and expands the objectives and implementation guidance from ISO/IEC 27002, while considering recommendations on sector-specific standards by ISO/IEC 27009. In addition, this guidance introduces a security grading, as well as lifecycle phase-specific controls. The grading aligns with the stringency of security controls, starting with Baseline Requirements (BR), Security Degree S3 and up to S1 (from lowest to highest degree). The lifecycle phase concerns the I&C development (D), project engineering (E) and operation and maintenance phases (O). This paper applies a sub-clause of IEC 63096 clause 15 (Supplier Relationships), to a programmable logic controller (PLC) that is typically used in power plants, to show the intended use of this standard and how it complements highest safety requirements in power plants. The Supplier Relationship clause concerns topics related to supply chain security, and is used to develop a use case example for the PLC. This example demonstrates how the controls and security degrees fits the implementation guidance from ISO/IEC 27002 and how they can be methodically applied to an I&C system.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"130 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":"121931159","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}
The High Temperature Gas-cooled Reactor (HTGR) designed by Tsinghua university is under development in China. The electrical equipment in Pressure Vessel, such as magnetic bearings and helium circulator are operating in the high pressure helium environment. Design research of insulation property of the electrical device in helium under high pressure is necessary. Nevertheless, it is challenging to investigate by experimental technique. We propose a similarity law, converting the high pressure to a lower pressure to simplify the experimental conditions. Similarity theory of gas discharge is that two geometrically similar gaps with scaling coefficients of k, based on the same product number of pressure p and the gas length d, have the similar discharge characteristics. We research the validity of discharge similarity theory by simulation. A fluid model of direct-current discharge in helium atmosphere was established, referencing experimental results. And the discharge models of gaps were solved by finite-element method respectively. Four geometrically similar gaps were designed, the prototype gap is 2cm long and operating at a pressure of 600Pa while the pressure of three similar gaps are 300Pa, 200Pa, and 120Pa, and corresponding to the length of gaps are 4cm, 6cm and 10cm, with scaled-down factor k of 2, 3, and 5 respectively. The simulation results show that as long as the scaled-down factor of pressure k is less than 5 and the reduced length relation meets the condition p1d1 = p2d2, the discharge characteristics of two geometrically similar gaps are similar. As a result, it is realizable to predict the insulation property of electrical device in helium with similarity law.
{"title":"Application of Similarity Law in Electrical Device Design in Helium for High Temperature Gas-Cooled Reactor","authors":"Xiaohua Chen, Y. Geng, Jie Wang","doi":"10.1115/ICONE26-82520","DOIUrl":"https://doi.org/10.1115/ICONE26-82520","url":null,"abstract":"The High Temperature Gas-cooled Reactor (HTGR) designed by Tsinghua university is under development in China. The electrical equipment in Pressure Vessel, such as magnetic bearings and helium circulator are operating in the high pressure helium environment. Design research of insulation property of the electrical device in helium under high pressure is necessary. Nevertheless, it is challenging to investigate by experimental technique. We propose a similarity law, converting the high pressure to a lower pressure to simplify the experimental conditions. Similarity theory of gas discharge is that two geometrically similar gaps with scaling coefficients of k, based on the same product number of pressure p and the gas length d, have the similar discharge characteristics. We research the validity of discharge similarity theory by simulation. A fluid model of direct-current discharge in helium atmosphere was established, referencing experimental results. And the discharge models of gaps were solved by finite-element method respectively. Four geometrically similar gaps were designed, the prototype gap is 2cm long and operating at a pressure of 600Pa while the pressure of three similar gaps are 300Pa, 200Pa, and 120Pa, and corresponding to the length of gaps are 4cm, 6cm and 10cm, with scaled-down factor k of 2, 3, and 5 respectively. The simulation results show that as long as the scaled-down factor of pressure k is less than 5 and the reduced length relation meets the condition p1d1 = p2d2, the discharge characteristics of two geometrically similar gaps are similar. As a result, it is realizable to predict the insulation property of electrical device in helium with similarity law.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"13 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":"127893062","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}
The Steam Generator Tube Rupture (SGTR) postulated event constitutes one of the most hazardous safety issues for Gen IV pool reactors, cooled by heavy liquid metals. This accidental scenario is characterized by quick water flashing when in contact with primary coolant liquid metal, causing pressure wave propagation, cover gas pressurization in the reactor main vessel as well as possible tube rupture propagation, vapour dragged through the core, oxides precipitation and consequent slugs and plugs formation. The design phase of Gen IV MYRRHA reactor addressed the SGTR scenario issues in the framework of MAXSIMA project, supported by the European Commission. This research activity was fully executed at ENEA CR Brasimone, where a new test section was designed, assembled, instrumented and implemented in the large scale pool facility CIRCE. It was supported by the execution of preliminary and detailed pre-tests analysis performed adopting SIMME-III and -IV code, respectively. This paper details the test section main features, able to host four full scale portions (each one constituted by 31 tubes) of the MYRRHA Primary Heat eXchanger (PHX), for carrying out four independent SGTR experiments. A couple of tests investigated the tube rupture at middle position between two spacer grids of the bundle. The other two tests analysed instead the rupture near the bottom tube plate. Auxiliary systems were adopted for reaching primary (Lead Bismuth Eutectic alloy, LBE) and secondary (water) coolant initial conditions in accordance with MYRRHA design. Water was injected at 16 bar and 200°C in LBE at 350°C under an argon cover gas at about atmospheric pressure. The experimental results of the first test (middle rupture), in terms of CIRCE vessel pressurization, vapour flow path through tube bundle and tubes deformation, are presented. The post-test analysis was performed by SIMMER-IV code adopting the 3D Cartesian code version. The whole main vessel of CIRCE facility and implemented test section were modelled conserving heights and flowing areas. The experimental initial conditions were successfully matched by numerical results as well as the vessel pressurization and temperature time trends in the tube bundle following the SGTR. An important engineering feedback, for MYRRHA designer, was the evidence of rupture propagation absence. Moreover, the effectiveness of implemented safety devices, rupture disks, was evaluated and characterized for pressure relief feedbacks. A wide series of high quality measured data (pressure, temperature, strain and mass flow rate) was acquired and constitutes a database enlargement for future codes validation and possible new model development.
{"title":"Experimental and Numerical Analysis of Steam Generator Tube Rupture Event for MYRRHA Reactor in CIRCE Facility With SIMMER-IV Code","authors":"A. Pesetti, M. Tarantino, N. Forgione","doi":"10.1115/ICONE26-82503","DOIUrl":"https://doi.org/10.1115/ICONE26-82503","url":null,"abstract":"The Steam Generator Tube Rupture (SGTR) postulated event constitutes one of the most hazardous safety issues for Gen IV pool reactors, cooled by heavy liquid metals. This accidental scenario is characterized by quick water flashing when in contact with primary coolant liquid metal, causing pressure wave propagation, cover gas pressurization in the reactor main vessel as well as possible tube rupture propagation, vapour dragged through the core, oxides precipitation and consequent slugs and plugs formation.\u0000 The design phase of Gen IV MYRRHA reactor addressed the SGTR scenario issues in the framework of MAXSIMA project, supported by the European Commission. This research activity was fully executed at ENEA CR Brasimone, where a new test section was designed, assembled, instrumented and implemented in the large scale pool facility CIRCE. It was supported by the execution of preliminary and detailed pre-tests analysis performed adopting SIMME-III and -IV code, respectively.\u0000 This paper details the test section main features, able to host four full scale portions (each one constituted by 31 tubes) of the MYRRHA Primary Heat eXchanger (PHX), for carrying out four independent SGTR experiments. A couple of tests investigated the tube rupture at middle position between two spacer grids of the bundle. The other two tests analysed instead the rupture near the bottom tube plate. Auxiliary systems were adopted for reaching primary (Lead Bismuth Eutectic alloy, LBE) and secondary (water) coolant initial conditions in accordance with MYRRHA design. Water was injected at 16 bar and 200°C in LBE at 350°C under an argon cover gas at about atmospheric pressure. The experimental results of the first test (middle rupture), in terms of CIRCE vessel pressurization, vapour flow path through tube bundle and tubes deformation, are presented.\u0000 The post-test analysis was performed by SIMMER-IV code adopting the 3D Cartesian code version. The whole main vessel of CIRCE facility and implemented test section were modelled conserving heights and flowing areas. The experimental initial conditions were successfully matched by numerical results as well as the vessel pressurization and temperature time trends in the tube bundle following the SGTR. An important engineering feedback, for MYRRHA designer, was the evidence of rupture propagation absence. Moreover, the effectiveness of implemented safety devices, rupture disks, was evaluated and characterized for pressure relief feedbacks.\u0000 A wide series of high quality measured data (pressure, temperature, strain and mass flow rate) was acquired and constitutes a database enlargement for future codes validation and possible new model development.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"22 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":"132649417","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}
The DEMOnstration Fusion power Plant (DEMO) will be a key step towards Fusion Power Plant technology. It represents the single step to a commercial fusion power plant, in charge of demonstrating the viability of relevant technologies. Indeed, the development of tokamak sub-systems has to take into account interface, structural and functional requirements and multi-physics issues that can be completely known only during the development of the design process. This leads to difficulties to be faced during the conceptual design, mainly related to the identification of the main requirements, the change management and the sub-system integration. The Systems Engineering approach aims to support the design and management of complex systems over their life-cycles, providing a systematic approach for the definition of customer needs and required functionality from the early stage of the design, as well as for the design synthesis and the system validation and verification. Among the tokamak sub-systems, the divertor is the one devoted to power exhaust management and represents, at the same time, one of the most challenging components, in terms of materials, technologies and manufacturing. Current design activities, conducted in the in the framework of EUROfusion Consortium are in a pre-conceptual phase. Despite the early design stage, a systems engineering approach is being applied as an integrated, interdisciplinary R&D effort. The paper therefore presents the modeling effort to the conceptual design of DEMO divertor aimed at identifying both system main functions and expected behavior, given the constraints imposed from either project requirement or from current viability of technological solutions. To allow for flexibility in design needed to explore the feasibility of different solutions at this pre-conceptual stage, the impact of possible changes in high level requirement or interfaces is also investigated. This is also achieved through the allocation of the requirements to the affected components and providing efficient traceability. Therefore, the proposed modelling approach is intended to support the whole divertor conceptual design stage, allowing for requirements identification, traceability and change management.
{"title":"Systems Engineering Approach for Pre-Conceptual Design of DEMO Divertor","authors":"D. Marzullo, D. Dongiovanni, J. You","doi":"10.1115/ICONE26-82421","DOIUrl":"https://doi.org/10.1115/ICONE26-82421","url":null,"abstract":"The DEMOnstration Fusion power Plant (DEMO) will be a key step towards Fusion Power Plant technology. It represents the single step to a commercial fusion power plant, in charge of demonstrating the viability of relevant technologies. Indeed, the development of tokamak sub-systems has to take into account interface, structural and functional requirements and multi-physics issues that can be completely known only during the development of the design process. This leads to difficulties to be faced during the conceptual design, mainly related to the identification of the main requirements, the change management and the sub-system integration. The Systems Engineering approach aims to support the design and management of complex systems over their life-cycles, providing a systematic approach for the definition of customer needs and required functionality from the early stage of the design, as well as for the design synthesis and the system validation and verification.\u0000 Among the tokamak sub-systems, the divertor is the one devoted to power exhaust management and represents, at the same time, one of the most challenging components, in terms of materials, technologies and manufacturing. Current design activities, conducted in the in the framework of EUROfusion Consortium are in a pre-conceptual phase. Despite the early design stage, a systems engineering approach is being applied as an integrated, interdisciplinary R&D effort.\u0000 The paper therefore presents the modeling effort to the conceptual design of DEMO divertor aimed at identifying both system main functions and expected behavior, given the constraints imposed from either project requirement or from current viability of technological solutions. To allow for flexibility in design needed to explore the feasibility of different solutions at this pre-conceptual stage, the impact of possible changes in high level requirement or interfaces is also investigated. This is also achieved through the allocation of the requirements to the affected components and providing efficient traceability. Therefore, the proposed modelling approach is intended to support the whole divertor conceptual design stage, allowing for requirements identification, traceability and change management.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"49 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":"122835321","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}
Studies are currently on-going on the cycle performance of Generation IV (Gen IV) Nuclear Power Plants (NPPs) for the purpose of determining optimum operating conditions for efficiency and economic reasons. For Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs), the cycle layout is predominantly driven by the choice of components, the component configuration and the coolant. The purpose of this paper to present and review the cycles currently being considered — the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR). In all cases, the cycles utilise helium as the coolant in a closed Brayton gas turbine configuration. Comparisons between the cycles are made for Design Point (DP) and Off-Design Point (ODP) analyses to emphasise the benefits and drawbacks of each cycle. The paper also talks about future trends which include higher Core Outlet Temperatures in excess of 1000 degrees Celsius and the proposal of a simplified cycle configuration which eliminates the need for the recuperator.
{"title":"A Review of Brayton Helium Gas Turbine Cycles for GFR and VHTR Generation IV Nuclear Power Plants","authors":"A. Gad-Briggs, P. Pilidis, T. Nikolaidis","doi":"10.1115/ICONE26-81681","DOIUrl":"https://doi.org/10.1115/ICONE26-81681","url":null,"abstract":"Studies are currently on-going on the cycle performance of Generation IV (Gen IV) Nuclear Power Plants (NPPs) for the purpose of determining optimum operating conditions for efficiency and economic reasons. For Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs), the cycle layout is predominantly driven by the choice of components, the component configuration and the coolant. The purpose of this paper to present and review the cycles currently being considered — the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR). In all cases, the cycles utilise helium as the coolant in a closed Brayton gas turbine configuration. Comparisons between the cycles are made for Design Point (DP) and Off-Design Point (ODP) analyses to emphasise the benefits and drawbacks of each cycle. The paper also talks about future trends which include higher Core Outlet Temperatures in excess of 1000 degrees Celsius and the proposal of a simplified cycle configuration which eliminates the need for the recuperator.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"1 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":"128947404","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}
High Temperature Reactor-Pebblebed Modules (HTR-PM) is a typical high-temperature gas cooled reactor (HTGR) [1]. Tritium is one of the most important radionuclides in reactors owing to its very harmful β-radiation and long half-life. In the HTR-PM, Silicon carbide (SiC) is the main barrier of triisotropic (TRISO) particles to prevent the diffusion of Tritium into the primary circuit [2]. When Tritium into the primary circuit and circulate to Steam Generator (SG), the Incoloy800H alloy is another important material to prevent the diffusion of Tritium into the secondary circuit [3]. When analyzing the source term of Tritium in HTR-PM primary and secondary circuit, it is important and necessary to know the diffusion behavior of Tritium in SiC and Incoloy800H and furthermore, the detail mechanism of diffusion is also essential, which could not be obtained from traditional phenomenological analysis and conservative estimation. In order to solve this challenge, a framework with ab-initio methods is established. In this paper, the detail theory of ab-initio theory and the actual usage in the calculation of the diffusion path, barrier energy are given firstly. And then, the most physical path and the minimum energy barrier will be determined, which can be considered as the diffusion activation energy. The calculated results of activation energy of Tritium in SiC and Incoloy800H are 0.442eV and 0.757eV respectively. Furthermore, the theoretical results are compared with the experimental data, and it is found that both are in agreement with each other. These results are very helpful for understanding the diffusion behaviors of Tritium in HTR-PM materials and can be used to guide the tritium source term analysis in HTR-PM, which are first studied from a micro perspective.
{"title":"First-Principles Studies of Diffusion Behaviors of Tritium in HTR-PM Materials: From Framework to Preliminary Result","authors":"Chao Fang, Wenyi Wang, Hongyu Chen, C. Li","doi":"10.1115/ICONE26-81481","DOIUrl":"https://doi.org/10.1115/ICONE26-81481","url":null,"abstract":"High Temperature Reactor-Pebblebed Modules (HTR-PM) is a typical high-temperature gas cooled reactor (HTGR) [1]. Tritium is one of the most important radionuclides in reactors owing to its very harmful β-radiation and long half-life. In the HTR-PM, Silicon carbide (SiC) is the main barrier of triisotropic (TRISO) particles to prevent the diffusion of Tritium into the primary circuit [2]. When Tritium into the primary circuit and circulate to Steam Generator (SG), the Incoloy800H alloy is another important material to prevent the diffusion of Tritium into the secondary circuit [3]. When analyzing the source term of Tritium in HTR-PM primary and secondary circuit, it is important and necessary to know the diffusion behavior of Tritium in SiC and Incoloy800H and furthermore, the detail mechanism of diffusion is also essential, which could not be obtained from traditional phenomenological analysis and conservative estimation. In order to solve this challenge, a framework with ab-initio methods is established. In this paper, the detail theory of ab-initio theory and the actual usage in the calculation of the diffusion path, barrier energy are given firstly. And then, the most physical path and the minimum energy barrier will be determined, which can be considered as the diffusion activation energy. The calculated results of activation energy of Tritium in SiC and Incoloy800H are 0.442eV and 0.757eV respectively. Furthermore, the theoretical results are compared with the experimental data, and it is found that both are in agreement with each other. These results are very helpful for understanding the diffusion behaviors of Tritium in HTR-PM materials and can be used to guide the tritium source term analysis in HTR-PM, which are first studied from a micro perspective.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"26 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":"121340197","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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