A floating nuclear power plant (FNPP) with small modular reactor (SMR) is a combination of a civilian nuclear infrastructure and an offshore installation, which is defined as a floating nuclear facility. The article draws the lessons from studying of the engineer combination like Floating Production Storage and Offloading (FPSO) under the regulation of several government departments. It puts forward recommendations for license application and government regulation as follows in consideration with current license application for nuclear power plant and ship survey. A FNPP shall follow the requirements of construction, fueling and operation for civil nuclear installation combined with ship survey. Application is submitted to nuclear safety regulator for construction permit, while the design drawings shall be submitted to department of ship survey which checks the drawings whether meet the requirements of ship survey, considering some nuclear safety needs. The result of ship survey shall be represented in the safety analysis reports. The construction and important devices manufacturing shall be under the supervision of nuclear installation regulators and ship survey departments. In conclusion, National Nuclear Safety Administration (NNSA) and Maritime Safety Administration of the People’s Republic of China (MSA) shall establish united supervisory system for SMR on sea in China. It is suggested that NNSA is in charge of the overall safety of a FNPP, while MSA is responsible of the ship survey. The operator shall undertake obligation of a FNPP and evaluate the ship cooperating with experienced agency. It is suggested that government departments build the mutual recognition agreement of safety review. It is better to solve the vague questions by coordination.
{"title":"Study on Supervision Mode of Floating Nuclear Power Plant With Small Modular Reactor","authors":"W. Lei, G. Li, Min Rui, Yongkang Liu, Yang Jue","doi":"10.1115/ICONE26-82138","DOIUrl":"https://doi.org/10.1115/ICONE26-82138","url":null,"abstract":"A floating nuclear power plant (FNPP) with small modular reactor (SMR) is a combination of a civilian nuclear infrastructure and an offshore installation, which is defined as a floating nuclear facility. The article draws the lessons from studying of the engineer combination like Floating Production Storage and Offloading (FPSO) under the regulation of several government departments. It puts forward recommendations for license application and government regulation as follows in consideration with current license application for nuclear power plant and ship survey. A FNPP shall follow the requirements of construction, fueling and operation for civil nuclear installation combined with ship survey. Application is submitted to nuclear safety regulator for construction permit, while the design drawings shall be submitted to department of ship survey which checks the drawings whether meet the requirements of ship survey, considering some nuclear safety needs. The result of ship survey shall be represented in the safety analysis reports. The construction and important devices manufacturing shall be under the supervision of nuclear installation regulators and ship survey departments. In conclusion, National Nuclear Safety Administration (NNSA) and Maritime Safety Administration of the People’s Republic of China (MSA) shall establish united supervisory system for SMR on sea in China. It is suggested that NNSA is in charge of the overall safety of a FNPP, while MSA is responsible of the ship survey. The operator shall undertake obligation of a FNPP and evaluate the ship cooperating with experienced agency. It is suggested that government departments build the mutual recognition agreement of safety review. It is better to solve the vague questions by coordination.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"58 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":"134442786","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}
Cui Mao, Yi-Bao Liu, Liguo Zhang, J. Tong, Bingfeng Xia, Z. Yin
The efficient and accurate burn-up measurement of the spherical fuel element is the key component of the operation of the pebble bed high temperature gas-cooled reactor. The accuracy of the method that determine burnup by the activity of Cs-137 degrades due to operation characteristics of HTR-10. HTR-10, as an test reactor, operated on and off during the past years. It stayed shutdown more than power operation. In order to improve the measurement accuracy of Cs-137 activity and enhance the possibility to detect radionuclides with low activity, which can be used to correct the classic burnup assay method, a new measurement system is now discussed using anti-coincidence technology, which suppresses the Compton plateau. In this paper, Geant4 is used to simulate the anticoincidence measurement process taking high purity germanium γ-ray spectrometer as main detector and plastic scintillator as the annular detector. By analyzing the signal to noise ratio in different detection scenarios with all kinds of shape parameters of the annular detector, the annular detector with the best anti-coincidence effect are optimaized. The above research results provide an important theoretical basis for the construction of online burn-up measurement system based on anti-Compton technology.
{"title":"Simulation of HTR-10 Anti-Compton HPGE Gamma-Ray Spectrometer With Geant4","authors":"Cui Mao, Yi-Bao Liu, Liguo Zhang, J. Tong, Bingfeng Xia, Z. Yin","doi":"10.1115/ICONE26-81254","DOIUrl":"https://doi.org/10.1115/ICONE26-81254","url":null,"abstract":"The efficient and accurate burn-up measurement of the spherical fuel element is the key component of the operation of the pebble bed high temperature gas-cooled reactor. The accuracy of the method that determine burnup by the activity of Cs-137 degrades due to operation characteristics of HTR-10. HTR-10, as an test reactor, operated on and off during the past years. It stayed shutdown more than power operation. In order to improve the measurement accuracy of Cs-137 activity and enhance the possibility to detect radionuclides with low activity, which can be used to correct the classic burnup assay method, a new measurement system is now discussed using anti-coincidence technology, which suppresses the Compton plateau. In this paper, Geant4 is used to simulate the anticoincidence measurement process taking high purity germanium γ-ray spectrometer as main detector and plastic scintillator as the annular detector. By analyzing the signal to noise ratio in different detection scenarios with all kinds of shape parameters of the annular detector, the annular detector with the best anti-coincidence effect are optimaized. The above research results provide an important theoretical basis for the construction of online burn-up measurement system based on anti-Compton technology.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"44 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":"134498034","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}
Jin-Yang Li, L. Gu, C. Yao, Da-Wei Wang, Tianji Peng, Zhu Yanlei
Accelerator driven subcritical system (ADS) has been considered as an advanced nuclear waste transmutation facility with inherent safety feature. Nowadays, Institute of Modern Physics, Chinese Academy of Science (IMPCAS) has made a plan to research and develop ADS, which project was formally approved by the National Development and Reform Commission at the end of 2015. In order to pursue high transmutation rate of Minor Actinide (MA), a new concept design of the lead-bismuth cooled ADS is proposed in present work, in which a gravity-driven dense granular flow spallation target has been designed to substitute the traditional solid or liquid heavy metal spallation target. Sensitive parameters of the spallation target have been investigated, such as the radius and axial position of the spallation target, the size of heavy metal grains, the granular filling rate, the spallation neutron yield, and the distribution of neutron source. Furthermore, an optimized design of sub-critical reactor has been proposed with three zone loading pattern as a consequence of eliminating the power peak factor, which thermal power is 10MW with the operation cycle 600 EFPD. Finally, some important physical parameters in the coupling depletion system have been studied, including the effective multiplication factor k-eff, subcritical multiplication factor k-s, the change of beam current, time-dependent void and Doppler effect, fraction of effective delayed neutron, and the transmutation supporting ratio.
{"title":"Neutronic Study on a New Concept of Accelerator Driven Subcritical System in China","authors":"Jin-Yang Li, L. Gu, C. Yao, Da-Wei Wang, Tianji Peng, Zhu Yanlei","doi":"10.1115/ICONE26-81329","DOIUrl":"https://doi.org/10.1115/ICONE26-81329","url":null,"abstract":"Accelerator driven subcritical system (ADS) has been considered as an advanced nuclear waste transmutation facility with inherent safety feature. Nowadays, Institute of Modern Physics, Chinese Academy of Science (IMPCAS) has made a plan to research and develop ADS, which project was formally approved by the National Development and Reform Commission at the end of 2015. In order to pursue high transmutation rate of Minor Actinide (MA), a new concept design of the lead-bismuth cooled ADS is proposed in present work, in which a gravity-driven dense granular flow spallation target has been designed to substitute the traditional solid or liquid heavy metal spallation target. Sensitive parameters of the spallation target have been investigated, such as the radius and axial position of the spallation target, the size of heavy metal grains, the granular filling rate, the spallation neutron yield, and the distribution of neutron source. Furthermore, an optimized design of sub-critical reactor has been proposed with three zone loading pattern as a consequence of eliminating the power peak factor, which thermal power is 10MW with the operation cycle 600 EFPD. Finally, some important physical parameters in the coupling depletion system have been studied, including the effective multiplication factor k-eff, subcritical multiplication factor k-s, the change of beam current, time-dependent void and Doppler effect, fraction of effective delayed neutron, and the transmutation supporting ratio.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"94 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":"124174189","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}
Shiyan Sun, Youjie Zhang, Yanhua Zheng, X. Fang, Xiaoyong Yang
During the operation of the High Temperature Gas-cooled Reactor (HTGR), the hot-spot temperature in the reactor core must be lower than the maximum permissible temperature of the fuel elements and the materials of construction, so that the reactor kept safe. However, no fixed temperature-measuring devices can be set in a pebble-bed core. A special spherical temperature-measuring device is adopted to make sure it brings as small impact to the reactor operation as possible. There are several metal wires with different melting points inside. The graphite thermometric balls will be put onto the top of HTR-10 reactor core, and they record and reflect the highest temperature in different positions in the core when flowing in the pebble bed. Before the reactor core temperature-measuring experiment of HTR-10, we must study the heat transfer characteristics of the graphite thermometric sphere to find out the relationship of the melting conditions and the temperature in the reactor core. A 3-D model of the graphite thermometric ball is established, and CFD method is adopted to research and figure out the thermal equilibrium time and temperature difference between the metal wires in the ball and the hot fluid outside the balls. Multiple situations are simulated, and the heat transfer process of the thermometric sphere is comprehensively studied. The heat convection is certified the most important aspect. Thermal equilibrium can be achieved within 19 minutes, far shorter than the period while the spheres flowing through the core. The simulation results can also applied to derive the thermal fluid temperature backward.
{"title":"Research on Heat Transfer Characteristics of the Thermometric Sphere in HTR-10","authors":"Shiyan Sun, Youjie Zhang, Yanhua Zheng, X. Fang, Xiaoyong Yang","doi":"10.1115/ICONE26-81768","DOIUrl":"https://doi.org/10.1115/ICONE26-81768","url":null,"abstract":"During the operation of the High Temperature Gas-cooled Reactor (HTGR), the hot-spot temperature in the reactor core must be lower than the maximum permissible temperature of the fuel elements and the materials of construction, so that the reactor kept safe. However, no fixed temperature-measuring devices can be set in a pebble-bed core. A special spherical temperature-measuring device is adopted to make sure it brings as small impact to the reactor operation as possible. There are several metal wires with different melting points inside. The graphite thermometric balls will be put onto the top of HTR-10 reactor core, and they record and reflect the highest temperature in different positions in the core when flowing in the pebble bed. Before the reactor core temperature-measuring experiment of HTR-10, we must study the heat transfer characteristics of the graphite thermometric sphere to find out the relationship of the melting conditions and the temperature in the reactor core. A 3-D model of the graphite thermometric ball is established, and CFD method is adopted to research and figure out the thermal equilibrium time and temperature difference between the metal wires in the ball and the hot fluid outside the balls. Multiple situations are simulated, and the heat transfer process of the thermometric sphere is comprehensively studied. The heat convection is certified the most important aspect. Thermal equilibrium can be achieved within 19 minutes, far shorter than the period while the spheres flowing through the core. The simulation results can also applied to derive the thermal fluid temperature backward.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"5 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":"122001344","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}
In the ESNII+ EU FP7 project, a reactor physics benchmark aiming at the whole core calculation with the reflectors and detailed description of the structural elements was specified. This benchmark is based on the 2009 CEA concept of the ALLEGRO core. Fixed nominal technological data at nominal reactor state (geometry, composition) were prescribed which had to be modified in specified calculation branches according to different types of the thermal expansion and control rod positions. The parameters of the point kinetic model to be applied in a system thermal hydraulic code had to be determined this way. Static mechanical models of the expansion processes were specified by the benchmark. The goal of the calculation exercise was to verify the reactor physics codes, namely to get information about the modelling uncertainties and — after — their influence on the calculated results of the safety analyses. The obtained deviations between the participants are characterizing the user effects, the modelling uncertainties and the influence of the nuclear data differences all, without the possibility of their separation because of the complexity of the benchmark problem. A conclusion could be drawn that a step by step procedure starting from simple problems (homogenous material, Wigner-Seitz cell or subassembly in asymptotic approach) is necessary if we wish to identify the reasons of the deviations. For the Doppler effect, a decision was made in this direction already in the ESNII+ project where an infinite regular lattice problem without any leakage had to be solved. This approach of the simplicity is followed by the present benchmarks (one rod and one assembly), but extending the simple benchmarks with burnup calculations and taking into account leakage in asymptotic approximation by neglecting the complicated processes necessary in the reflector regions.
在ESNII+ EU FP7项目中，指定了一个针对整个堆芯计算的反应堆物理基准，包括反射器和结构元件的详细描述。这个基准是基于2009年CEA的ALLEGRO核心概念。规定了反应器标称状态(几何、组成)下的固定标称工艺数据，这些数据必须根据不同类型的热膨胀和控制棒位置在指定的计算分支中进行修改。应用于系统热工规范的点动力学模型的参数必须以这种方式确定。根据基准确定了膨胀过程的静态力学模型。计算练习的目的是验证反应堆物理代码，即获得有关建模不确定性及其对安全分析计算结果的影响的信息。获得的参与者之间的偏差描述了用户效应、建模不确定性和核数据差异的影响，由于基准问题的复杂性，它们不可能分离。可以得出结论，如果我们希望确定偏差的原因，就必须从简单问题(均质材料，Wigner-Seitz单元或渐近方法中的子装配)开始逐步进行程序。对于多普勒效应，ESNII+项目已经在这个方向上做出了决定，该项目需要解决无泄漏的无限规则晶格问题。这种简单的方法适用于目前的基准测试(一杆和一个组件)，但将简单的基准测试扩展为燃耗计算，并通过忽略反射区域中必要的复杂过程来考虑渐近近似中的泄漏。
{"title":"Neutronic Methodological Benchmarks With Simplified Geometries for the Gas Cooled Reactor Group Constant Generating Tools","authors":"E. Temesvári, B. Batki, M. Gren","doi":"10.1115/ICONE26-81427","DOIUrl":"https://doi.org/10.1115/ICONE26-81427","url":null,"abstract":"In the ESNII+ EU FP7 project, a reactor physics benchmark aiming at the whole core calculation with the reflectors and detailed description of the structural elements was specified. This benchmark is based on the 2009 CEA concept of the ALLEGRO core. Fixed nominal technological data at nominal reactor state (geometry, composition) were prescribed which had to be modified in specified calculation branches according to different types of the thermal expansion and control rod positions. The parameters of the point kinetic model to be applied in a system thermal hydraulic code had to be determined this way. Static mechanical models of the expansion processes were specified by the benchmark.\u0000 The goal of the calculation exercise was to verify the reactor physics codes, namely to get information about the modelling uncertainties and — after — their influence on the calculated results of the safety analyses. The obtained deviations between the participants are characterizing the user effects, the modelling uncertainties and the influence of the nuclear data differences all, without the possibility of their separation because of the complexity of the benchmark problem. A conclusion could be drawn that a step by step procedure starting from simple problems (homogenous material, Wigner-Seitz cell or subassembly in asymptotic approach) is necessary if we wish to identify the reasons of the deviations. For the Doppler effect, a decision was made in this direction already in the ESNII+ project where an infinite regular lattice problem without any leakage had to be solved. This approach of the simplicity is followed by the present benchmarks (one rod and one assembly), but extending the simple benchmarks with burnup calculations and taking into account leakage in asymptotic approximation by neglecting the complicated processes necessary in the reflector regions.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"36 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":"123512901","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}
J. Garcia, C. Pétesch, T. Lebarbé, P. Lamagnère, Y. Lejeail
The 2018 edition of the RCC-MRx Code [1] will be issued by the end of the 2018, in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires). This Code set up design and construction rules of research reactor components (coming from the RCC-MX 2008 code developed within the context of the Jules Horowitz Reactor project), and to components operating at high temperature and to the Vacuum Vessel of ITER (coming from the RCC-MR 2007). The extension of the scope of the code to innovative systems such as fusion reactors leads to revisit the background of the code to define the requirements to introduce a new process or a new material. The developed methodology has been applied to the introduction of the Fe–9%Cr–1%W–TaV steel (Eurofer), today in the Probationary Phase Rules of RCC-MRx. It was the first time to introduce a “new” material into the code, new in the sense of non-existing in any current standardization. This process, still in progress, highlights the need to have a minimum of information on the expectation of the code regarding the material data. This paper describes the different steps of the introduction of the Eurofer in the RCC-MRx code as well as the tools developed to facilitate the process.
{"title":"Development of a Standard for Fusion Needs: Example of Introduction of Eurofer in RCC-MRx","authors":"J. Garcia, C. Pétesch, T. Lebarbé, P. Lamagnère, Y. Lejeail","doi":"10.1115/ICONE26-82337","DOIUrl":"https://doi.org/10.1115/ICONE26-82337","url":null,"abstract":"The 2018 edition of the RCC-MRx Code [1] will be issued by the end of the 2018, in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires). This Code set up design and construction rules of research reactor components (coming from the RCC-MX 2008 code developed within the context of the Jules Horowitz Reactor project), and to components operating at high temperature and to the Vacuum Vessel of ITER (coming from the RCC-MR 2007).\u0000 The extension of the scope of the code to innovative systems such as fusion reactors leads to revisit the background of the code to define the requirements to introduce a new process or a new material.\u0000 The developed methodology has been applied to the introduction of the Fe–9%Cr–1%W–TaV steel (Eurofer), today in the Probationary Phase Rules of RCC-MRx. It was the first time to introduce a “new” material into the code, new in the sense of non-existing in any current standardization. This process, still in progress, highlights the need to have a minimum of information on the expectation of the code regarding the material data.\u0000 This paper describes the different steps of the introduction of the Eurofer in the RCC-MRx code as well as the tools developed to facilitate the process.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"35 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":"130535441","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}
Y. Eguchi, T. Sugawara, K. Nishihara, Y. Tazawa, K. Tsujimoto
The Japan Atomic Energy Agency (JAEA) has been conducting the research and development (R&D) on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. To foster the R&D of ADS, the Transmutation Physics Experimental Facility (TEF-P) in the J-PARC project has been planned to build by JAEA [1]. The TEF-P is used minor actinide (MA) fuel which has large decay heat, so during the failure of the core cooling system, the evaluation of the core temperature increase is important. This study aims to evaluate the natural cooling characteristics of TEF-P core and to achieve a design that does not damage the core and the fuels during an accident (the failure of the core cooling system). The experiments using mockup device was performed to validate the heat transfer characteristics in the empty rectangular lattice tube. It was obtained that the actual heat transfer coefficient of empty rectangular lattice tube was about 2.2 times larger than the theoretical free convection model. It was also confirmed that the insertion of any block into the empty rectangular lattice tube could achieve the higher heat transfer coefficient. Using the heat transfer coefficient obtained by experiment results, thermal analysis was performed by the three-dimensional heat transfer analysis. As a result, the calculation results showed that the maximum core temperature will be 294 °C which is less than the design criterion of temperature, 327 °C. It was presented that the design condition which the core temperature will be below the design criterion during the failure of the core cooling system through this study.
{"title":"Evaluation of Heat Removal During the Failure of the Core Cooling for New Critical Assembly","authors":"Y. Eguchi, T. Sugawara, K. Nishihara, Y. Tazawa, K. Tsujimoto","doi":"10.1115/ICONE26-82012","DOIUrl":"https://doi.org/10.1115/ICONE26-82012","url":null,"abstract":"The Japan Atomic Energy Agency (JAEA) has been conducting the research and development (R&D) on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. To foster the R&D of ADS, the Transmutation Physics Experimental Facility (TEF-P) in the J-PARC project has been planned to build by JAEA [1]. The TEF-P is used minor actinide (MA) fuel which has large decay heat, so during the failure of the core cooling system, the evaluation of the core temperature increase is important.\u0000 This study aims to evaluate the natural cooling characteristics of TEF-P core and to achieve a design that does not damage the core and the fuels during an accident (the failure of the core cooling system).\u0000 The experiments using mockup device was performed to validate the heat transfer characteristics in the empty rectangular lattice tube. It was obtained that the actual heat transfer coefficient of empty rectangular lattice tube was about 2.2 times larger than the theoretical free convection model. It was also confirmed that the insertion of any block into the empty rectangular lattice tube could achieve the higher heat transfer coefficient.\u0000 Using the heat transfer coefficient obtained by experiment results, thermal analysis was performed by the three-dimensional heat transfer analysis. As a result, the calculation results showed that the maximum core temperature will be 294 °C which is less than the design criterion of temperature, 327 °C. It was presented that the design condition which the core temperature will be below the design criterion during the failure of the core cooling system through this study.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"116 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":"134456255","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 Backfit Rule, 10 CFR §50.109, requires the NRC staff to produce cost-benefit evaluations to justify any changes it may make in its positions, or any new requirements it might impose on licensees, for the purpose of enhancing plant safety. The Backfit Rule also allows the NRC staff to forgo cost-benefit evaluations when it identifies errors or omissions in licensing submittals, or in its own reviews of licensing submittals. In such cases, there is no safety enhancement to be realized. Instead, the NRC staff seeks to obtain reasonable assurance that the level of safety, required by regulations, and licensees’ commitments, is maintained or, if necessary, restored. This provision in the Backfit Rule, called the “Compliance Exception”, has been proven to be very difficult to apply. In 1998 and in 2001 two licensees had argued, in License Amendment Requests (LARs) that their pressurizer safety valves (PSVs) were qualified for water relief duty. Consequently, their PSVs were safety grade components that could be assumed to be available, in licensing basis accident analyses, to open, relieve water, and then reseat. This capability was thought to be required in order to mitigate certain accidents that caused the pressurizer to become water-solid. The 1998 LAR was withdrawn when the licensee was informed that its PSV test results did not demonstrate a capability to relieve water. However, the 2001 LAR was approved, based upon the licensee’s claim that it had acceptable PSV test results. Later, in 2013, the NRC staff realized that the PSV tests, cited in the 2001 application, did not actually exist. So, after careful consideration, over a two-year period, the NRC staff issued a compliance-based backfit order to the licensee. The licensee appealed the order, and the NRC staff denied the appeal. Then the licensee filed a second appeal, this time directly with the NRC’s Executive Director of Operations (EDO). (Such appeals are allowed by the Backfit Rule.) The EDO granted this appeal. So, the compliance-based backfit order, which was intended to address the missing PSV test results, was ultimately overturned. The PSV test results are still missing; but the licensee now has the NRC’s approval to assume the operation of water-qualified PSVs in its licensing basis accident analyses. This paper follows the writing, issuance, and appeal of this compliance-based backfit order, and describes how difficult it is to apply the Compliance Exception of the Backfit Rule.
{"title":"The Backfit Rule’s Compliance Exception","authors":"Samuel Miranda","doi":"10.1115/ICONE26-81905","DOIUrl":"https://doi.org/10.1115/ICONE26-81905","url":null,"abstract":"The Backfit Rule, 10 CFR §50.109, requires the NRC staff to produce cost-benefit evaluations to justify any changes it may make in its positions, or any new requirements it might impose on licensees, for the purpose of enhancing plant safety. The Backfit Rule also allows the NRC staff to forgo cost-benefit evaluations when it identifies errors or omissions in licensing submittals, or in its own reviews of licensing submittals. In such cases, there is no safety enhancement to be realized. Instead, the NRC staff seeks to obtain reasonable assurance that the level of safety, required by regulations, and licensees’ commitments, is maintained or, if necessary, restored. This provision in the Backfit Rule, called the “Compliance Exception”, has been proven to be very difficult to apply.\u0000 In 1998 and in 2001 two licensees had argued, in License Amendment Requests (LARs) that their pressurizer safety valves (PSVs) were qualified for water relief duty. Consequently, their PSVs were safety grade components that could be assumed to be available, in licensing basis accident analyses, to open, relieve water, and then reseat. This capability was thought to be required in order to mitigate certain accidents that caused the pressurizer to become water-solid. The 1998 LAR was withdrawn when the licensee was informed that its PSV test results did not demonstrate a capability to relieve water. However, the 2001 LAR was approved, based upon the licensee’s claim that it had acceptable PSV test results. Later, in 2013, the NRC staff realized that the PSV tests, cited in the 2001 application, did not actually exist. So, after careful consideration, over a two-year period, the NRC staff issued a compliance-based backfit order to the licensee. The licensee appealed the order, and the NRC staff denied the appeal. Then the licensee filed a second appeal, this time directly with the NRC’s Executive Director of Operations (EDO). (Such appeals are allowed by the Backfit Rule.) The EDO granted this appeal. So, the compliance-based backfit order, which was intended to address the missing PSV test results, was ultimately overturned. The PSV test results are still missing; but the licensee now has the NRC’s approval to assume the operation of water-qualified PSVs in its licensing basis accident analyses.\u0000 This paper follows the writing, issuance, and appeal of this compliance-based backfit order, and describes how difficult it is to apply the Compliance Exception of the Backfit Rule.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"18 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":"116168933","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}
Nuclear power plant is composed of many structures, systems and components. In the design and development of nuclear power plants, in order to improve their safety, more and more designers realized that the standards and methods of design, manufacture and supervision of nuclear facilities and equipment have changed a lot. In the new period, facing the deteriorating environment, green, safety and environmental protection are all the things we have to pursue in our real life. Therefore, it is urgent to develop clean and efficient nuclear power sources and focus on improving the safety, reliability and economy of nuclear power plants. For extreme accidents that are extremely unlikely, the equipment of nuclear power plant should be kept safe and worthy of further exploration under the design extension conditions. In order to ensure nuclear safety, according to the practical experience at home and abroad and the practice of new nuclear power plant, it is proposed to improve the safety classification theory and method of nuclear power plant equipment. At the same time, combined with the in service inspection, equipment identification, quality assurance requirements, maintainability and technical maturity of nuclear power plant equipment, we comprehensively and deeply study the safety classification of nuclear power plant equipment.
{"title":"Discussion on Safety Classification for Equipment of Nuclear Power Plants","authors":"He Dan, Zhang Yue","doi":"10.1115/ICONE26-82105","DOIUrl":"https://doi.org/10.1115/ICONE26-82105","url":null,"abstract":"Nuclear power plant is composed of many structures, systems and components. In the design and development of nuclear power plants, in order to improve their safety, more and more designers realized that the standards and methods of design, manufacture and supervision of nuclear facilities and equipment have changed a lot. In the new period, facing the deteriorating environment, green, safety and environmental protection are all the things we have to pursue in our real life. Therefore, it is urgent to develop clean and efficient nuclear power sources and focus on improving the safety, reliability and economy of nuclear power plants. For extreme accidents that are extremely unlikely, the equipment of nuclear power plant should be kept safe and worthy of further exploration under the design extension conditions. In order to ensure nuclear safety, according to the practical experience at home and abroad and the practice of new nuclear power plant, it is proposed to improve the safety classification theory and method of nuclear power plant equipment. At the same time, combined with the in service inspection, equipment identification, quality assurance requirements, maintainability and technical maturity of nuclear power plant equipment, we comprehensively and deeply study the safety classification of nuclear power plant equipment.","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":"121917581","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. Ang, Nuh Mohamud, Hiromasa Chitose, Naoki Hirokawa, Ryusuke Kimura
The demand for continuous improvement in safety of nuclear power plants has led to an international expectation that early or large releases of fission products as a result of severe accidents be practically eliminated for new reactor designs. The UK Department for Business, Energy and Industrial Strategy has recently published the Seventh UK Report which provided confirmation of UK demonstrating compliance with the obligations of the Convention on Nuclear Safety [Ref-1]. Relating to Article 14 on the Assessment and Verification of Safety, the UK nuclear Regulator (Office for Nuclear Regulation (ONR)) has the expectation of the demonstration of ‘practical elimination’ of potential severe accident states be included in the safety cases for new nuclear power plants [Ref-2]. In order to achieve this, the safety case should show either that it is physically impossible for the accident states to occur or the states can be considered to be extremely unlikely with a high degree of confidence by design provisions [Ref-3] [Ref-4] [Ref-5]. A demonstration framework was developed and applied successfully in the UK ABWR Generic Design Assessment (GDA) Pre-Construction Safety Report (PCSR) which was submitted to the ONR in August 2017 [Ref-6]. A summary of this demonstration is provided in this paper.
{"title":"A Demonstration of Practical Elimination of Early or Large Fission Product Release for the UK ABWR Generic Design Assessment","authors":"M. Ang, Nuh Mohamud, Hiromasa Chitose, Naoki Hirokawa, Ryusuke Kimura","doi":"10.1115/ICONE26-82045","DOIUrl":"https://doi.org/10.1115/ICONE26-82045","url":null,"abstract":"The demand for continuous improvement in safety of nuclear power plants has led to an international expectation that early or large releases of fission products as a result of severe accidents be practically eliminated for new reactor designs. The UK Department for Business, Energy and Industrial Strategy has recently published the Seventh UK Report which provided confirmation of UK demonstrating compliance with the obligations of the Convention on Nuclear Safety [Ref-1]. Relating to Article 14 on the Assessment and Verification of Safety, the UK nuclear Regulator (Office for Nuclear Regulation (ONR)) has the expectation of the demonstration of ‘practical elimination’ of potential severe accident states be included in the safety cases for new nuclear power plants [Ref-2]. In order to achieve this, the safety case should show either that it is physically impossible for the accident states to occur or the states can be considered to be extremely unlikely with a high degree of confidence by design provisions [Ref-3] [Ref-4] [Ref-5]. A demonstration framework was developed and applied successfully in the UK ABWR Generic Design Assessment (GDA) Pre-Construction Safety Report (PCSR) which was submitted to the ONR in August 2017 [Ref-6].\u0000 A summary of this demonstration is provided in this paper.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"44 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":"128971178","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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