G. Cartland-Glover, S. Rolfo, A. Skillen, D. Emerson, C. Moulinec, D. Litskevich, B. Merk
Molten salt reactors are a very promising option for the development of highly innovative solutions for the nuclear energy production of the future. New results of the application of the nodal diffusion code, DYN3D-MG, to model the steady state operation of molten salt fast reactors based on circulating fluoride salt, are presented here. The cross-section set was prepared using the Monte Carlo particle transport code, SERPENT. Full core comparisons between SERPENT and DYN3D-MG demonstrate that the molten salt fast reactor steady state operation can be modelled by DYN3D-MG within the required accuracy. The work is an important initial step for the development of a coupled CFD/neutron physics code system. This code system will finally be applied for the investigation of an innovative material protection scheme to avoid the contact of the liquid salt with the structural material.
{"title":"Modelling the Neutronics of a Molten Salt Fast Reactor Using DYN3D-MG for the Investigation of the Application of Frozen Wall Technology","authors":"G. Cartland-Glover, S. Rolfo, A. Skillen, D. Emerson, C. Moulinec, D. Litskevich, B. Merk","doi":"10.1115/ICONE26-82170","DOIUrl":"https://doi.org/10.1115/ICONE26-82170","url":null,"abstract":"Molten salt reactors are a very promising option for the development of highly innovative solutions for the nuclear energy production of the future. New results of the application of the nodal diffusion code, DYN3D-MG, to model the steady state operation of molten salt fast reactors based on circulating fluoride salt, are presented here. The cross-section set was prepared using the Monte Carlo particle transport code, SERPENT. Full core comparisons between SERPENT and DYN3D-MG demonstrate that the molten salt fast reactor steady state operation can be modelled by DYN3D-MG within the required accuracy. The work is an important initial step for the development of a coupled CFD/neutron physics code system. This code system will finally be applied for the investigation of an innovative material protection scheme to avoid the contact of the liquid salt with the structural material.","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":"115877731","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}
Chinese Fusion Engineering Test Reactor (CFETR) is a new test Tokamak device which is now being designed in China to make the transition from the International Thermonuclear Experimental Reactor (ITER) to the future Fusion Power Plant (FPP). Breeding blanket is the key component of fusion reactor which is mainly responsible for the tritium self-sufficiency and fusion energy conversion. In the past few years, three kinds of blanket conceptual design schemes have been proposed and tested in parallel for CFETR Phase I, in which the helium cooled solid breeder (HCSB) blanket concept is acknowledged as the most promising one. However, nowadays, the design phase of CFETR has gradually changed from I to II aiming for the future DEMO operation condition, the main parameters of which are quite different from the previous one. Therefore, it’s necessary to perform conceptual design and various analyses for the HCSB blanket under the new working condition. In this work, firstly, a new conceptual design scheme of HCSB blanket for Phase II is put forward. Then, the radial build arrangements, of the two typical blanket modules are optimized by using the NTCOC. This work can provide valuable references for further conceptual design and neutronics/thermal-hydraulic coupling analyses of the HCSB blanket for CFETR Phase II.
{"title":"Conceptual Design and Neutronics/Thermal-Hydraulic Coupling Optimization Analyses of Two Typical Helium Cooled Solid Breeder Blanket Modules for CFETR Phase II","authors":"Shijie Cui, Dalin Zhang, W. Tian, G. Su, S. Qiu","doi":"10.1115/ICONE26-81539","DOIUrl":"https://doi.org/10.1115/ICONE26-81539","url":null,"abstract":"Chinese Fusion Engineering Test Reactor (CFETR) is a new test Tokamak device which is now being designed in China to make the transition from the International Thermonuclear Experimental Reactor (ITER) to the future Fusion Power Plant (FPP). Breeding blanket is the key component of fusion reactor which is mainly responsible for the tritium self-sufficiency and fusion energy conversion. In the past few years, three kinds of blanket conceptual design schemes have been proposed and tested in parallel for CFETR Phase I, in which the helium cooled solid breeder (HCSB) blanket concept is acknowledged as the most promising one. However, nowadays, the design phase of CFETR has gradually changed from I to II aiming for the future DEMO operation condition, the main parameters of which are quite different from the previous one. Therefore, it’s necessary to perform conceptual design and various analyses for the HCSB blanket under the new working condition. In this work, firstly, a new conceptual design scheme of HCSB blanket for Phase II is put forward. Then, the radial build arrangements, of the two typical blanket modules are optimized by using the NTCOC. This work can provide valuable references for further conceptual design and neutronics/thermal-hydraulic coupling analyses of the HCSB blanket for CFETR Phase II.","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":"115901925","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}
Shogo Noda, S. Sukarman, A. Yamaji, Tetsuo Takei, T. Fukuda, Arisa Ayukawa
The Super Fast Breeder Reactor (Super FBR) utilizes supercritical light water as coolant, which changes from liquidlike high density state to gas-like low density state continuously in the core without phase change. In the preceding study (Noda et al., 2017), new concept of axially heterogeneous core with multi-axial fuel shuffling was proposed. The core consisted of two layers of mixed oxide (MOX) fuel and two layers of blanket fuel with depleted uranium (DU), which were arranged alternatively in the axial direction. The study showed that, with independent fuel shuffling in the upper part and lower part of the core, breeding performance could be improved by increasing the upper blanket fuel batch number while keeping the fuel batch number of the rest of the core unchanged, because of increased neutron flux in the upper blanket. However, the study did not consider influence of different coolant density histories in the different axial level of the core on the core neutronics. Hence, this study aims to reveal influence of the different coolant density histories through design and analyses of the multi-axial fuel shuffling core with two MOX layers and three blanket layers. The three levels correspond to the coolant density below, around, and above the pseudo-critical temperature. The neutronics calculations are carried out with SRAC 2006 code and JENDL-3.3 nuclear data library. Unit cell burnup calculations based on collision probability method are carried out for 5 different coolant density histories to consider influence of different neutron spectrum on breeding performance of the core. Influence of instantaneous coolant density changes on the core neutronics are considered by coupling core burnup calculations with thermal-hydraulics calculations based on single channel model. Influence of independent fuel shuffling of the upper blanket on the core neutronics (breeding performance and void reactivity characteristics) is investigated, followed by a similar investigation on the lower blanket. The differences between the two schemes are investigated since coolant density histories are greatly different between the upper blanket and the lower blanket.
超级快中子增殖反应堆(Super FBR)利用超临界轻水作为冷却剂，在堆芯内由液态高密度状态连续变化到气态低密度状态，无相变。在之前的研究中(Noda et al.， 2017)，提出了轴向非均质堆芯多轴向燃料混叠的新概念。堆芯由两层混合氧化物(MOX)燃料和两层贫铀(DU)包层燃料组成，沿轴向交替排列。研究表明，在堆芯上下部分燃料独立混叠的情况下，由于堆芯上包层的中子通量增加，在增加堆芯上包层燃料批号的同时保持堆芯其余部分燃料批号不变，可以提高堆芯的增殖性能。然而，该研究没有考虑不同冷却剂密度历史在堆芯不同轴向水平对堆芯中子的影响。因此，本研究旨在通过设计和分析两层MOX和三层包层的多轴燃料混流堆芯，揭示不同冷却剂密度历史对堆芯的影响。这三个级别对应于冷却剂密度低于、接近和高于伪临界温度。利用SRAC 2006程序和JENDL-3.3核数据库进行中子计算。为考虑不同中子谱对堆芯增殖性能的影响，对5种不同的冷却剂密度历史进行了基于碰撞概率法的单体电池燃耗计算。通过将堆芯燃耗计算与基于单通道模型的热工水力学计算耦合，考虑了冷却剂密度瞬时变化对堆芯中子电子学的影响。研究了上包层的独立燃料变换对堆芯中子(增殖性能和空洞反应特性)的影响，随后对下包层进行了类似的研究。由于上包层和下包层之间的冷却剂密度历史差异很大，因此研究了两种方案之间的差异。
{"title":"Core Design Study of Super FBR With Multi-Axial Fuel Shuffling and Different Coolant Density","authors":"Shogo Noda, S. Sukarman, A. Yamaji, Tetsuo Takei, T. Fukuda, Arisa Ayukawa","doi":"10.1115/ICONE26-81501","DOIUrl":"https://doi.org/10.1115/ICONE26-81501","url":null,"abstract":"The Super Fast Breeder Reactor (Super FBR) utilizes supercritical light water as coolant, which changes from liquidlike high density state to gas-like low density state continuously in the core without phase change. In the preceding study (Noda et al., 2017), new concept of axially heterogeneous core with multi-axial fuel shuffling was proposed. The core consisted of two layers of mixed oxide (MOX) fuel and two layers of blanket fuel with depleted uranium (DU), which were arranged alternatively in the axial direction. The study showed that, with independent fuel shuffling in the upper part and lower part of the core, breeding performance could be improved by increasing the upper blanket fuel batch number while keeping the fuel batch number of the rest of the core unchanged, because of increased neutron flux in the upper blanket. However, the study did not consider influence of different coolant density histories in the different axial level of the core on the core neutronics. Hence, this study aims to reveal influence of the different coolant density histories through design and analyses of the multi-axial fuel shuffling core with two MOX layers and three blanket layers. The three levels correspond to the coolant density below, around, and above the pseudo-critical temperature. The neutronics calculations are carried out with SRAC 2006 code and JENDL-3.3 nuclear data library. Unit cell burnup calculations based on collision probability method are carried out for 5 different coolant density histories to consider influence of different neutron spectrum on breeding performance of the core. Influence of instantaneous coolant density changes on the core neutronics are considered by coupling core burnup calculations with thermal-hydraulics calculations based on single channel model. Influence of independent fuel shuffling of the upper blanket on the core neutronics (breeding performance and void reactivity characteristics) is investigated, followed by a similar investigation on the lower blanket. The differences between the two schemes are investigated since coolant density histories are greatly different between the upper blanket and the lower blanket.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"131 3 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":"130778181","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 goal of basic research is to gather information and build upon the knowledge base. This type of study identifies a problem and suggests solutions that applied research can consider and employ using empirical methodologies. Basic theory is a valid and accepted research tool, it is often mathematical in nature, and it is formulated to solve a particular problem. The problem is to find the mathematical link between the natural energy of stars and the energy created through fission. An equation is shown here to have a place in the field of nuclear engineering because it provides insight into the question of what makes the binding and decaying of atomic particles mathematically possible. This equation shows that when the two particles of unequal mass move close enough to be caught up in each other’s flow fields, they are entangled in the fields and remain bound together. When two particles are of equal mass however their encounter does not produce a binding mechanism because their flow fields are the same strength and cancel each other out. The fission of uranium-236 into the elements krypton-92 and barium-141 is discussed according to the equation. Then the fusion of deuterium and tritium is analyzed as an example of flow fields at work. Finally the first three steps of the proton-proton chain reaction are compared to the equation.
{"title":"A Mathematical Link Between the Natural Energy of Stars and Fission","authors":"B. Bayles","doi":"10.1115/ICONE26-81093","DOIUrl":"https://doi.org/10.1115/ICONE26-81093","url":null,"abstract":"The goal of basic research is to gather information and build upon the knowledge base. This type of study identifies a problem and suggests solutions that applied research can consider and employ using empirical methodologies. Basic theory is a valid and accepted research tool, it is often mathematical in nature, and it is formulated to solve a particular problem. The problem is to find the mathematical link between the natural energy of stars and the energy created through fission. An equation is shown here to have a place in the field of nuclear engineering because it provides insight into the question of what makes the binding and decaying of atomic particles mathematically possible. This equation shows that when the two particles of unequal mass move close enough to be caught up in each other’s flow fields, they are entangled in the fields and remain bound together. When two particles are of equal mass however their encounter does not produce a binding mechanism because their flow fields are the same strength and cancel each other out. The fission of uranium-236 into the elements krypton-92 and barium-141 is discussed according to the equation. Then the fusion of deuterium and tritium is analyzed as an example of flow fields at work. Finally the first three steps of the proton-proton chain reaction are compared to the equation.","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":"130237880","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}
Qiudong Wang, B. Xia, Jiong Guo, D. She, Lei Shi, Zuoyi Zhang
In this work, a two-zone reactor core, which contains an inner driving zone and an outer ThO2 breeding zone, is designed under the framework of the HTR-PM. The main aim of this work is to investigate the feasibility of thorium utilization in the mature design of the HTR-PM with the inherent safety features. The neutronics and thermal-hydraulics characteristics are investigated to optimize the design parameters by using VSOP. The aim of optimization is to maximize the conversion of thorium to 233U in the breeding zone. The preliminary results indicate that the volume ratio of the breeding zone to the driving zone has significant influence on the power peaking factor and the maximum fuel temperature in normal operation and accidental conditions. On the other hand, the increase of reactor power will lead to increase of maximum fuel temperature after DLOFC accident. More heavy metal loading in the breeding zone will raise 233U yield, while the influence of fuel particle radius on the conversion ratio is negligible. An optimized 200 MWt two-zone reactor design is obtained with volume ratio of the driving zone to the breeding zone of 4:1, and 7 g and 30 g heavy metal per fuel sphere in the driving zone and the breeding zone, respectively.
{"title":"Preliminary Neutronics and Thermal-Hydraulics Study on Thorium-Based HTR-PM With Outer Breeding Zone","authors":"Qiudong Wang, B. Xia, Jiong Guo, D. She, Lei Shi, Zuoyi Zhang","doi":"10.1115/ICONE26-81975","DOIUrl":"https://doi.org/10.1115/ICONE26-81975","url":null,"abstract":"In this work, a two-zone reactor core, which contains an inner driving zone and an outer ThO2 breeding zone, is designed under the framework of the HTR-PM. The main aim of this work is to investigate the feasibility of thorium utilization in the mature design of the HTR-PM with the inherent safety features. The neutronics and thermal-hydraulics characteristics are investigated to optimize the design parameters by using VSOP. The aim of optimization is to maximize the conversion of thorium to 233U in the breeding zone. The preliminary results indicate that the volume ratio of the breeding zone to the driving zone has significant influence on the power peaking factor and the maximum fuel temperature in normal operation and accidental conditions. On the other hand, the increase of reactor power will lead to increase of maximum fuel temperature after DLOFC accident. More heavy metal loading in the breeding zone will raise 233U yield, while the influence of fuel particle radius on the conversion ratio is negligible. An optimized 200 MWt two-zone reactor design is obtained with volume ratio of the driving zone to the breeding zone of 4:1, and 7 g and 30 g heavy metal per fuel sphere in the driving zone and the breeding zone, respectively.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"10 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":"126316230","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 NRC staff, and licensees will sometimes cite a Known and Established Standard, especially when there is evidence that the staff’s position, with respect to a Known and Established Standard, may have changed. Licensees will often flag such changes as backfits, and demand that they meet the requirements of the Backfit Rule in the Code of Federal Regulations (10 CFR §50.109). However, the term, Known and Established Standard, is not clearly defined. If the term is not defined by the NRC staff, then it’s possible that it will be defined by others, most notably by licensees with technical issues or license amendment requests (LARs) that are undergoing the staff’s review. What follows is a discussion of the origin and application of the NRC’s Known and Established Standard, and how it might be defined. The Known and Established Standard is first deconstructed into its two components, Known and Established, and then each is defined separately. When this is done, it becomes apparent that some Known and Established Standards can be more important than others. The two definitions can be used, together, to distinguish between unequal, even incompatible Known and Established Standards. This could help the NRC staff identify the most relevant Known and Established Standard to apply in its review or evaluation of a particular technical issue or license amendment request (LAR). Simply choosing the most recent of Known and Established Standards might not be adequate.
{"title":"What Is a “Known and Established” Standard?","authors":"Samuel Miranda","doi":"10.1115/ICONE26-81901","DOIUrl":"https://doi.org/10.1115/ICONE26-81901","url":null,"abstract":"The NRC staff, and licensees will sometimes cite a Known and Established Standard, especially when there is evidence that the staff’s position, with respect to a Known and Established Standard, may have changed. Licensees will often flag such changes as backfits, and demand that they meet the requirements of the Backfit Rule in the Code of Federal Regulations (10 CFR §50.109). However, the term, Known and Established Standard, is not clearly defined. If the term is not defined by the NRC staff, then it’s possible that it will be defined by others, most notably by licensees with technical issues or license amendment requests (LARs) that are undergoing the staff’s review.\u0000 What follows is a discussion of the origin and application of the NRC’s Known and Established Standard, and how it might be defined. The Known and Established Standard is first deconstructed into its two components, Known and Established, and then each is defined separately. When this is done, it becomes apparent that some Known and Established Standards can be more important than others. The two definitions can be used, together, to distinguish between unequal, even incompatible Known and Established Standards. This could help the NRC staff identify the most relevant Known and Established Standard to apply in its review or evaluation of a particular technical issue or license amendment request (LAR). Simply choosing the most recent of Known and Established Standards might not be adequate.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"76 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":"126063812","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}
Liquid lead-bismuth eutectic (LBE) has been studied as a new type of coolant for an accelerator driven sub-critical system (ADS). And the gas-lift pump has been used to enhance the circulation capacity of coolant in ADS instead of mechanical pumps due to its simpler mechanical structure and higher security. The purpose of this experiment is to study the circulation capacity enhancement by gas-lift technique and the flow and heat transfer characteristic of liquid LBE in an annular channel. The experimental results show that: gas-injection can significantly increase liquid LBE mass flow rate, but the growth of liquid LBE mass flow rate will be reduced when gas flow rate reaches a value; The friction coefficient of liquid LBE in an annular channel decreases with the increase of Re and is larger than that calculated by Blasius formula at the same Re. For the convection heat transfer of liquid LBE in an annular channel, the heat conduction term is dominant, and Nusselt number increases with the increase of Peclet number. The experimental correlations of friction coefficient and convection heat transfer of liquid LBE in annular channel were fitted based on experimental data and compared with those from literature.
{"title":"Experimental Research on Heat Transfer Characteristic of Liquid Lead-Bismuth Eutectic Flowing in Annular Channel","authors":"F. Zhu, Junmei Wu, Leitai Shi, G. Su","doi":"10.1115/ICONE26-82042","DOIUrl":"https://doi.org/10.1115/ICONE26-82042","url":null,"abstract":"Liquid lead-bismuth eutectic (LBE) has been studied as a new type of coolant for an accelerator driven sub-critical system (ADS). And the gas-lift pump has been used to enhance the circulation capacity of coolant in ADS instead of mechanical pumps due to its simpler mechanical structure and higher security. The purpose of this experiment is to study the circulation capacity enhancement by gas-lift technique and the flow and heat transfer characteristic of liquid LBE in an annular channel. The experimental results show that: gas-injection can significantly increase liquid LBE mass flow rate, but the growth of liquid LBE mass flow rate will be reduced when gas flow rate reaches a value; The friction coefficient of liquid LBE in an annular channel decreases with the increase of Re and is larger than that calculated by Blasius formula at the same Re. For the convection heat transfer of liquid LBE in an annular channel, the heat conduction term is dominant, and Nusselt number increases with the increase of Peclet number. The experimental correlations of friction coefficient and convection heat transfer of liquid LBE in annular channel were fitted based on experimental data and compared with those from literature.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"31 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":"128092878","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 EU-APR standard design is a two-loop pressurised water reactor of the range of 1400MWe, developed by Korean Hydro & Nuclear Power (KHNP) from the Korean APR1400 reference plant. It is the objective of the EU-APR to comply with the current main European regulatory and utility requirements for safe and competitive LWR NPPs. A full assessment of the EU-APR against the European Utility Requirements (EUR) can therefore be regarded as an important step in its introduction to the European market. KHNP’s request to assess its EU-APR standard design was granted by the EUR Organisation and in autumn of 2015, a two-year assessment process started, during which the EU-APR was assessed against the twenty chapters of the EUR Document Volume 2, encompassing over four thousand individual requirements having in view principally NPP safety and performance. The EU-APR standard design was assessed against each single requirement by experts of the appropriate discipline from within the EUR member companies. The resultant assessment reports were reviewed by the Coordination Group (working level team) and then validated by the EUR Administration Group and EUR Steering Committee. The syntheses of the requirement-by-requirement assessments per chapter were collected in the principal project output: the EU-APR dedicated subset of EUR Document Volume 3. This Volume 3 subset (Subset I) also contains a technical description of the plant, issued by KHNP and reviewed by the EUR organisation. Furthermore, it explains the assessment process and highlights the main assessment results. Volume 3 Subset I was approved by the EUR Steering Committee in September 2017. The EU-APR standard design was found to be fully compliant for a majority (almost 90%) of all requirements applicable in this assessment. Furthermore, the EU-APR was found to be compliant with the objectives of an additional 6% of the investigated requirements. The EU-APR standard design can be considered to be in the late basic design phase. Nonetheless, the portion of requirements which were not assessable is very low (below 3%), explained by the abundant availability of reference plant (detailed design) documentation, which often helped demonstrate by means of extrapolatable examples KHNP’s adherence to design principles established for the EU-APR. In total, less than 1% of assessed requirements were found to be non-compliant. The main non-compliances relate to choices of design principles and analysis methods that are explicitly different from those stipulated by EUR, and to the relatively low level of development of reactor core and fuel studies to demonstrate operational nuclear plant manoeuvrability and fuel cycle optimisation in view of safety margins and economics.
{"title":"The EUR Assessment Process and Highlights of the Compliance Analysis for the EU-APR Standard Design","authors":"C. Declercq, A. Ballard, G. Ferraro, Anicet Touré","doi":"10.1115/ICONE26-81889","DOIUrl":"https://doi.org/10.1115/ICONE26-81889","url":null,"abstract":"The EU-APR standard design is a two-loop pressurised water reactor of the range of 1400MWe, developed by Korean Hydro & Nuclear Power (KHNP) from the Korean APR1400 reference plant. It is the objective of the EU-APR to comply with the current main European regulatory and utility requirements for safe and competitive LWR NPPs. A full assessment of the EU-APR against the European Utility Requirements (EUR) can therefore be regarded as an important step in its introduction to the European market.\u0000 KHNP’s request to assess its EU-APR standard design was granted by the EUR Organisation and in autumn of 2015, a two-year assessment process started, during which the EU-APR was assessed against the twenty chapters of the EUR Document Volume 2, encompassing over four thousand individual requirements having in view principally NPP safety and performance.\u0000 The EU-APR standard design was assessed against each single requirement by experts of the appropriate discipline from within the EUR member companies. The resultant assessment reports were reviewed by the Coordination Group (working level team) and then validated by the EUR Administration Group and EUR Steering Committee. The syntheses of the requirement-by-requirement assessments per chapter were collected in the principal project output: the EU-APR dedicated subset of EUR Document Volume 3. This Volume 3 subset (Subset I) also contains a technical description of the plant, issued by KHNP and reviewed by the EUR organisation. Furthermore, it explains the assessment process and highlights the main assessment results. Volume 3 Subset I was approved by the EUR Steering Committee in September 2017.\u0000 The EU-APR standard design was found to be fully compliant for a majority (almost 90%) of all requirements applicable in this assessment. Furthermore, the EU-APR was found to be compliant with the objectives of an additional 6% of the investigated requirements. The EU-APR standard design can be considered to be in the late basic design phase. Nonetheless, the portion of requirements which were not assessable is very low (below 3%), explained by the abundant availability of reference plant (detailed design) documentation, which often helped demonstrate by means of extrapolatable examples KHNP’s adherence to design principles established for the EU-APR. In total, less than 1% of assessed requirements were found to be non-compliant. The main non-compliances relate to choices of design principles and analysis methods that are explicitly different from those stipulated by EUR, and to the relatively low level of development of reactor core and fuel studies to demonstrate operational nuclear plant manoeuvrability and fuel cycle optimisation in view of safety margins and economics.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"48 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":"126994687","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}
So far, the licenses of more than 85% of US operating plants have been renewed, authorizing them to continue operations for an additional 20 years past the end of their original, 40-year operating licenses. 10 CFR §54, which governs the NRC’s renewal of plant operating licenses, defines the Current Licensing Basis (CLB), and its role in license renewal applications. The CLB includes, inter alia, compliance with all the regulations in 10 CFR §50, and its appendices, including the design-basis information presented in final safety analysis reports (FSARs). Appendix A of 10 CFR §50 specifies the General Design Criteria (GDCs), and defines the conditions under which they must be satisfied. For example, Appendix A defines anticipated operational occurrences (AOOs) as incidents that may occur during the lifetime of a particular plant. Two standards of the American Nuclear Society (ANS) [1] [2] redefined AOOs as incidents that may occur during the calendar year for a particular plant, and added a class of events that may occur during the lifetime of a particular plant. The ANS standards defined a categorization scheme that puts all the various types of incidents into four categories (i.e., Conditions: I, II, III, and IV), and specifies the GDCs and other criteria that must be met for each of the categories. AOOs, or Condition II events, for example, must not result in any fuel damage. Licensees have committed to abide by the ANS categorization system, and to comply with all of the categories’ acceptance criteria. These commitments are in their Final Safety Analysis Reports (FSARs), which are part of their CLBs. Conditions I, and II define relatively frequently occurring incidents, and require that their consequences to be benign. Condition III events, however, are limited to only a very few incidents during the lifetime of a plant. Their consequences are not benign. (Condition IV is the most severe category. Condition IV events are not expect to occur at all.) Since the frequency of occurrence of Condition III events is defined in terms of plant lifetime, it follows that lengthening the plant lifetime (e.g., from 40 to 60 years) will lead to the occurrence of more Condition III events. These events can result in fuel damage, or worse. The CLB of a renewed license plant, therefore, will have to account for more Condition III events. This paper focuses upon how Condition III events can affect the CLB during an extended plant lifetime. It also discusses the concept of extending plant operating licenses by 20, or 40 years, and its potential impact upon the public health and safety.
{"title":"The Role of the NRC in License Renewals","authors":"Samuel Miranda","doi":"10.1115/ICONE26-81904","DOIUrl":"https://doi.org/10.1115/ICONE26-81904","url":null,"abstract":"So far, the licenses of more than 85% of US operating plants have been renewed, authorizing them to continue operations for an additional 20 years past the end of their original, 40-year operating licenses.\u0000 10 CFR §54, which governs the NRC’s renewal of plant operating licenses, defines the Current Licensing Basis (CLB), and its role in license renewal applications. The CLB includes, inter alia, compliance with all the regulations in 10 CFR §50, and its appendices, including the design-basis information presented in final safety analysis reports (FSARs). Appendix A of 10 CFR §50 specifies the General Design Criteria (GDCs), and defines the conditions under which they must be satisfied. For example, Appendix A defines anticipated operational occurrences (AOOs) as incidents that may occur during the lifetime of a particular plant. Two standards of the American Nuclear Society (ANS) [1] [2] redefined AOOs as incidents that may occur during the calendar year for a particular plant, and added a class of events that may occur during the lifetime of a particular plant.\u0000 The ANS standards defined a categorization scheme that puts all the various types of incidents into four categories (i.e., Conditions: I, II, III, and IV), and specifies the GDCs and other criteria that must be met for each of the categories. AOOs, or Condition II events, for example, must not result in any fuel damage. Licensees have committed to abide by the ANS categorization system, and to comply with all of the categories’ acceptance criteria. These commitments are in their Final Safety Analysis Reports (FSARs), which are part of their CLBs.\u0000 Conditions I, and II define relatively frequently occurring incidents, and require that their consequences to be benign. Condition III events, however, are limited to only a very few incidents during the lifetime of a plant. Their consequences are not benign. (Condition IV is the most severe category. Condition IV events are not expect to occur at all.)\u0000 Since the frequency of occurrence of Condition III events is defined in terms of plant lifetime, it follows that lengthening the plant lifetime (e.g., from 40 to 60 years) will lead to the occurrence of more Condition III events. These events can result in fuel damage, or worse. The CLB of a renewed license plant, therefore, will have to account for more Condition III events.\u0000 This paper focuses upon how Condition III events can affect the CLB during an extended plant lifetime. It also discusses the concept of extending plant operating licenses by 20, or 40 years, and its potential impact upon the public health and safety.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"41 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":"127326820","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 Fluoride-salt-cooled High temperature Reactors (FHRs) are an advanced concept using a novel combination of high-temperature coated-particle fuel, low-pressure fluoride-salt coolant and air-Brayton power conversion system. Prismatic fuel or pebble fuel are adopted for the conceptual core designs of FHRs like TMSR-SF, MK1 PB-FHR and SM-AHTR. The high-Prandtl-number FLiBe is mainly adopted as the primary coolant, which specifies in high melting and boiling point and high volumetric capacity. The experimental results obtained from the air, water or inert gas prove reliable for the Prandtl number vary from 0.7 to 7. Little experimental research has been conducted to prove applicability of the above results to the high-Prandtl fluid, fluoride salts in the packed pebble bed. In this paper, a pebble bed experimental facility has been designed and constructed for the FHRs to explore the thermal-hydraulic characteristics of fluoride salts in the reactor pebble bed core. Dowtherm A is adopted as a simulant fluid for the fluoride salts. The cylindrical test section is packed with steel pebbles. The electromagnetic induction heating system is used to provide internal heat source for the pebble beds. The forced flow and convective heat transfer of high-Prandtl-number fluid in the pebble bed with internal heat generation are investigated in the experiment. The fluid inlet temperature and mass flow rate are studied on the thermal-hydraulic characteristics.
{"title":"Experimental Studies on the Thermal-Hydraulics of Dowtherm A Through the Pebble Bed With Internal Heat Generation","authors":"Limin Liu, Dalin Zhang, Linfeng Li, Yichen Yang, Chenglong Wang, S. Qiu","doi":"10.1115/ICONE26-81917","DOIUrl":"https://doi.org/10.1115/ICONE26-81917","url":null,"abstract":"The Fluoride-salt-cooled High temperature Reactors (FHRs) are an advanced concept using a novel combination of high-temperature coated-particle fuel, low-pressure fluoride-salt coolant and air-Brayton power conversion system. Prismatic fuel or pebble fuel are adopted for the conceptual core designs of FHRs like TMSR-SF, MK1 PB-FHR and SM-AHTR. The high-Prandtl-number FLiBe is mainly adopted as the primary coolant, which specifies in high melting and boiling point and high volumetric capacity. The experimental results obtained from the air, water or inert gas prove reliable for the Prandtl number vary from 0.7 to 7. Little experimental research has been conducted to prove applicability of the above results to the high-Prandtl fluid, fluoride salts in the packed pebble bed. In this paper, a pebble bed experimental facility has been designed and constructed for the FHRs to explore the thermal-hydraulic characteristics of fluoride salts in the reactor pebble bed core. Dowtherm A is adopted as a simulant fluid for the fluoride salts. The cylindrical test section is packed with steel pebbles. The electromagnetic induction heating system is used to provide internal heat source for the pebble beds. The forced flow and convective heat transfer of high-Prandtl-number fluid in the pebble bed with internal heat generation are investigated in the experiment. The fluid inlet temperature and mass flow rate are studied on the thermal-hydraulic characteristics.","PeriodicalId":354697,"journal":{"name":"Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues","volume":"235 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":"133319754","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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