Ye-Rin Choi, Min Kyu Kim, Jae-Hee Kim, Tae-Young Ryu, Jun-seog Yang, Moon-Ki Kim, Jaeboong Choi
Demands for the safety of the Nuclear Power Plant (NPP) are increasing because of some accidents during decades. For the safety of the NPP, maintaining integrity of the reactor is one of the most important part. For the integrity evaluation of the Reactor Pressure Vessel (RPV), evaluation methods such as Upper Shelf Energy (USE), Pressurized Thermal Shock (PTS), and Pressure-Temperature (P-T) limit curve, etc. have been suggested by the ASME Code with consideration of neutron irradiation embrittlement. Among them, The P-T limit curve suggests limitations for the temperature and pressure during the operation of the RPV. The ASME Code Section XI Appendix G (Sec. XI App. G) suggests a method to generate P-T limit curves of the RPV [1]. There is restriction on the operation procedure; the cooling rate of the reactor is limited to 100 °F/hr or less and the available temperature range for the equations at the ASME Code is also limited to 100 °F/hr. However it is needed to cool down the reactor very fast at the severe accident condition to control the reactor to the stable condition and this sudden temperature drop can cause a thermal shock in the reactor. Therefore it is important to compensate the risk by accurately prepared P-T limit curve with high cooling rate for severe accidents in the NPP. In this study, researchers try to expand the limitation of the cooling rate for the P-T limit curve from 100 °F/hr to 200 °F/hr. Finite Element Analysis (FEA) for integrity evaluation and comparison of results using ASME Code equations were carried out.
{"title":"A Study on the Pressure-Temperature Limit Curve Under High Cooling Rate for the Reactor Using Finite Element Method","authors":"Ye-Rin Choi, Min Kyu Kim, Jae-Hee Kim, Tae-Young Ryu, Jun-seog Yang, Moon-Ki Kim, Jaeboong Choi","doi":"10.1115/PVP2018-84850","DOIUrl":"https://doi.org/10.1115/PVP2018-84850","url":null,"abstract":"Demands for the safety of the Nuclear Power Plant (NPP) are increasing because of some accidents during decades. For the safety of the NPP, maintaining integrity of the reactor is one of the most important part. For the integrity evaluation of the Reactor Pressure Vessel (RPV), evaluation methods such as Upper Shelf Energy (USE), Pressurized Thermal Shock (PTS), and Pressure-Temperature (P-T) limit curve, etc. have been suggested by the ASME Code with consideration of neutron irradiation embrittlement. Among them, The P-T limit curve suggests limitations for the temperature and pressure during the operation of the RPV. The ASME Code Section XI Appendix G (Sec. XI App. G) suggests a method to generate P-T limit curves of the RPV [1]. There is restriction on the operation procedure; the cooling rate of the reactor is limited to 100 °F/hr or less and the available temperature range for the equations at the ASME Code is also limited to 100 °F/hr. However it is needed to cool down the reactor very fast at the severe accident condition to control the reactor to the stable condition and this sudden temperature drop can cause a thermal shock in the reactor. Therefore it is important to compensate the risk by accurately prepared P-T limit curve with high cooling rate for severe accidents in the NPP. In this study, researchers try to expand the limitation of the cooling rate for the P-T limit curve from 100 °F/hr to 200 °F/hr. Finite Element Analysis (FEA) for integrity evaluation and comparison of results using ASME Code equations were carried out.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"13 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90429925","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 U.S. Nuclear Regulatory Commission staff has analyzed results from the weld residual stress round robin study, conducted in 2014. An uncertainty quantification scheme was applied to the dataset in order to compare and contrast results from independent analysts. The uncertainty quantification scheme provides a rigorous framework within which to make judgement calls about appropriate modeling guidelines and potential validation schemes. This paper will explore various options for guidelines and validation approaches, as informed by a statistical analysis of the dataset.
{"title":"Exploring Finite Element Validation for Weld Residual Stress Prediction","authors":"M. Benson, P. Raynaud, J. Wallace","doi":"10.1115/PVP2018-84931","DOIUrl":"https://doi.org/10.1115/PVP2018-84931","url":null,"abstract":"The U.S. Nuclear Regulatory Commission staff has analyzed results from the weld residual stress round robin study, conducted in 2014. An uncertainty quantification scheme was applied to the dataset in order to compare and contrast results from independent analysts. The uncertainty quantification scheme provides a rigorous framework within which to make judgement calls about appropriate modeling guidelines and potential validation schemes. This paper will explore various options for guidelines and validation approaches, as informed by a statistical analysis of the dataset.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"205 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74449716","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 recent years, hydrogen, one of renewable energy, has attracted attention. For widespread commercialization of the hydrogen-energy systems, a useful and reliable evaluation method should be developed for capturing the degradation of strength and fatigue properties of metals in presence of hydrogen. This paper implemented transient hydrogen diffusion-elastoplastic coupling analysis program into a commercial software of Finite Element Analysis (Abaqus) to predict the fatigue crack growth (FCG) acceleration of a low carbon steel (JIS-SM490B) in high-pressure hydrogen gas. For this simulation, hydrogen-diffusion properties (concentration and diffusivity) depending on plastic strain were experimentally obtained. Our thorough numerical results proposed a practical technique to predict an onset of hydrogen-enhanced FCG acceleration measured in experiments, via the numerically obtained gradient of hydrogen concentration at the crack tip. In addition, a practical technique to predict the hydrogen-enhanced FCG acceleration ratio was also discussed based on the gradient of hydrogen concentration.
{"title":"Transient Hydrogen Diffusion/Elastoplastic Coupling Analysis for Predicting Fatigue Crack Growth Acceleration of Low-Carbon Steel in Gaseous Hydrogen","authors":"Kai Kawahara, M. Fujikawa, J. Yamabe","doi":"10.1115/PVP2018-84390","DOIUrl":"https://doi.org/10.1115/PVP2018-84390","url":null,"abstract":"In recent years, hydrogen, one of renewable energy, has attracted attention. For widespread commercialization of the hydrogen-energy systems, a useful and reliable evaluation method should be developed for capturing the degradation of strength and fatigue properties of metals in presence of hydrogen. This paper implemented transient hydrogen diffusion-elastoplastic coupling analysis program into a commercial software of Finite Element Analysis (Abaqus) to predict the fatigue crack growth (FCG) acceleration of a low carbon steel (JIS-SM490B) in high-pressure hydrogen gas. For this simulation, hydrogen-diffusion properties (concentration and diffusivity) depending on plastic strain were experimentally obtained. Our thorough numerical results proposed a practical technique to predict an onset of hydrogen-enhanced FCG acceleration measured in experiments, via the numerically obtained gradient of hydrogen concentration at the crack tip. In addition, a practical technique to predict the hydrogen-enhanced FCG acceleration ratio was also discussed based on the gradient of hydrogen concentration.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"64 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76865319","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 super pipe nozzles in nuclear power plants are usually designed to be in compliance with the requirements of Class 2 piping of Section III of the ASME Boiler and Pressure Vessel Code. The stress indices B2 and stress intensification factor i are required for the stress evaluation. In the past two decades, the hot extrusion forming technology has been widely used to manufacture those nozzles, instead of traditional insert weldolets. However, previous extruded nozzle stress analyses have shown B2 that the calculated stresses may exceed the limits in some working conditions. The objective of present study is to determine the stress indices and stress intensification factor for an extruded nozzle of the supper pipe by the finite element method and to evaluate the conservatism of those factors from the ASME Code formulae. In this paper, a three-dimensional finite element model of an extruded nozzle is developed. Four load cases are considered, which are corresponding to an in-plane bending moment and an out-plane bending moment applied at the run pipe side and at the branch pipe side, respectively. The magnitude of bending moment is assumed to be 1000Nm. The stress indices B2r, B2b, C2r, C2b, K2r and K2b, where the subscript r and b refer to the run pipe and B2r the branch pipe, are calculated based on the finite element analysis results. The stress intensification factor ir and ib are determined by the empirical formula: ir = C2r*K2r/2 and ib = C2b*K2b/2. Further, the developed factors are compared with those calculated from the ASME code formulae. It is found that the stress indices B2r and B2b obtained from the linear elastic finite element analysis are conservative. Currently, the values of B2r and B2b gained from the ASME code formulae are more appropriate for the stress evolution. The stress intensification factors ir and ib obtained from the analytical determination are lower than those calculated from the ASME code formula. For the extrude nozzle studied, the factor ir decreases 30% and the factor ib decreases about 3.3%.
{"title":"Analytical Determination of Stress Indices and Stress Intensification Factor for an Extruded Nozzle of Super Pipe","authors":"Lv Feng, Zhou Gengyu, Qian Haiyang","doi":"10.1115/PVP2018-85144","DOIUrl":"https://doi.org/10.1115/PVP2018-85144","url":null,"abstract":"The super pipe nozzles in nuclear power plants are usually designed to be in compliance with the requirements of Class 2 piping of Section III of the ASME Boiler and Pressure Vessel Code. The stress indices B2 and stress intensification factor i are required for the stress evaluation. In the past two decades, the hot extrusion forming technology has been widely used to manufacture those nozzles, instead of traditional insert weldolets. However, previous extruded nozzle stress analyses have shown B2 that the calculated stresses may exceed the limits in some working conditions. The objective of present study is to determine the stress indices and stress intensification factor for an extruded nozzle of the supper pipe by the finite element method and to evaluate the conservatism of those factors from the ASME Code formulae. In this paper, a three-dimensional finite element model of an extruded nozzle is developed. Four load cases are considered, which are corresponding to an in-plane bending moment and an out-plane bending moment applied at the run pipe side and at the branch pipe side, respectively. The magnitude of bending moment is assumed to be 1000Nm. The stress indices B2r, B2b, C2r, C2b, K2r and K2b, where the subscript r and b refer to the run pipe and B2r the branch pipe, are calculated based on the finite element analysis results. The stress intensification factor ir and ib are determined by the empirical formula: ir = C2r*K2r/2 and ib = C2b*K2b/2. Further, the developed factors are compared with those calculated from the ASME code formulae. It is found that the stress indices B2r and B2b obtained from the linear elastic finite element analysis are conservative. Currently, the values of B2r and B2b gained from the ASME code formulae are more appropriate for the stress evolution. The stress intensification factors ir and ib obtained from the analytical determination are lower than those calculated from the ASME code formula. For the extrude nozzle studied, the factor ir decreases 30% and the factor ib decreases about 3.3%.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"4 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91314126","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}
W. Gong, Xiaochun Zhang, Mingqiang Xie, Yuan Fu, Xiao Wang
Safety assessment and economy design of piping system at high temperature is highly important in power and nuclear engineering. ASME-NH code considers cyclic failure modes at elevated temperature and provides the rules and damage limits for creep-fatigue interaction. Based on the calculation methods for creep strain increment of ASME-NH code, this paper investigated the creep-fatigue damage of reactor coolant piping in Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF0) loop. By designing different stress cycles, the different creep strain increment and creep-fatigue assessments are systematically conducted. It was found that the creep strain increment of SF0 loop accumulated in one stress cycle time was higher than that accumulated during the entire service life divided by the number of stress cycles. Similarly, the creep-fatigue damage was also lower. And thus, an economical design method for SF0/SF1 loop was given.
{"title":"Numerical Study on the Creep Strain Characteristics for TMSR Reactor Coolant Piping Under Thermal Loading","authors":"W. Gong, Xiaochun Zhang, Mingqiang Xie, Yuan Fu, Xiao Wang","doi":"10.1115/PVP2018-84333","DOIUrl":"https://doi.org/10.1115/PVP2018-84333","url":null,"abstract":"Safety assessment and economy design of piping system at high temperature is highly important in power and nuclear engineering. ASME-NH code considers cyclic failure modes at elevated temperature and provides the rules and damage limits for creep-fatigue interaction. Based on the calculation methods for creep strain increment of ASME-NH code, this paper investigated the creep-fatigue damage of reactor coolant piping in Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF0) loop. By designing different stress cycles, the different creep strain increment and creep-fatigue assessments are systematically conducted. It was found that the creep strain increment of SF0 loop accumulated in one stress cycle time was higher than that accumulated during the entire service life divided by the number of stress cycles. Similarly, the creep-fatigue damage was also lower. And thus, an economical design method for SF0/SF1 loop was given.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"55 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88562962","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}
Fakun Zhuang, S. Tu, Guoshan Xie, S. Shao, Luowei Cao
Small specimen creep test techniques have been widely applied in the creep properties assessment of materials for the equipment in-service. In order to acquire the creep data accurately and conveniently, the creep test method with small cantilever beam specimens is proposed. On the basis of Norton-Bailey creep law, analytical creep model for the cantilever beam specimen is derived. With this model, the load can be converted to equivalent uniaxial stress and the displacement rate can be converted to equivalent uniaxial strain rate. The creep properties of Cr-Mo steel are assessed by the cantilever beam specimens creep tests. And the creep parameters are evaluated, which are compared to the uniaxial creep parameters. The results show that parameters obtained from the cantilever beam tests correspond reasonably well with those from uniaxial tests. It proves that the primary and secondary creep properties can be assessed by the cantilever beam specimen tests.
{"title":"Creep Properties Assessment of Materials by a Small Cantilever Beam Specimen","authors":"Fakun Zhuang, S. Tu, Guoshan Xie, S. Shao, Luowei Cao","doi":"10.1115/PVP2018-84135","DOIUrl":"https://doi.org/10.1115/PVP2018-84135","url":null,"abstract":"Small specimen creep test techniques have been widely applied in the creep properties assessment of materials for the equipment in-service. In order to acquire the creep data accurately and conveniently, the creep test method with small cantilever beam specimens is proposed. On the basis of Norton-Bailey creep law, analytical creep model for the cantilever beam specimen is derived. With this model, the load can be converted to equivalent uniaxial stress and the displacement rate can be converted to equivalent uniaxial strain rate. The creep properties of Cr-Mo steel are assessed by the cantilever beam specimens creep tests. And the creep parameters are evaluated, which are compared to the uniaxial creep parameters. The results show that parameters obtained from the cantilever beam tests correspond reasonably well with those from uniaxial tests. It proves that the primary and secondary creep properties can be assessed by the cantilever beam specimen tests.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89946718","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}
Evaluation of mechanical properties from a small amount of test material is essential in a process of development of new alloys and monitoring of degradation and damage of in-service components. Miniaturization of test specimen together with a specimen reconstitution technique enable direct method of evaluation of mechanical properties if only a limited amount of material is available. Modern types of construction steels and alloys possess improved metallurgical and mechanical properties and also demonstrate a high level of fracture toughness. In these cases, correlation and correction of size effect on fracture toughness between different specimen sizes are needed. Within this contribution the size effect of J-R curve for two sizes of three-point-bend specimens is solved. The J-R curves of three-point-bend specimens with cross-section 10 × 10 mm and 3 × 4 mm of P91 and Eurofer97 steel were determined at room temperature. The existence of size effect is proven as larger specimens demonstrate higher J-R curves than smaller specimens for both steels. Simplified mechanical model based on fracture energies initially proposed by Schindler et al. is applied to describe experimentally determined J-R curves and to predict the observed size effect between used specimen sizes.
{"title":"The Size Effect on J-R Curve for Construction Steels and its Prediction by Simplified Mechanical Model","authors":"L. Stratil, F. Šiška, H. Hadraba, I. Dlouhý","doi":"10.1115/PVP2018-84866","DOIUrl":"https://doi.org/10.1115/PVP2018-84866","url":null,"abstract":"Evaluation of mechanical properties from a small amount of test material is essential in a process of development of new alloys and monitoring of degradation and damage of in-service components. Miniaturization of test specimen together with a specimen reconstitution technique enable direct method of evaluation of mechanical properties if only a limited amount of material is available. Modern types of construction steels and alloys possess improved metallurgical and mechanical properties and also demonstrate a high level of fracture toughness. In these cases, correlation and correction of size effect on fracture toughness between different specimen sizes are needed. Within this contribution the size effect of J-R curve for two sizes of three-point-bend specimens is solved. The J-R curves of three-point-bend specimens with cross-section 10 × 10 mm and 3 × 4 mm of P91 and Eurofer97 steel were determined at room temperature. The existence of size effect is proven as larger specimens demonstrate higher J-R curves than smaller specimens for both steels. Simplified mechanical model based on fracture energies initially proposed by Schindler et al. is applied to describe experimentally determined J-R curves and to predict the observed size effect between used specimen sizes.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"17 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81718138","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}
A. Kovács, Á. Horváth, M. Horváth, I. Szenthe, F. Gillemot
In order to evaluate the fracture toughness and Master Curve, the exact values of the tensile and yield strengths determined at the fracture toughness testing temperature are required. The fracture toughness should be measured around the T0 reference temperature to obtain valid results. Tensile properties — if exists — are generally measured at room temperature, and at the maximum operation temperature of the pressure vessel. If T0 reference temperature is different from these temperatures a correction formula is used for interpolation or extrapolation. Several times even no reliable tensile results are available, especially in the case of irradiated specimens. Use of irradiated Charpy remnants is a solution to produce tensile bars. A specific method has been developed to produce 12 pieces of flat 2*1 mm cross section tensile specimens made from one half Charpy specimens using stud weld reconstitution. The specimen cutting is made using thin molybdenum wire spark cutting. In order to do so difficulties of handling the small radioactive specimens needed to be solved, special grips were developed for the tensile machine to clamp the miniature tensile specimens. A laser extensometer was used to measure the strain and a video microscope was used to determine the cross section after the fracture without touching the specimen. The paper presents the results obtained on more than 200 irradiated small tensile specimens. These results had been compared to the old surveillance ones.
{"title":"Reconstituted Mini Tensile Specimens","authors":"A. Kovács, Á. Horváth, M. Horváth, I. Szenthe, F. Gillemot","doi":"10.1115/PVP2018-84760","DOIUrl":"https://doi.org/10.1115/PVP2018-84760","url":null,"abstract":"In order to evaluate the fracture toughness and Master Curve, the exact values of the tensile and yield strengths determined at the fracture toughness testing temperature are required. The fracture toughness should be measured around the T0 reference temperature to obtain valid results. Tensile properties — if exists — are generally measured at room temperature, and at the maximum operation temperature of the pressure vessel. If T0 reference temperature is different from these temperatures a correction formula is used for interpolation or extrapolation. Several times even no reliable tensile results are available, especially in the case of irradiated specimens. Use of irradiated Charpy remnants is a solution to produce tensile bars. A specific method has been developed to produce 12 pieces of flat 2*1 mm cross section tensile specimens made from one half Charpy specimens using stud weld reconstitution. The specimen cutting is made using thin molybdenum wire spark cutting. In order to do so difficulties of handling the small radioactive specimens needed to be solved, special grips were developed for the tensile machine to clamp the miniature tensile specimens. A laser extensometer was used to measure the strain and a video microscope was used to determine the cross section after the fracture without touching the specimen. The paper presents the results obtained on more than 200 irradiated small tensile specimens. These results had been compared to the old surveillance ones.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"12 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74564273","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}
For commercial development and deployment of the molten salt reactor, a structural alloy that provides both strength at high temperature and resistance to very corrosive molten salt environment is required. To meet this requirement, a survey is conducted on domestic and international candidate alloys. Alloy N turns out to be the sole frontrunner in readiness for qualification to enable the desired deployment within an estimated 10 years. A review of the qualification for commercial nuclear applications indicates that Alloy N has met a large portion of the requirements. Gaps in the qualification are also identified. A search for historical data is underway to retrieve information needed for filling the gaps and upgrading the qualification. Scope of the discovered historical data is briefly discussed and strategic planning for research and development pathway is suggested to ensure successful evolution in commercial deployment of the molten salt reactor system.
{"title":"Considerations of Alloy N Code Extension for Commercial Molten Salt Reactor Development and Deployment","authors":"W. Ren","doi":"10.1115/PVP2018-84716","DOIUrl":"https://doi.org/10.1115/PVP2018-84716","url":null,"abstract":"For commercial development and deployment of the molten salt reactor, a structural alloy that provides both strength at high temperature and resistance to very corrosive molten salt environment is required. To meet this requirement, a survey is conducted on domestic and international candidate alloys. Alloy N turns out to be the sole frontrunner in readiness for qualification to enable the desired deployment within an estimated 10 years. A review of the qualification for commercial nuclear applications indicates that Alloy N has met a large portion of the requirements. Gaps in the qualification are also identified. A search for historical data is underway to retrieve information needed for filling the gaps and upgrading the qualification. Scope of the discovered historical data is briefly discussed and strategic planning for research and development pathway is suggested to ensure successful evolution in commercial deployment of the molten salt reactor system.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"58 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79027173","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. Koshiishi, Satoru Aoike, T. Hashimoto, Yuya Hideki
The Hashimoto-Koshiishi model was simplified from a mathematical viewpoint taking into consideration explicit crack growth rate (CGR) dependency on the stress intensity factor to enable practicable use of the model in the integrity assessment of boiling water reactor components. Then simplified equations were developed for different strain distribution cases in the model. The developed simplified equations were applied to analyze CGRs of cold worked stainless steels and the simulated weld heat affected zone of the primary loop recirculation piping and core shroud and these equations reproduced CGRs well.
{"title":"Formulation of the Stress Corrosion Crack Growth Rates for the Practical Use in the Integrity Assessment of Nuclear Reactor Components","authors":"M. Koshiishi, Satoru Aoike, T. Hashimoto, Yuya Hideki","doi":"10.1115/PVP2018-84213","DOIUrl":"https://doi.org/10.1115/PVP2018-84213","url":null,"abstract":"The Hashimoto-Koshiishi model was simplified from a mathematical viewpoint taking into consideration explicit crack growth rate (CGR) dependency on the stress intensity factor to enable practicable use of the model in the integrity assessment of boiling water reactor components. Then simplified equations were developed for different strain distribution cases in the model. The developed simplified equations were applied to analyze CGRs of cold worked stainless steels and the simulated weld heat affected zone of the primary loop recirculation piping and core shroud and these equations reproduced CGRs well.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"29 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75501324","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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