A program for a high-temperature design analysis and defect assessment has been developed for an elevated temperature evaluation according to the RCC-MRx for Generation IV and fusion reactor systems. The program, called ‘HITEP_RCC-MRx,’ consists of three modules: ‘HITEP_RCC-DBA,’ which computerizes the design-by-analysis (DBA) for class 1 components such as the pressure vessel and heat exchangers according to RB-3200 procedures, ‘HITEP_RCC-PIPE,’ which computerizes the design-by-rule (DBR) analysis for class 1 piping according to RB-3600 procedures and ‘HITEP_RCC-A16,’ which computerizes high-temperature defect assessment according to the A16 procedures. It is a web-based program, and thus can operate on a smartphone as well as on a personal computer once it is connected to the URL. The program has been verified with a number of relevant example problems on DBA, Pipe, and A16. It was shown from the verification works that HITEP_RCC-MRx with the three modules conducts a design evaluation and a defect assessment in an efficient and reliable way.
{"title":"Development of a Program for High-Temperature Design Analysis and Defect Assessment According to RCC-MRx","authors":"H. Lee, Min-Gu Won, N. Huh, Woo-Gon Kim","doi":"10.1115/PVP2018-84242","DOIUrl":"https://doi.org/10.1115/PVP2018-84242","url":null,"abstract":"A program for a high-temperature design analysis and defect assessment has been developed for an elevated temperature evaluation according to the RCC-MRx for Generation IV and fusion reactor systems. The program, called ‘HITEP_RCC-MRx,’ consists of three modules: ‘HITEP_RCC-DBA,’ which computerizes the design-by-analysis (DBA) for class 1 components such as the pressure vessel and heat exchangers according to RB-3200 procedures, ‘HITEP_RCC-PIPE,’ which computerizes the design-by-rule (DBR) analysis for class 1 piping according to RB-3600 procedures and ‘HITEP_RCC-A16,’ which computerizes high-temperature defect assessment according to the A16 procedures. It is a web-based program, and thus can operate on a smartphone as well as on a personal computer once it is connected to the URL. The program has been verified with a number of relevant example problems on DBA, Pipe, and A16. It was shown from the verification works that HITEP_RCC-MRx with the three modules conducts a design evaluation and a defect assessment in an efficient and reliable way.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"20 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121067829","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}
Reactor internals are components that are no typical pressure boundary but they are nevertheless very important as they hold fuel elements and all reactor control system elements and thus must ensure their safe and reliable operation during the whole reactor life under all operating and even beyond bases regimes. In principle, reactor internals can be replaced but their weight, quantity of very high activated material and cost such possibility practically excluded. Thus, evaluation of the reactor internals condition and prediction of their behavior during the whole or even extended lifetime is of high importance. Reactor internals are subjected to very high neutron irradiation that could initiated not only stress corrosion (irradiation assisted) cracking but also large embrittlement and changes in dimensions (swelling and creep). VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was initiated and co-ordinated by the Czech and was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for Nuclear Power Plants (NPPs) with WWER (Water-Water-Energetic-Reactor = PWR type) type reactors, as those codes were developed only for design and manufacture and were not changed since their second edition in 1989. VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes. Within the last upgrading and principal extending of this VERLIFE Procedure was developed within the 3-years IAEA project (in close co-operation with another project of the 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) that started in 2009 with final approval and editing in 2013”) a part dealing with the evaluation of reactor internals lifetime was elaborated.. This IAEA VERLIFE procedure for internals has been implemented into the existing Normative Technical Documentation (NTD) ASI (Czech Association of Mechanical Engineers), Section IV – Evaluation of Residual Lifetime of Components and Piping in WWER type NPPs. Main damaging mechanisms that should be taken into account in reactor internals and the procedure are described in detail with necessary formulae for materials of internals: - Radiation hardening - Radiation embrittlement - Radiation swelling - Radiation creep - Swelling under stress effect - Swelling inducing embrittlement - Irradiated assisted stress corrosion cracking - Transformation austenite-ferrite and also the method for evaluation of the resistance against non-ductile failure of postulated defect. The paper will describe these main principles and also more detailed information on the procedure for evaluation of reactor internals w
{"title":"Evaluation of Reactor Internals Integrity and Lifetime According to the NTD ASI","authors":"M. Brumovsky","doi":"10.1115/PVP2018-84140","DOIUrl":"https://doi.org/10.1115/PVP2018-84140","url":null,"abstract":"Reactor internals are components that are no typical pressure boundary but they are nevertheless very important as they hold fuel elements and all reactor control system elements and thus must ensure their safe and reliable operation during the whole reactor life under all operating and even beyond bases regimes.\u0000 In principle, reactor internals can be replaced but their weight, quantity of very high activated material and cost such possibility practically excluded.\u0000 Thus, evaluation of the reactor internals condition and prediction of their behavior during the whole or even extended lifetime is of high importance. Reactor internals are subjected to very high neutron irradiation that could initiated not only stress corrosion (irradiation assisted) cracking but also large embrittlement and changes in dimensions (swelling and creep).\u0000 VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation” was initiated and co-ordinated by the Czech and was developed within the 5th Framework Program of the European Union in 2003 and later upgraded within the 6th Framework Program “COVERS – Safety of WWER NPPs” of the European Union in 2008. This Procedure had to fill the gap in original Soviet/Russian Codes and Rules for Nuclear Power Plants (NPPs) with WWER (Water-Water-Energetic-Reactor = PWR type) type reactors, as those codes were developed only for design and manufacture and were not changed since their second edition in 1989.\u0000 VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes.\u0000 Within the last upgrading and principal extending of this VERLIFE Procedure was developed within the 3-years IAEA project (in close co-operation with another project of the 6th Framework Program of the European Union “NULIFE – Plant Life Management of NPPs”) that started in 2009 with final approval and editing in 2013”) a part dealing with the evaluation of reactor internals lifetime was elaborated..\u0000 This IAEA VERLIFE procedure for internals has been implemented into the existing Normative Technical Documentation (NTD) ASI (Czech Association of Mechanical Engineers), Section IV – Evaluation of Residual Lifetime of Components and Piping in WWER type NPPs.\u0000 Main damaging mechanisms that should be taken into account in reactor internals and the procedure are described in detail with necessary formulae for materials of internals:\u0000 - Radiation hardening\u0000 - Radiation embrittlement\u0000 - Radiation swelling\u0000 - Radiation creep\u0000 - Swelling under stress effect\u0000 - Swelling inducing embrittlement\u0000 - Irradiated assisted stress corrosion cracking\u0000 - Transformation austenite-ferrite\u0000 and also the method for evaluation of the resistance against non-ductile failure of postulated defect.\u0000 The paper will describe these main principles and also more detailed information on the procedure for evaluation of reactor internals w","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123592902","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}
Environmentally Assisted Fatigue (EAF) is receiving nowadays an increased level of attention for existing Nuclear Power Plants (NPPs) as utilities are now working to extend their life. In the wake of numerous experimental fatigue tests carried out in air and also in a PWR environment, the French RCC-M code [1] has recently been amended (in its 2016 edition) with two Rules in Probatory Phase (RPP), equivalent to ASME code-cases, “RPP-2” and “RPP-3” [2] [3]. RPP-2 consists of an update of the design fatigue curve in air for stainless steels (SSs) and nickel-based alloys, and is also associated with RPP-3 which provides guidelines for incorporating the environmental penalty “Fen” factor in fatigue usage factor calculations. Alongside this codification effort, an EAF screening has recently been carried out within EDF DT [4] on various areas of the primary circuit of the 900 MWe plants of the EDF fleet. This screening led to the identification of a list of 35 “sentinel locations” which are defined as areas most prone to EAF degradation process. These locations will be subjected to detailed EAF analysis in the stress report calculations (according to the above-mentioned RCC-M code cases) for the fourth decennial inspection of the 900 MWe (VD4 900 MWe) power plants. The potential impact of EAF on the secondary circuit components is another question to address in anticipation of the VD4 900 MWe, as they may be considered as class 1 or class 2 equipment for RCC-M application according to the equipment specification. This paper presents the approach proposed by EDF towards an exemption of environmental effects consideration for secondary circuit components. The argument is first based on a review of experimental campaigns led in Japan and France (respectively on fatigue test specimens and at the component scale) which indicate a Dissolved Oxygen (DO) content threshold below which environmental effects are almost inexistent. The (conservative) value of 40 ppb has been selected consistently with NUREG/CR-6909 revision 0 [5]. The second part of the argument is built, on the one hand, on the analysis of the EDF Technical Specifications for Operation (STE) which narrows the scope of the study only to unit outages, and, on the other hand, on the analysis of 5 years of operations of all 900 MWe plants of the EDF fleet (equivalent to 170 reactor-years). It has been shown that the DO content rarely exceeded the 40 ppb threshold in the secondary coolant, and that in this case, the considered locations were not submitted to any fatigue loading.
{"title":"Environmental Assisted Fatigue and EDF 900 MWe PWRs Fleet: Towards an Exemption of Environmental Effects Consideration for Secondary Circuit Components","authors":"Sam Cuvilliez, G. Léopold, T. Métais","doi":"10.1115/PVP2018-84301","DOIUrl":"https://doi.org/10.1115/PVP2018-84301","url":null,"abstract":"Environmentally Assisted Fatigue (EAF) is receiving nowadays an increased level of attention for existing Nuclear Power Plants (NPPs) as utilities are now working to extend their life. In the wake of numerous experimental fatigue tests carried out in air and also in a PWR environment, the French RCC-M code [1] has recently been amended (in its 2016 edition) with two Rules in Probatory Phase (RPP), equivalent to ASME code-cases, “RPP-2” and “RPP-3” [2] [3]. RPP-2 consists of an update of the design fatigue curve in air for stainless steels (SSs) and nickel-based alloys, and is also associated with RPP-3 which provides guidelines for incorporating the environmental penalty “Fen” factor in fatigue usage factor calculations.\u0000 Alongside this codification effort, an EAF screening has recently been carried out within EDF DT [4] on various areas of the primary circuit of the 900 MWe plants of the EDF fleet. This screening led to the identification of a list of 35 “sentinel locations” which are defined as areas most prone to EAF degradation process. These locations will be subjected to detailed EAF analysis in the stress report calculations (according to the above-mentioned RCC-M code cases) for the fourth decennial inspection of the 900 MWe (VD4 900 MWe) power plants.\u0000 The potential impact of EAF on the secondary circuit components is another question to address in anticipation of the VD4 900 MWe, as they may be considered as class 1 or class 2 equipment for RCC-M application according to the equipment specification. This paper presents the approach proposed by EDF towards an exemption of environmental effects consideration for secondary circuit components. The argument is first based on a review of experimental campaigns led in Japan and France (respectively on fatigue test specimens and at the component scale) which indicate a Dissolved Oxygen (DO) content threshold below which environmental effects are almost inexistent. The (conservative) value of 40 ppb has been selected consistently with NUREG/CR-6909 revision 0 [5]. The second part of the argument is built, on the one hand, on the analysis of the EDF Technical Specifications for Operation (STE) which narrows the scope of the study only to unit outages, and, on the other hand, on the analysis of 5 years of operations of all 900 MWe plants of the EDF fleet (equivalent to 170 reactor-years). It has been shown that the DO content rarely exceeded the 40 ppb threshold in the secondary coolant, and that in this case, the considered locations were not submitted to any fatigue loading.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"2013 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114451350","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}
An irradiated low-upper-shelf Linde 80 weld metal has been tested by four laboratories as part of an inter-laboratory assessment of use of the miniature compact tension [mini-C(T)] test specimen for Master Curve fracture toughness evaluation following ASTM E1921. The preliminary results from each of the laboratories have been compiled and evaluated together to assess the validity and use of the mini-C(T) specimen for an irradiated reactor pressure vessel material which can exhibit ductile crack growth at low temperatures relative to cleavage initiation fracture toughness. The preliminary results from this mini-C(T) testing can also be compared to extensive specimen test results from larger C(T) specimens of the same irradiated material. Comparisons of the results from each of the laboratories and some inter-laboratory differences in the fracture testing are assessed. The evaluations indicate reasonable agreement between the mini-C(T) and larger specimen results, but the selection of test temperature and the number of test specimens needed to obtain reliable results are more difficult when testing a low-upper-shelf toughness material.
{"title":"Inter-Laboratory Results and Analyses of Mini-C(T) Specimen Testing of an Irradiated Linde 80 Weld Metal","authors":"W. Server, M. Sokolov, Masato Yamamoto, R. Carter","doi":"10.1115/PVP2018-84950","DOIUrl":"https://doi.org/10.1115/PVP2018-84950","url":null,"abstract":"An irradiated low-upper-shelf Linde 80 weld metal has been tested by four laboratories as part of an inter-laboratory assessment of use of the miniature compact tension [mini-C(T)] test specimen for Master Curve fracture toughness evaluation following ASTM E1921. The preliminary results from each of the laboratories have been compiled and evaluated together to assess the validity and use of the mini-C(T) specimen for an irradiated reactor pressure vessel material which can exhibit ductile crack growth at low temperatures relative to cleavage initiation fracture toughness. The preliminary results from this mini-C(T) testing can also be compared to extensive specimen test results from larger C(T) specimens of the same irradiated material. Comparisons of the results from each of the laboratories and some inter-laboratory differences in the fracture testing are assessed. The evaluations indicate reasonable agreement between the mini-C(T) and larger specimen results, but the selection of test temperature and the number of test specimens needed to obtain reliable results are more difficult when testing a low-upper-shelf toughness material.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"14 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128054457","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}
Uijeong Ro, Jeong Hwan Kim, Hoomin Lee, Seok-Jun Kang, M. Kim
The Sodium Fast-cooled Reactor (SFR), are generation IV nuclear power plants, have a target operating temperature of 550°C which makes creep-fatigue behavior more critical than a generation III nuclear power plants. So it is important to understand the nature of creep-fatigue behavior of the piping material, Grade 91 steel. The creep-fatigue damage diagram of Grade 91 steel used in ASME-NH was derived using a conventional time-fraction testing method which was originally developed for type 300 stainless steels. Multiple studies indicate that the creep-fatigue damage diagram of Grade 91 steel developed using this testing method has excessive conservatism in it. Therefore, an alternative testing method was suggested by separating creep and fatigue using interrupted creep tests. The suggested method makes it possible to control creep life consumption freely which was difficult with the previous method. It also makes it easier to observe the interaction between creep and fatigue mechanisms and microstructural evolution. In conclusion, an alternative creep-fatigue damage diagram for Grade 91 steel at 550°C was developed using an interrupt creep fatigue testing method and FE model simulation.
{"title":"Creep-Fatigue Damage Evaluation of Grade 91 Steel Using Interrupt Creep Fatigue Test","authors":"Uijeong Ro, Jeong Hwan Kim, Hoomin Lee, Seok-Jun Kang, M. Kim","doi":"10.1115/PVP2018-84561","DOIUrl":"https://doi.org/10.1115/PVP2018-84561","url":null,"abstract":"The Sodium Fast-cooled Reactor (SFR), are generation IV nuclear power plants, have a target operating temperature of 550°C which makes creep-fatigue behavior more critical than a generation III nuclear power plants. So it is important to understand the nature of creep-fatigue behavior of the piping material, Grade 91 steel. The creep-fatigue damage diagram of Grade 91 steel used in ASME-NH was derived using a conventional time-fraction testing method which was originally developed for type 300 stainless steels. Multiple studies indicate that the creep-fatigue damage diagram of Grade 91 steel developed using this testing method has excessive conservatism in it. Therefore, an alternative testing method was suggested by separating creep and fatigue using interrupted creep tests. The suggested method makes it possible to control creep life consumption freely which was difficult with the previous method. It also makes it easier to observe the interaction between creep and fatigue mechanisms and microstructural evolution. In conclusion, an alternative creep-fatigue damage diagram for Grade 91 steel at 550°C was developed using an interrupt creep fatigue testing method and FE model simulation.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"22 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115492651","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 use of miniature compact tension (mini-CT) specimens for fracture mechanics was experimentally demonstrated to allow the characterization of ferritic steels in the transition regime. In particular, the master curve transition temperature T0 can confidently be determined according to the ASTM E1921 standard using mini-CT specimens. This means that specimen size effect is well taken into account if loss of constraint is limited by restricting the test temperature range to remain below the allowed maximum loading level. In the upper shelf ductile regime, where stable crack growth occurs, a number of challenges should be overcome to use such a geometry to derive the crack resistance curve, or JR-curve, transferrable to a structure. Indeed, despite a large scatter, the experimental data on several materials suggest a size effect that underestimates the crack resistance when reducing specimen size. The crack resistance behavior of several reactor pressure vessel materials was investigated with square-sized ligament compact tension specimens of various size ranging from 1 inch-thickness (B = 25 mm) to the smallest thickness (B = 4.2 mm) of the mini-CT. The main objective of this paper is to estimate the crack resistance behavior of RPV steels that would be obtained with a standard 1T-CT specimen by using mini-CT with the appropriate specimen size correction. After a series of scaling attempts that were not successful, based on a simple dimensional analysis, a simple analytical formulation based on specimen thickness and ligament is suggested to account for specimen size effect for the CT geometry. Reasonable agreement could generally be found on a number of RPV materials between crack resistance measured with mini-CT and standard 1T-CT specimens.
{"title":"Crack Resistance Curve Measurement With Miniaturized CT Specimen","authors":"R. Chaouadi, M. Lambrecht, R. Gérard","doi":"10.1115/PVP2018-84690","DOIUrl":"https://doi.org/10.1115/PVP2018-84690","url":null,"abstract":"The use of miniature compact tension (mini-CT) specimens for fracture mechanics was experimentally demonstrated to allow the characterization of ferritic steels in the transition regime. In particular, the master curve transition temperature T0 can confidently be determined according to the ASTM E1921 standard using mini-CT specimens. This means that specimen size effect is well taken into account if loss of constraint is limited by restricting the test temperature range to remain below the allowed maximum loading level. In the upper shelf ductile regime, where stable crack growth occurs, a number of challenges should be overcome to use such a geometry to derive the crack resistance curve, or JR-curve, transferrable to a structure. Indeed, despite a large scatter, the experimental data on several materials suggest a size effect that underestimates the crack resistance when reducing specimen size.\u0000 The crack resistance behavior of several reactor pressure vessel materials was investigated with square-sized ligament compact tension specimens of various size ranging from 1 inch-thickness (B = 25 mm) to the smallest thickness (B = 4.2 mm) of the mini-CT. The main objective of this paper is to estimate the crack resistance behavior of RPV steels that would be obtained with a standard 1T-CT specimen by using mini-CT with the appropriate specimen size correction. After a series of scaling attempts that were not successful, based on a simple dimensional analysis, a simple analytical formulation based on specimen thickness and ligament is suggested to account for specimen size effect for the CT geometry. Reasonable agreement could generally be found on a number of RPV materials between crack resistance measured with mini-CT and standard 1T-CT specimens.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124418846","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}
F. Gillemot, M. Horváth, Á. Horváth, I. Szenthe, A. Kovács
The original WWER-440 surveillance had 6 sets of specimens and each set had 12 Charpy, 12 COD (crack opening displacement) and 6 tensile specimens made from base material, weldment and HAZ (heat affected zone). The Charpy size precrack TPB (three point bend) COD specimens were located at the end of the chains, where the flux is rapidly decreasing. During the period of 1970–90, when the WWER-440-V213 units were designed, built and started to operate, the Charpy impact transition curve measurement was the accepted method to evaluate the radiation embrittlement. The technology and the standards to use small size fracture mechanical specimens in surveillance capsules were not developed at the time period when most of the second generation reactors — including the WWER-440 V 213 type — were designed, therefore the fracture toughness specimens were considered less interesting for the utilities and the safety authorities. Fracture toughness curves were elaborated in the laboratories on large size unirradiated specimens and radiation embrittlement adjustments were made according to the Charpy shift. However, during the past 30 years fracture mechanics has rapidly developed, and the testing moved to the direction of using small and mini sized specimens. The development of the Master Curve evaluation method [4,5] allowed the use of small specimens for fracture toughness testing in surveillance programs, and the results obtained on irradiated specimens may be used directly in the lifetime evaluation. The purpose of this work was to develop a specimen production technology and testing procedure to measure these data using the remnants of irradiated surveillance Charpy specimens, and the comparison of the data calculated from CMOD and LLD on irradiated CrMoV type RPV material and weldment.
原始WWER-440监测有6组试件，每组试件有12个Charpy、12个COD(裂纹张开位移)和6个由母材、焊件和HAZ(热影响区)制成的拉伸试件。chpy尺寸预裂TPB(三点弯曲)COD试样位于链的末端，其通量迅速下降。在1970 - 1990年WWER-440-V213机组设计、建造和开始运行期间，Charpy冲击过渡曲线测量是公认的辐射脆化评价方法。在大多数第二代反应堆(包括WWER-440 V - 213型)设计时，在监测胶囊中使用小尺寸断裂力学试样的技术和标准尚未开发，因此，对公用事业和安全当局来说，断裂韧性试样被认为不那么有趣。在实验室对大尺寸未辐照试样进行了断裂韧性曲线的绘制，并根据Charpy位移进行了辐射脆化调整。然而，近30年来，断裂力学得到了迅速发展，试验向着使用小、微型试样的方向发展。主曲线评估方法的发展[4,5]允许在监测项目中使用小试样进行断裂韧性测试，并且在辐照试样上获得的结果可直接用于寿命评估。这项工作的目的是开发一种样品生产技术和测试程序，使用辐照后的监视Charpy样品的残余来测量这些数据，并将辐照后的CrMoV型RPV材料和焊件的CMOD和LLD计算的数据进行比较。
{"title":"Master Curve Testing on Reconstituted Surveillance Charpy Specimens","authors":"F. Gillemot, M. Horváth, Á. Horváth, I. Szenthe, A. Kovács","doi":"10.1115/PVP2018-84749","DOIUrl":"https://doi.org/10.1115/PVP2018-84749","url":null,"abstract":"The original WWER-440 surveillance had 6 sets of specimens and each set had 12 Charpy, 12 COD (crack opening displacement) and 6 tensile specimens made from base material, weldment and HAZ (heat affected zone). The Charpy size precrack TPB (three point bend) COD specimens were located at the end of the chains, where the flux is rapidly decreasing.\u0000 During the period of 1970–90, when the WWER-440-V213 units were designed, built and started to operate, the Charpy impact transition curve measurement was the accepted method to evaluate the radiation embrittlement. The technology and the standards to use small size fracture mechanical specimens in surveillance capsules were not developed at the time period when most of the second generation reactors — including the WWER-440 V 213 type — were designed, therefore the fracture toughness specimens were considered less interesting for the utilities and the safety authorities. Fracture toughness curves were elaborated in the laboratories on large size unirradiated specimens and radiation embrittlement adjustments were made according to the Charpy shift. However, during the past 30 years fracture mechanics has rapidly developed, and the testing moved to the direction of using small and mini sized specimens. The development of the Master Curve evaluation method [4,5] allowed the use of small specimens for fracture toughness testing in surveillance programs, and the results obtained on irradiated specimens may be used directly in the lifetime evaluation. The purpose of this work was to develop a specimen production technology and testing procedure to measure these data using the remnants of irradiated surveillance Charpy specimens, and the comparison of the data calculated from CMOD and LLD on irradiated CrMoV type RPV material and weldment.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"30 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123923347","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}
Mini-CT specimens are becoming a highly popular geometry for use in reactor pressure vessel (RPV) community for direct measurement of fracture toughness in the transition region using the Master Curve methodology. In the present study, Mini-CT specimens were machined from previously tested Charpy specimens of the Midland low upper-shelf Linde 80 weld in both, unirradiated and irradiated conditions. The irradiated specimens have been characterized as part of a joint ORNL-EPRI-CRIEPI collaborative program. The Linde 80 weld was selected because it has been extensively characterized in the irradiated condition by conventional specimens, and because of the need to validate application of Mini-CT specimens for low upper-shelf materials — a more likely case for some irradiated materials of older generation RPVs. It is shown that the fracture toughness reference temperatures, To, derived from these Mini-CT specimens are in good agreement with To values previously recorded for this material in the unirradiated and irradiated conditions. However, this study indicates that in real practice it is highly advisable to use a much larger number of specimens than the minimum number prescribed in ASTM E1921.
{"title":"Use of Mini-CT Specimens for Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Before and After Irradiation","authors":"M. Sokolov","doi":"10.1115/PVP2018-84804","DOIUrl":"https://doi.org/10.1115/PVP2018-84804","url":null,"abstract":"Mini-CT specimens are becoming a highly popular geometry for use in reactor pressure vessel (RPV) community for direct measurement of fracture toughness in the transition region using the Master Curve methodology. In the present study, Mini-CT specimens were machined from previously tested Charpy specimens of the Midland low upper-shelf Linde 80 weld in both, unirradiated and irradiated conditions. The irradiated specimens have been characterized as part of a joint ORNL-EPRI-CRIEPI collaborative program. The Linde 80 weld was selected because it has been extensively characterized in the irradiated condition by conventional specimens, and because of the need to validate application of Mini-CT specimens for low upper-shelf materials — a more likely case for some irradiated materials of older generation RPVs. It is shown that the fracture toughness reference temperatures, To, derived from these Mini-CT specimens are in good agreement with To values previously recorded for this material in the unirradiated and irradiated conditions. However, this study indicates that in real practice it is highly advisable to use a much larger number of specimens than the minimum number prescribed in ASTM E1921.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"37 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125975641","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. Ha, T. Tobita, H. Takamizawa, S. Hanawa, Y. Nishiyama
An evaluation of the fracture toughness of the heat-affected zone (HAZ), which is located under the weld overlay cladding of a reactor pressure vessel (RPV), was performed. Considering inhomogeneous microstructures of the HAZ, 0.4T-C(T) specimens were manufactured from the cladding strips locations, and Mini-C(T) specimens were fabricated from the distanced location as well as under the cladding. The reference temperature (To) of specimens that were aligned with the middle section of a cladding strip (HAZMCS) was ∼12°C higher than that of specimens that were aligned with cladding strips at the overlap (HAZOCS). To values of partial area in the HAZ were obtained using Mini-C(T) specimen. The To values obtained near the side of the cladding were ∼13°C higher than those away from the cladding. To values of HAZ for both 0.4T-C(T) and Mini-C(T) specimens were significantly lower than that of the base metal at a quarter thickness by 40°C–60°C. Compared to the literature data that indicated fracture toughness at the surface without overlay cladding and base metal of a quarter thickness in a pressure vessel plate, this study concluded that the welding thermal history showed no significant effect on the fracture toughness of the inner surface of RPV steel.
{"title":"Fracture Toughness Evaluation of Heat-Affected Zone Under Weld Overlay Cladding in Reactor Pressure Vessel Steel","authors":"Y. Ha, T. Tobita, H. Takamizawa, S. Hanawa, Y. Nishiyama","doi":"10.1115/PVP2018-84535","DOIUrl":"https://doi.org/10.1115/PVP2018-84535","url":null,"abstract":"An evaluation of the fracture toughness of the heat-affected zone (HAZ), which is located under the weld overlay cladding of a reactor pressure vessel (RPV), was performed. Considering inhomogeneous microstructures of the HAZ, 0.4T-C(T) specimens were manufactured from the cladding strips locations, and Mini-C(T) specimens were fabricated from the distanced location as well as under the cladding. The reference temperature (To) of specimens that were aligned with the middle section of a cladding strip (HAZMCS) was ∼12°C higher than that of specimens that were aligned with cladding strips at the overlap (HAZOCS). To values of partial area in the HAZ were obtained using Mini-C(T) specimen. The To values obtained near the side of the cladding were ∼13°C higher than those away from the cladding. To values of HAZ for both 0.4T-C(T) and Mini-C(T) specimens were significantly lower than that of the base metal at a quarter thickness by 40°C–60°C. Compared to the literature data that indicated fracture toughness at the surface without overlay cladding and base metal of a quarter thickness in a pressure vessel plate, this study concluded that the welding thermal history showed no significant effect on the fracture toughness of the inner surface of RPV steel.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"13 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126115266","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 load and temperature history during pressurized thermal shock (PTS) event is highly depending on the crack edge location in wall thickness direction of a reactor pressure vessel (RPV) beltline region. Therefore, the consideration of plant specific through-wall fracture toughness distribution, which is not considered in the current codes and regulations [1,2], may improve the structural integrity assessment for PTS event. The Master Curve (MC) method [3,4] is one of the methods, which can directory evaluate the fracture toughness of ferritic materials with relatively low number of any size of specimens. CRIEPI has proposed the use of very small C(T) (Mini-C(T)) specimens for the MC method. The appropriateness of Mini-C(T) technology has been demonstrated through a series of researches and round robin activities [5, 6, 7, 8, 9]. The present study evaluated the through-wall fracture toughness distribution of irradiated IAEA reference material (JRQ) by means of combination of MC method and Mini-C(T) specimens. Four thickness locations between inner surface to 1/4-T was selected. Those four layers were separately subjected to the Mini-C(T) MC evaluation in two different laboratories. Both laboratories could separately obtain valid and consistent reference temperature, To, from all the tested layers. Inner most layer exhibits 80 °C lower To compared to the 1/4-T location even though the layer has the highest fluence of 5.38 × 1019 n/cm2, while that in 1/4-T location is 2.54 × 1019 n/cm2. The results demonstrate that initial toughness distribution is dominant in the general trend of fracture toughness distribution even after the material was highly irradiated.
{"title":"Evaluation of Through Wall Fracture Toughness Distribution of IAEA Reference Material JRQ by Mini-C(T) Specimens and the Master Curve Method","authors":"Masato Yamamoto, Tomohiro Kobayashi","doi":"10.1115/PVP2018-84889","DOIUrl":"https://doi.org/10.1115/PVP2018-84889","url":null,"abstract":"The load and temperature history during pressurized thermal shock (PTS) event is highly depending on the crack edge location in wall thickness direction of a reactor pressure vessel (RPV) beltline region. Therefore, the consideration of plant specific through-wall fracture toughness distribution, which is not considered in the current codes and regulations [1,2], may improve the structural integrity assessment for PTS event.\u0000 The Master Curve (MC) method [3,4] is one of the methods, which can directory evaluate the fracture toughness of ferritic materials with relatively low number of any size of specimens. CRIEPI has proposed the use of very small C(T) (Mini-C(T)) specimens for the MC method. The appropriateness of Mini-C(T) technology has been demonstrated through a series of researches and round robin activities [5, 6, 7, 8, 9].\u0000 The present study evaluated the through-wall fracture toughness distribution of irradiated IAEA reference material (JRQ) by means of combination of MC method and Mini-C(T) specimens. Four thickness locations between inner surface to 1/4-T was selected. Those four layers were separately subjected to the Mini-C(T) MC evaluation in two different laboratories. Both laboratories could separately obtain valid and consistent reference temperature, To, from all the tested layers. Inner most layer exhibits 80 °C lower To compared to the 1/4-T location even though the layer has the highest fluence of 5.38 × 1019 n/cm2, while that in 1/4-T location is 2.54 × 1019 n/cm2. The results demonstrate that initial toughness distribution is dominant in the general trend of fracture toughness distribution even after the material was highly irradiated.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"77 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124957415","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}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：