Overview of the Doel 3 and Tihange 2 Reactor Pressure Vessel Safety Cases

2016 ◽  
Author(s):  
Michel De Smet ◽  
Jean Van Vyve
Keyword(s):  
Structural Integrity ◽  
Case Reports ◽  
Operating Mode ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Safety Case ◽  
On Line ◽  
Safety Cases ◽  
Integrity Analysis ◽  
Reactor Pressure

During the summer outages of 2012, large numbers of indications were found inside the shell material of the Doel 3 and Tihange 2 reactor pressure vessels (RPV). Therefore both plants remained in cold shutdown with their core unloaded. A series of examinations, tests and inspections were performed leading to the conclusion that the indications are hydrogen flakes and that they do not affect the structural integrity of the RPVs, regardless the operating mode, transient or accident condition. All this was documented in the Safety Case Reports issued in December 2012 and the Safety Case Addenda issued in April 2013. Based on those reports, the Belgian Federal Agency for Nuclear Control (FANC) authorized the restart of both units that went back on-line in June 2013. In parallel, a number of requirements from the FANC were addressed such as the qualification of the applied Ultrasonic Testing (UT) procedure, and mechanical testing of irradiated specimens containing hydrogen flakes. The preliminary tests showed unexpected results regarding the shift in RTNDT (Reference Temperature for Nil Ductility Transition) under irradiation that could not confirm the hypotheses considered in the initial Safety Case Reports. Therefore, the Licensee Electrabel decided to shut down both plants immediately. In order to fully address this concern, the material test programme was extended including several RPV materials and covering additional irradiation campaigns. This led to a modification of the irradiation embrittlement trend curves considered in the structural integrity analysis. In addition, the qualification of the UT procedure led to an updated cartography of the flakes. The structural integrity assessments of both RPVs were revised accordingly. The final Safety Case Reports, confirming the fitness-for-continued operation of both RPVs, were submitted to the FANC in October 2015. FANC allowed restart of both units on November 17th, 2015. The paper gives a historical overview of the Doel 3 and Tihange 2 RPV Safety Cases and explains how the roadmap was built in order to demonstrate the RPV’s structural integrity in the presence of hydrogen flakes.

Material Properties of Reactor Pressure Vessel Shells Affected by Hydrogen Flaking

2016 ◽  
Author(s):  
Robert Gérard ◽  
Michel De Smet ◽  
Rachid Chaouadi
Keyword(s):  
Fracture Toughness ◽  
Material Properties ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Operating Cycle ◽  
Safety Case ◽  
Safety Cases ◽  
Definition Of ◽  
Reactor Pressure ◽  
Trend Curve

During the summer outages of 2012, large numbers of nearly-laminar indications were found in the core shells of the Doel 3 and Tihange 2 reactor pressure vessels (RPV). As a consequence, both units remained in cold shutdown with their core unloaded. A series of examinations, tests and inspections were performed leading to the conclusion that the indications are hydrogen flakes and that they do not affect the structural integrity of the RPV, regardless of the operating mode, transient or accident condition. All this was documented in the Safety Case reports issued in December 2012 and in the Safety Case Addenda issued in April 2013 [1]. Based on those reports, the Belgian Federal Agency for Nuclear Control (FANC) authorized the restart of both units which went back on-line in June 2013. A key input required for this Safety Case was the definition of the appropriate material properties, in particular fracture toughness, for the RPV shells affected by hydrogen flakes. A material testing program on non-irradiated materials evaluated aspects like the possible effects of macro-segregations and local segregations (ghost lines) and of specimen orientation on the fracture toughness. The irradiation embrittlement sensitivity of the zone of macro-segregation in which the flakes are located was evaluated on the basis of the maximum enrichment in Cu, P and Ni in macro-segregations based on literature data. This was the basis of the trend curve of RTNDT evolution vs. fluence used in the Safety Cases submitted in 2012–2013. The restart authorization in 2013 was accompanied by a number of “mid-term” requirements, to be completed during the first operating cycle after the restart. One of these requirements was the mechanical testing of irradiated specimens containing hydrogen flakes, in order to confirm the conservativeness of the RTNDT trend curve used for the structural integrity analyses. After a first irradiation campaign of a material containing hydrogen flakes in the BR2 reactor of the Belgian Nuclear Research Center SCK.CEN, atypical results were obtained and the utility decided to shut down the units in March 2014. Detailed investigations involving three additional irradiation campaigns in BR2 including other reference materials, among which another material affected by hydrogen flakes, were performed in order to characterize this atypical behaviour and to derive a new conservative RTNDT trend curve. The resulting trend curve was accepted by the FANC and was used in the 2015 Safety Cases [1]. An overview of the Doel 3 and Tihange 2 safety cases is given in [6]. The paper summarizes the results of the material investigations on non-irradiated and irradiated materials and the process leading to the definition of this conservative RTNDT trend curve.

Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel

2021 ◽  
Vol 143 (4) ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PFM analysis of structural components in aging light water reactor (PASCAL) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against nonductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities, and institutes voluntarily participating as members. Through one-year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches, and results.

ORNL Evaluation of Safety Cases for Two Belgian Reactor Pressure Vessels Containing Quasi-Laminar Defects

2017 ◽  
Author(s):  
Hilda B. Klasky ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
Sarma B. Gorti ◽  
Randy K. Nanstad ◽  
...  
Keyword(s):  
New York ◽  
Finite Element ◽  
Structural Integrity ◽  
National Laboratory ◽  
Relevant Literature ◽  
Pressure Vessels ◽  
Oak Ridge ◽  
Safety Margins ◽  
Safety Cases ◽  
Reactor Pressure

The Oak Ridge National Laboratory (ORNL) performed a detailed technical review of the 2015 Electrabel (EBL) Safety Cases prepared for the Belgium reactor pressure vessels (RPVs) at Doel 3 and Tihange 2 (D3/T2). The Federal Agency for Nuclear Control (FANC) in Belgium commissioned ORNL to provide a thorough assessment of the existing safety margins against cracking of the RPVs due to the presence of almost laminar flaws found in each RPV. Initial efforts focused on surveying relevant literature that provided necessary background knowledge on the issues related to the quasi-laminar flaws observed in D3/T2 reactors. Next, ORNL proceeded to develop an independent quantitative assessment of the entire flaw population in the two Belgian reactors according to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Appendix G, “Fracture Toughness Criteria for Protection Against Failure,” New York (both 1992 and 2004 versions). That screening assessment of the EBL-characterized flaws in D3/T2 used ORNL tools, methodologies, and the ASME Code Case N-848, “Alternative Characterization Rules for Quasi-Laminar Flaws”. Results and conclusions derived from comparisons of the ORNL flaw acceptance assessments of D3/T2 with those from the 2015 EBL Safety Cases are presented in the paper. The ORNL screening analyses identified fewer flaws than EBL that were not compliant with the ASME Section XI (1992) criterion; the EBL criterion imposed additional conservatisms not included in ASME Section XI. Furthermore, ORNL’s application of the updated ASME Section XI (2004) criterion produced only four non-compliant flaws, all due to design-basis loss-of-coolant loading transients. Among the latter, only one flaw remained non-compliant when analyzed using the warm-prestress (WPS) cleavage fracture model typically applied in USA flaw assessments. ORNL’s independent refined analysis of that flaw (#1660, which was also non-compliant in the EBL screening assessments) rendered it compliant when modeled as a more realistic individual quasi-laminar flaw using a 3-dimensional XFEM (eXtended Finite Element Method) approach available in the ABAQUS© finite element code. Taken as a whole, the ORNL-specific results and conclusions confirmed the structural integrity of Doel 3 and Tihange 2 under all design transients with ample margin in the presence of the 16,196 detected flaws.

Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During PTS Events

2012 ◽  
Author(s):  
Kunio Onizawa ◽  
Koichi Masaki ◽  
Jinya Katsuyama
Keyword(s):  
Crack Initiation ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Conditional Probabilities ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Integrity Analysis ◽  
Reactor Pressure

In order to apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), PFM analysis code has been developed at JAEA. Using the PFM analysis code, PASCAL version 3, the conditional probabilities of crack initiation and fracture for an RPV during pressurized thermal shock events have been analyzed. Sensitivity analyses on some input parameters were performed to clarify the effect on the conditional fracture probability. Comparison between the conditional probabilities and temperature margin (ΔTm) from current deterministic analysis method were made for some model plant conditions of domestic typical old-type RPVs. From the analyses, a good correlation between ΔTm and the conditional probability of crack initiation was obtained.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs Affected by Hydrogen Flakes: Refined X-FEM Analyses

2016 ◽  
Author(s):  
Pierre Dulieu ◽  
Valéry Lacroix
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Extended Finite Element Method ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Extended Finite Element ◽  
Integrity Assessment ◽  
Safety Case ◽  
Component Manufacturing ◽  
Reactor Pressure

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, specific ultrasonic in-service inspections revealed a large number of quasi-laminar indications in the base metal of the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, a Flaw Acceptability Assessment had to be performed as a part of the Safety Case demonstrating the fitness-for-service of these units. In that framework, detailed analyses using eXtended Finite Element Method were conducted to model the specific character of hydrogen flakes. Their quasi-laminar orientation as well as their high density required setting up 3D multi-flaws model accounting for flaw interaction. These calculations highlighted that even the most penalizing flaw configurations are harmless in terms of structural integrity despite the consideration of higher degradation of irradiated material toughness.

Development of Probabilistic Fracture Mechanics Analysis Code PASCAL Version 4 for Reactor Pressure Vessels

2018 ◽  
Author(s):  
Kai Lu ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Yinsheng Li ◽  
Shumpei Uno
Keyword(s):  
Structural Integrity ◽  
Prediction Method ◽  
Pressure Vessels ◽  
Stress Distributions ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Integrity Assessment ◽  
Failure Frequency ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWRs) for structural integrity assessment of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock transients. By reflecting the latest knowledge and findings, the PASCAL code has been continuously improved. In this paper, the development of PASCAL Version 4 (hereafter, PASCAL4) is described. Several analysis functions incorporated into PASCAL4 for evaluating the failure frequency of RPVs are introduced, for example, the evaluation function of confidence level of failure frequency considering epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions and KI calculation methods considering complicated stress distributions, and the recent Japanese irradiation embrittlement prediction method. Finally, using PASCAL4, a PFM analysis example for a Japanese model RPV is presented.

Development of Crack Evaluation Models for Probabilistic Fracture Mechanics Analyses of Japanese Reactor Pressure Vessels

2018 ◽  
Author(s):  
Kai Lu ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Irradiation Embrittlement ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Evaluation Models ◽  
Reactor Pressure

In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency for structural integrity assessment of reactor pressure vessels (RPVs). The most recent release is PASCAL Version 4 (hereafter, PSACAL4) which can be used to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and pressurized thermal shock events. For the integrity assessment of RPVs, development of crack evaluation models is important. In this study, finite element analyses are performed firstly to verify the stress intensity factor calculations of cracks in PASCAL4. In addition, the applicability of the crack evaluation models in PASCAL4 such as the location of embedded cracks, crack shape and depth of surface cracks, and the increment of crack propagation is investigated. Based on sensitivity analyses of crack evaluation models for Japanese RPVs using PASCAL4, the effects of these evaluation models on failure frequency are clarified. From the analysis results, crack evaluation models recommended to the failure frequency evaluation for a Japanese model RPV are discussed.

Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels

2017 ◽  
Author(s):  
Jinya Katsuyama ◽  
Kazuya Osakabe ◽  
Shumpei Uno ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Neutron Irradiation ◽  
Structural Integrity ◽  
The United States ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Reactor Pressure

In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the standards developed by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the statistical distributions of various influence factors related to ageing due to the long-term operation. At Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, JAEA has developed a guideline for the PFM based structural integrity assessment of RPVs to promote the applicability of PFM in Japan and achieve the objective that an engineer/analyst who familiar with the fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body (general requirements), explanation (guidance), and several supplements. The technical basis for PFM analysis is also provided, and the new information and better fracture mechanics models are included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

Improvements on Evaluation Functions of a Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessels

2019 ◽  
Vol 142 (2) ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Structural Integrity ◽  
Evaluation Model ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Warm Prestressing ◽  
Evaluation Functions ◽  
Reactor Pressure

Abstract In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL was developed for structural integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. By reflecting the latest knowledge and findings, the evaluation functions are continuously improved and have been incorporated into PASCAL4 which is the most recent version of the PASCAL code. In this paper, the improvements made in PASCAL4 are explained in detail, such as the evaluation model of warm prestressing (WPS) effect, evaluation function of confidence levels for PFM analysis results by considering the epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions, and improved methods for KI calculations when considering complicated stress distributions. Moreover, using PASCAL4, PFM analysis examples considering these improvements are presented.

Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels

2020 ◽  
Vol 142 (2) ◽  
Author(s):  
Jinya Katsuyama ◽  
Kazuya Osakabe ◽  
Shumpei Uno ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Neutron Irradiation ◽  
Structural Integrity ◽  
The United States ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Reactor Pressure

Abstract In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the codes provided by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the probabilistic distributions of various influence factors related to aged degradation due to the long-term operation. In Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, we have developed a guideline for structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and enable persons who have knowledge on fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body, explanation, and several supplements. The technical basis for PFM analysis is provided, and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

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