NESC-VII: Fracture Mechanics Analyses of WPS Experiments on Large-Scale Cruciform Specimen

This paper describes numerical analyses performed to simulate warm pre-stress (WPS) experiments conducted with large-scale cruciform specimens within the Network for Evaluation of Structural Components (NESC-VII) project. NESC-VII is a European cooperative action in support of WPS application in reactor pressure vessel (RPV) integrity assessment. The project aims in evaluation of the influence of WPS when assessing the structural integrity of RPVs. Advanced fracture mechanics models will be developed and performed to validate experiments concerning the effect of different WPS scenarios on RPV components. The Oak Ridge National Laboratory (ORNL), USA contributes to the Work Package-2 (Analyses of WPS experiments) within the NESC-VII network. A series of WPS type experiments on large-scale cruciform specimens have been conducted at CEA Saclay, France, within the framework of NESC VII project. This paper first describes NESC-VII feasibility test analyses conducted at ORNL. Very good agreement was achieved between AREVA NP SAS and ORNL. Further analyses were conducted to evaluate the NESC-VII WPS tests conducted under Load-Cool-Transient-Fracture (LCTF) and Load-Cool-Fracture (LCF) conditions. This objective of this work is to provide a definitive quantification of WPS effects when assessing the structural integrity of reactor pressure vessels. This information will be utilized to further validate, refine, and improve the WPS models that are being used in probabilistic fracture mechanics computer codes now in use by the NRC staff in their effort to develop risk-informed updates to Title 10 of the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G.

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

Volume 1A: Codes and Standards ◽

10.1115/pvp2017-65305 ◽

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 ◽

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.

Design and Analysis of Pressure Vessels and Piping: Implementation of ASME B31, Fatigue, ASME Section VIII, and Buckling Analyses ◽

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

10.1115/pvp2003-2190 ◽

2003 ◽

Cited By ~ 2

Author(s):

Silvia Turato ◽

Vincent Venturini ◽

Eric Meister ◽

B. Richard Bass ◽

Terry L. Dickson ◽

...

Keyword(s):

Fracture Mechanics ◽

Pressure Vessel ◽

Structural Integrity ◽

Safety Concern ◽

Nuclear Regulatory Commission ◽

Reactor Pressure Vessel ◽

Oak Ridge ◽

Integrity Assessment ◽

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

Probabilistic Fracture Mechanics Evaluations That Consider Nozzles in the Extended Beltline Region of Reactor Pressure Vessels

Volume 1B: Codes and Standards ◽

10.1115/pvp2015-45065 ◽

2015 ◽

Author(s):

Gary L. Stevens ◽

Mark T. Kirk ◽

Terry Dickson

Keyword(s):

Fracture Mechanics ◽

National Laboratory ◽

Computer Code ◽

Nuclear Regulatory Commission ◽

Pressure Vessels ◽

Stress Concentrations ◽

Oak Ridge ◽

Regulatory Issue ◽

For many years, ASME Section XI committees have discussed the assessment of nozzle penetrations in various flaw evaluations for reactor pressure vessels (RPVs). As summarized in Reference [1], linear elastic fracture mechanics (LEFM) solutions for nozzle penetrations have been in use since the 1970s. In 2013, one of these solutions was adopted into ASME Code, Section XI, Nonmandatory Appendix G (ASME App. G) [2] for use in developing RPV pressure-temperature (P-T) operating limits. That change to ASME App. G was based on compilation of past work [3] and additional evaluations of fracture driving force [4][5]. To establish the P-T limits on RPV operation, consideration should be given to both the RPV shell material with the highest reference temperature as well as regions of the RPV (e.g., nozzles, flange) that contain structural discontinuities. In October 2014, the U.S. Nuclear Regulatory Commission (NRC) highlighted these requirements in Regulatory Issue Summary (RIS) 2014-11 [6]. Probabilistic fracture mechanics (PFM) analyses performed to support pressurized thermal shock (PTS) evaluations using the Fracture Analysis Vessels Oak Ridge (FAVOR) computer code [7] currently evaluate only the RPV beltline shell regions. These evaluations are based on the assumption that the PFM results are controlled by the higher embrittlement characteristic of the shell region rather than the stress concentration characteristic of the nozzle, which does not experience nearly the embrittlement of the shell due to its greater distance from the core. To evaluate this assumption, the NRC and the Oak Ridge National Laboratory (ORNL) performed PFM analyses to quantify the effect of these stress concentrations on the results of the RPV PFM analyses. This paper summarizes the methods and evaluates the results of these analyses.

An Improved Method for Postulating Fabrication Flaws in Reactor Pressure Vessels for Structural Integrity Evaluation

10th International Conference on Nuclear Engineering, Volume 1 ◽

10.1115/icone10-22646 ◽

2002 ◽

Cited By ~ 2

Author(s):

F. A. Simonen ◽

T. L. Dickson

Keyword(s):

Fracture Mechanics ◽

Structural Integrity ◽

National Laboratory ◽

Pacific Northwest National Laboratory ◽

Nuclear Regulatory Commission ◽

The United States ◽

Pressure Vessels ◽

Inner Surface ◽

This paper presents an improved model for postulating fabrication flaws in reactor pressure vessels (RPVs) and for the treatment of measured flaw data by probabilistic fracture mechanics (PFM) codes that are used for structural integrity evaluations. The model used to develop the current pressurized thermal shock (PTS) regulations conservatively postulated that all fabrication flaws were inner-surface breaking flaws. To reduce conservatisms and uncertainties in flaw-related inputs, the United States Nuclear Regulatory Commission (USNRC) has supported research at Pacific Northwest National Laboratory (PNNL) that has resulted in data on fabrication flaws from non-destructive and destructive examinations of actual RPV material. Statistical distributions have been developed to characterize the number and sizes of flaws in the various material regions of a vessel. The regions include the main seam welds, repair welds, base metal of plates and forgings, and the cladding that is applied to the inner surface of the vessel. Flaws are also characterized as being located within the interior of these regions or along the weld fusion lines that join the regions. Flaws are taken that occur at random locations relative to the embrittled inner region of the vessel. The probabilistic fracture mechanics model associates each of the simulated flaw types with the fracture properties of the region being addressed.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs Linear Elastic and Elastic-Plastic X-FEM Calculations Performed for Large Scale Tests

Volume 1A: Codes and Standards ◽

10.1115/pvp2016-63767 ◽

2016 ◽

Author(s):

Valéry Lacroix ◽

Pierre Dulieu

Keyword(s):

Large Scale ◽

Structural Integrity ◽

Load Capacity ◽

Pressure Vessels ◽

Integrity Assessment ◽

Elastic Plastic ◽

Linear Elastic ◽

Nuclear Control ◽

Large Scale Tests ◽

In the framework of the hydrogen flakes issue concerning the reactor pressure vessels of the two Belgian NPP’s Doel 3 and Tihange 2, the Federal Agency of Nuclear Control required to perform tests on large scale specimens taken from a block representative of the pressure vessels with the double objective of validating the structural integrity approach and of verifying the load capacity of the specimens affected by flakes. The large scale tests were led on many kinds of specimens: 4 points bending specimens, CT specimens and tensile specimens containing hydrogen flakes or flawed with EDM notches. All of these tests have been simulated using extend finite element method (XFEM). The paper describes the linear elastic and elastic-plastic fracture mechanics calculations performed in the frame of these large scale tests using XFEM and presents the comparison between simulations and experiments. A focus is done on the XFEM capabilities to model 3D complex shaped flaws like hydrogen flakes.

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

Volume 1B: Codes and Standards ◽

10.1115/pvp2018-84965 ◽

2018 ◽

Author(s):

Kai Lu ◽

Koichi Masaki ◽

Jinya Katsuyama ◽

Yinsheng Li

Keyword(s):

Fracture Mechanics ◽

Structural Integrity ◽

Pressure Vessels ◽

Sensitivity Analyses ◽

Integrity Assessment ◽

Failure Frequency ◽

Evaluation Models ◽

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

Volume 1B: Codes and Standards ◽

10.1115/pvp2017-65921 ◽

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 ◽

Integrity Assessment ◽

Failure Frequency ◽

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

Journal of Pressure Vessel Technology ◽

10.1115/1.4045512 ◽

2019 ◽

Vol 142 (2) ◽

Author(s):

Kai Lu ◽

Jinya Katsuyama ◽

Yinsheng Li

Keyword(s):

Fracture Mechanics ◽

Structural Integrity ◽

Evaluation Model ◽

Pressure Vessels ◽

Integrity Assessment ◽

Warm Prestressing ◽

Evaluation Functions ◽

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

Journal of Pressure Vessel Technology ◽

10.1115/1.4045874 ◽

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 ◽

Integrity Assessment ◽

Failure Frequency ◽

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.

Application of Probabilistic Fracture Mechanics to Reactor Pressure Vessel Using PASCAL4 Code

Journal of Pressure Vessel Technology ◽

10.1115/1.4048103 ◽

2020 ◽

Vol 143 (2) ◽

Author(s):

Kai Lu ◽

Jinya Katsuyama ◽

Yinsheng Li ◽

Shinobu Yoshimura

Keyword(s):

Fracture Mechanics ◽

Power Plants ◽

Structural Integrity ◽

Pressure Vessels ◽

Pressurized Water ◽

Integrity Assessment ◽

Thermal Transients ◽

Integrity Assessments ◽

Abstract Probabilistic fracture mechanics (PFM) is considered to be a promising methodology in structural integrity assessments of pressure-boundary components in nuclear power plants since it can rationally represent the inherent probabilistic distributions for influence parameters without over-conservativeness. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 which enables the failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor (PWR) was conducted using PASCAL4, and the effects of nondestructive examination (NDE) and neutron flux reduction on failure frequencies of the RPV were quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for probabilistic integrity assessments of embrittled RPVs and can enhance the applicability of PFM methodology.