A Proposal for the Maximum KIc for Use in ASME Code Flaw and Fracture Toughness Evaluations

2011 ◽  
Author(s):  
Mark T. Kirk ◽  
Gary L. Stevens ◽  
Marjorie Erickson ◽  
Shengjun Sean Yin
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Empirical Models ◽  
Ferritic Steels ◽  
Transition Regime ◽  
Upper Limit ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
The One ◽  
Reactor Pressure

Nonmandatory Appendices A [1] and G [2] of Section XI of the ASME Code use the KIc curve (indexed to the material reference transition temperature, RTNDT) in reactor pressure vessel (RPV) flaw evaluations, and for the purpose of establishing RPV pressure-temperature (P-T) limits. Neither of these appendices places an upper-limit on the KIc value that may be used in these assessments. Over the years, it has often been suggested by some of the members of the ASME Section XI Code committees that are responsible for maintaining Appendices A and G that there is a practical upper limit of 200 ksi√in (220 MPa√m) [4]. This upper limit is not well recognized by all users of the ASME Code, is not explicitly documented within the Code itself, and the one source known to the authors where it is defended [4] relies on data that is either in error, or is less than 220 MPa√m. However, as part of the NRC/industry pressurized thermal shock (PTS) reevaluation effort, empirical models were developed that propose common temperature dependencies for all ferritic steels operating on the upper shelf. These models relate the fracture toughness properties in the transition regime to those on the upper shelf and, combined with data for a wide variety of RPV steels and welds on which they are based, suggest that the practical upper limit of 220 MPa√m exceeds the upper shelf fracture toughness of most RPV steels by a considerable amount, especially for irradiated steels. In this paper, available models and data are used to propose upper bound limits of applicability on the KIc curve for use in ASME Code, Section XI, Nonmandatory Appendices A and G evaluations that are consistent with available data for RPV steels.

To Determination of the WWER RPV Steels Crack Resistance Characteristics

2013 ◽  
Author(s):  
Igor Orynyak ◽  
Maksym Zarazovskii ◽  
Sergii Radchenko ◽  
Volodymyr Kozlov
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Crack Resistance ◽  
Temperature Shift ◽  
Charpy Impact ◽  
Temperature Range ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure ◽  
15Kh2nmfa Steel

The efficiency of fracture toughness determined by the methodology of normative document PNAE G-7-002-86 has been analyzed. Crack resistance characteristics of WWER-1000 reactor pressure vessel base metal at unirradiated condition are obtained by experimental way. All specimens were made of the RPV support forging (15Kh2NMFA steel) of abandoned Crimean NPP Fracture toughness experiments were carried out on three types of specimens CT 1T, CT 0.5T and SEB over a temperature range from −130°C to −40°C in fully accordance to the ASTM E1921. Charpy impact energy data obtained on twelve specimens over a temperature range from −80°C to 80°C has been used to determine the 47J transition temperature. Comparison of obtained fracture toughness data with normative curve shows that the last one has unreasonably high lower shelf. It has been found that the PNAE G-7-002-86 Code, which uses the ideology of transition temperature shift, is too conservative to estimate WWER-1000 RPVs resistance against brittle fracture for the pressurized thermal shock (over 90 MPa·√m area of stress intensity factor).

In-Service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution Per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

2011 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Michael T. Anderson ◽  
Wallace Norris
Keyword(s):  
Fracture Toughness ◽  
Density Distribution ◽  
Thermal Shock ◽  
Size Distribution ◽  
Pressure Vessel ◽  
Ultrasonic Testing ◽  
Reactor Vessel ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Reactor Pressure

The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking of a reactor pressure vessel (RPV) due to a pressurized thermal shock (PTS) event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.” The §50.61a rule, which is optional, requires licensees to analyze the results from periodic volumetric examinations required by the American Society of Mechanical Engineers (ASME) Code. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory has been working on a program to assess the ability of current inservice inspection ultrasonic testing (UT) techniques, as qualified through the ASME Code to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, “Spirit of Appendix VIII reactor vessel examination,” a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

Assessment of Fracture Toughness Models for Ferritic Steels Used in Section XI of the ASME Code Relative to Current Data-Based Models

2014 ◽  
Author(s):  
Mark Kirk ◽  
Marjorie Erickson ◽  
William Server ◽  
Gary Stevens ◽  
Russell Cipolla
Keyword(s):  
Fracture Toughness ◽  
Current Data ◽  
Ferritic Steels ◽  
Irradiation Effects ◽  
Irradiation Embrittlement ◽  
New Information ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
The Mean ◽  
The U.S

Section XI of the ASME Code provides models of the fracture toughness of ferritic steel. Recent efforts have been made to incorporate new information, such as the Code Cases that use the Master Curve, but the fracture toughness models in Section XI have, for the most part, remained unchanged since the KIc and KIa curves were first developed in Welding Research Council Bulletin 175 in 1972. Since 1972, considerable advancements to the state of knowledge, both theoretical and practical have occurred, particularly with regard to the amount of available data. For example, as part of the U.S. Nuclear Regulatory Commission’s (NRC’s) pressurized thermal shock (PTS) re-evaluation efforts the NRC and the industry jointly developed an integrated model that predicts the mean trends and scatter of the fracture toughness of ferritic steels throughout the temperature range from the lower shelf to the upper shelf. This collection of models was used by the NRC to establish the index temperature screening limits adopted in the Alternate PTS Rule documented in Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50.61a (10CFR50.61a). In this paper the predictions of the toughness models used by the ASME Code are compared with these newer models (that are based on considerably more data) to identify areas where the ASME Code could be improved. Such improvements include the following: • On the lower shelf, the low-temperature asymptote of the KIc curve does not represent a lower bound to all available data. • On the upper shelf, the de facto KIc limit of applicability of 220 MPa√m exceeds available data, especially after consideration of irradiation effects. • The separation between the KIc and KIa curves depends on the amount of irradiation embrittlement, a functionality not captured by the ASME Section XI equations. • The temperature above which upper shelf behavior can be expected depends on the amount of irradiation embrittlement, a functionality not captured in the ASME Section XI equations.

Applications of the Master Curve Approach to Irradiated Steels

2004 ◽  
Author(s):  
A. Ballesteros ◽  
J. Bros ◽  
M. Brumovsky
Keyword(s):  
Fracture Toughness ◽  
Master Curve ◽  
Pressure Vessels ◽  
Current Situation ◽  
Ferritic Steels ◽  
Preliminary Results ◽  
Fundamental Understanding ◽  
Asme Code ◽  
Reactor Pressure ◽  
Open Issues

The Master Curve approach is being introduced in the evaluation of the fracture toughness of low-alloy ferritic steels used in reactor pressure vessels. The Master Curve provides significant technical enhancements compared with the methodology currently incorporated in the ASME code. The fundamental understanding of Master Curve has been achieved, but application issues are emerging facing the integrity evaluation of reactor pressure vessels. The current situation of the Master Curve methodology is reviewed in this paper, including the last developments for WWER reactors. The open issues are discussed, and the preliminary results from investigations being carried out in Spain are presented.

Alternative RTNDT for Linde 80 Weld Materials

2002 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Screening Criteria ◽  
Master Curve Method ◽  
Asme Code ◽  
Curve Method

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.

Study on Flaw Acceptance Standard of ASME Code Sec. XI Based on Failure Probability

2004 ◽  
Author(s):  
Katsuyuki Shibata ◽  
Kunio Onizawa ◽  
YinSheng Li ◽  
Yasuhiro Kanto ◽  
Shinobu Yoshimura
Keyword(s):  
Case Studies ◽  
Failure Probability ◽  
Pressure Vessels ◽  
Aspect Ratios ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Order Of Magnitude ◽  
The Difference ◽  
Reactor Pressure

Based on the failure probability, the flaw acceptance standard of ASME Code Sec. XI is examined with some concerns weather the failure probability is uniform for flaws with various aspect ratios and failure frequencies are small enough. In this paper, the results of preliminary case studies are described on the failure probability of reactor pressure vessels (RPVs) with a surface flaw specified in Sec. XI. PFM code PASCAL was used for case studies. A PTS (Pressurized Thermal Shock) transient prescribed by NRC/EPRI PTS Benchmark Study was used as an applied load. Analysis results showed that the conditional failure probability of a RPV with an initial flaw of acceptable depth depends on the aspect ratio. In the case flaw shapes are close to semi-circular, the failure probability are higher than that of the cases aspect ration are less than 0.6 by one order of magnitude due to the difference of fracture behavior at the surface point. A case study for determining the acceptable flaws based on failure probability was also carried out.

Effects of Cladding on Reactor Pressure Vessel Integrity Based on the Revised PTS Rule (10 CFR 50.61)

2009 ◽  
Author(s):  
Vikram Marthandam ◽  
Timothy J. Griesbach ◽  
Jack Spanner
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Historical Perspective ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Potential Benefits ◽  
Vessel Integrity ◽  
Technical Basis ◽  
Reactor Pressure

This paper provides a historical perspective of the effects of cladding and the analyses techniques used to evaluate the integrity of an RPV subjected to pressurized thermal shock (PTS) transients. A summary of the specific requirements of the draft revised PTS rule (10 CFR 50.61) and the role of cladding in the evaluation of the RPV integrity under the revised PTS Rule are discussed in detail. The technical basis for the revision of the PTS Rule is based on two main criteria: (1) NDE requirements and (2) Calculation of RTMAX-X and ΔT30. NDE requirements of the Rule include performing volumetric inspections using procedures, equipment and personnel qualified under ASME Section XI, Appendix VIII. The flaw density limits specified in the new Rule are more restrictive than those stipulated by Section XI of the ASME Code. The licensee is required to demonstrate by performing analysis based on the flaw size and density inputs that the through wall cracking frequency does not exceed 1E−6 per reactor year. Based on the understanding of the requirements of the revised PTS Rule, there may be an increase in the effort needed by the utility to meet these requirements. The potential benefits of the Rule for extending vessel life may be very large, but there are also some risks in using the Rule if flaws are detected in or near the cladding. This paper summarizes the potential impacts on operating plants that choose to request relief from existing PTS Rules by implementing the new PTS Rule.

Evaluation of P-T Limit Curves According to Alternative Fracture Toughness Requirements

2014 ◽  
Vol 1051 ◽  
pp. 896-901
Author(s):  
Sin Ae Lee ◽  
Sung Jun Lee ◽  
Sang Hwan Lee ◽  
Yoon Suk Chang
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Ductile Failure ◽  
Evaluation Procedure ◽  
Limit Curve ◽  
Asme Code ◽  
Temperature And Pressure ◽  
Reactor Pressure

During the heat-up and cool-down processes of nuclear power plants, temperature and pressure histories are to be maintained below the P-T limit curve to prevent the non-ductile failure of the RPV(Reactor Pressure Vessel). The ASME Code Sec. XI, App. G describe the detailed procedure for generating the P-T limit curve. The evaluation procedure is containing the evaluation methods of RTNDT using 10CFR50.61. However, recently, Alternative fracture toughness requirements were released 10CFR50.61a. Therefore, in this study, RTNDT of RPV according to the 10CFR50.61a was calculated and used for evaluation of P-T limit curve of a typical RPV under cool-down condition. As a result, it was proven that the P-T curve obtained from 10CFR50.61 is conservative because RTNDT value obtained from the alternative fracture toughness requirements are significantly low.

Influence Factors of Fracture Mechanics Analysis of Reactor Pressure Vessel under Pressurized Thermal Shock

2019 ◽  
Vol 795 ◽  
pp. 333-339
Author(s):  
Juan Luo ◽  
Jia Cheng Luo
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Influence Factors ◽  
Pressurized Thermal Shock ◽  
Inner Surface ◽  
Mechanics Analysis ◽  
Reactor Pressure

When the reactor pressure vessel (RPV) is subjected to pressurized thermal shock (PTS), the cooling water injected by the emergency core cooling system (ECCS) will generate a large temperature difference in the wall thickness of the pressure vessel. On the other hand, the fracture toughness of the RPV material decreases a lot under long-term neutron irradiation. Under this condition, the PTS transient may cause a rapid growth of defects in the inner surface of the vessel, resulting in failure of the pressure vessel. In this paper, the fracture mechanics analysis method of RPV under pressurized thermal shock is studied. The thermal analysis and structural analysis of the pressure vessel are performed by finite element method. The stress intensity factor and fracture toughness are obtained through calculation. At the same time, the influence factors of fracture mechanics analysis of RPV under PTS condition are analyzed. The effects of different crack size, crack type, load transient, and neutron irradiation flux on the PTS fracture mechanics analysis results are evaluated. Results show that the larger the ratio of length to depth for axial inner surface cracks, the easier RPV crack grows. Under small break condition, the circumferential cracks are safer than axial cracks. The longer the operating time, the more severe the embrittlement of RPV materials, which will lead to the failure of RPV more easily. For the two typical PTS transients studied in this paper, the re-pressurization condition is safer than the small break condition. The results can provide basis for structural integrity assessment of RPV under PTS condition.

Fracture Toughness in Postulated Crack Area of PTS Evaluation in Highly-Neutron Irradiated RPV Steel

2021 ◽  
Author(s):  
Yoosung Ha ◽  
Masaki Shimodaira ◽  
Hisashi Takamizawa ◽  
Tohru Tobita ◽  
Jinya Katsuyama ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Vickers Hardness ◽  
Heat Affected Zone ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Rpv Steel ◽  
Pressurized Thermal Shock ◽  
Inner Surface ◽  
Surveillance Specimens ◽  
Reactor Pressure

Abstract The Japanese Electric Association Code 4206-2016 requires that the semi-elliptical crack sized 10 mm in depth × 60 mm in length shall be postulated near the inner surface of a reactor pressure vessel (RPV) in pressurized thermal shock events. The fracture toughness distribution was investigated in the postulated crack area under the PTS events of unirradiated and highly-neutron irradiated RPV steels. Vickers hardness in heat-affected zone (HAZ) due to stainless overlay cladding and 10 mm from the cladding were higher than that of a quarter thickness position, where the surveillance specimens are machined, for both unirradiated (E1) and irradiated (up to 1 × 1020 n/cm2, WIM) materials. Fracture toughness of HAZ and 10 mm from the cladding was higher for the above highly-neutron irradiated material. The same result was obtained in the unirradiated material. Therefore, it was confirmed that fracture toughness obtained from surveillance specimens can provide conservative assessment of structural integrity of RPV.

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