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

Asme Code ◽

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.

Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

10.2172/1136240 ◽

2014 ◽

Author(s):

Edmund Sullivan ◽

Michael Anderson

Keyword(s):

Fracture Toughness ◽

Density Distribution ◽

Thermal Shock ◽

Pressure Vessel ◽

Ultrasonic Testing ◽

Reactor Pressure Vessel ◽

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

Volume 7: Operations, Applications and Components ◽

10.1115/pvp2009-78088 ◽

2009 ◽

Author(s):

Vikram Marthandam ◽

Timothy J. Griesbach ◽

Jack Spanner

Keyword(s):

Thermal Shock ◽

Pressure Vessel ◽

Historical Perspective ◽

Asme Code ◽

Potential Benefits ◽

Vessel Integrity ◽

Technical Basis ◽

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.

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

Key Engineering Materials ◽

10.4028/www.scientific.net/kem.795.333 ◽

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 ◽

Inner Surface ◽

Mechanics Analysis ◽

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.

A Risk-Informed Assessment of ASME Section XI, Appendix E

ASME 2011 Pressure Vessels and Piping Conference: Volume 6, Parts A and B ◽

10.1115/pvp2011-57201 ◽

2011 ◽

Author(s):

Ronald Gamble ◽

William Server ◽

Bruce Bishop ◽

Nathan Palm ◽

Carol Heinecke

Keyword(s):

Thermal Shock ◽

Pressure Vessel ◽

Evaluation Criteria ◽

Nuclear Regulatory Commission ◽

Asme Code ◽

American Society ◽

Regulatory Commission ◽

Reactor Pressure ◽

The U.S

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Non mandatory Appendix E, “Evaluation of Unanticipated Operating Events”, provides a deterministic procedure for evaluating reactor pressure vessel (RPV) integrity following an unanticipated event that exceeds the operational limits defined in plant operating procedures. The recently developed risk-informed procedure for Appendix G to Section XI of the ASME Code [2, 3], and the development by the U.S. Nuclear Regulatory Commission (NRC) of the alternate Pressurized Thermal Shock (PTS) rule [4, 5, 6] led to initiation of this study to determine if the Appendix E evaluation criteria are consistent with risk-informed acceptance criteria. The results of the work presented in this paper demonstrate that Appendix E is consistent with risk-informed criteria developed for PTS and Appendix G and ensures that evaluation of RPV integrity following an unanticipated event would not violate material or operational limits recently defined using risk-informed criteria. Currently, Appendix E does not have evaluation criteria for BWR vessels; however, as part of this study, risk-informed analyses were performed for unanticipated heat-up events and isothermal, overpressure events in BWR plant designs.

Weibull statistical models of KIc/KIa fracture toughness databases for pressure vessel steels with an application to pressurized thermal shock assessments of nuclear reactor pressure vessels

International Journal of Pressure Vessels and Piping ◽

10.1016/s0308-0161(01)00031-x ◽

2001 ◽

Vol 78 (2-3) ◽

pp. 165-178 ◽

Cited By ~ 9

Author(s):

P.T. Williams ◽

K.O. Bowman ◽

B.R. Bass ◽

T.L. Dickson

Keyword(s):

Fracture Toughness ◽

Thermal Shock ◽

Pressure Vessel ◽

Statistical Models ◽

Nuclear Reactor ◽

Pressure Vessels ◽

Pressure Vessel Steels ◽

Evaluation on Constraint Effect of Reactor Pressure Vessel Under Pressurized Thermal Shock

Volume 1B: Codes and Standards ◽

10.1115/pvp2013-97566 ◽

2013 ◽

Cited By ~ 1

Author(s):

Naoki Ogawa ◽

Kentaro Yoshimoto ◽

Takatoshi Hirota ◽

Shohei Sakaguchi ◽

Toru Oumaya

Keyword(s):

Fracture Toughness ◽

Thermal Shock ◽

Surface Crack ◽

Pressure Vessel ◽

Crack Tip ◽

Reactor Pressure Vessel ◽

Constraint Effect ◽

Elastic Plastic ◽

In recent years, the integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) accident has become controversial issue since the larger shift of RTNDT in some higher fluence surveillance data raised a concern on RPV integrity. Under PTS condition, the combination of thermal stress due to a temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the wall of RPV. Currently, RPV integrity is assessed by comparing stress intensity factor on a crack tip under PTS condition and a reference toughness curve based on the fracture toughness data of irradiated compact specimens. Since PTS loading is large enough to cause plastic deformation, a crack tip behavior on the inner surface of RPV can be explained by elastic-plastic fracture mechanics using the J-integral. In this study, 3D elastic plastic finite element analyses were performed to assess the crack tip behavior on surface of a RPV under Loss of coolant Accident, which causes one of the most severe PTS condition. In order to quantify the constraint effect on a surface crack, J-Q approach was applied. The constraint effect of a surface crack was compared with a compact specimen and its influence on the fracture toughness was assessed. As a result, the difference of constraint effect was clearly obtained. And it is recommended to consider constraint effects in the evaluation of structural integrity of RPV under PTS.