Understanding Equivalent Margins Analysis for Required Upper Shelf Energy

2012 ◽  
Author(s):  
Michael Benson ◽  
Gary L. Stevens ◽  
Mark Kirk ◽  
Russell Cipolla ◽  
Douglas Scarth
Keyword(s):  
Pressure Vessel ◽  
Energy Use ◽  
Shelf Energy ◽  
Robust Implementation ◽  
Growth Stability ◽  
Ductile Crack Growth ◽  
Asme Code ◽  
American Society ◽  
Crack Growth Stability ◽  
Reactor Pressure

Equivalent Margins Analysis (EMA) involves the calculation of an alternative minimum reactor pressure vessel (RPV) upper shelf energy (USE) when the projected value falls below current limits codified in Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), Appendix G. One set of calculation methodologies for performing the analysis are provided in the Nuclear Regulatory Commission’s (NRC’s) Regulatory Guide (RG) 1.161 and American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME Code), Section XI, nonmandatory Appendix K. Careful application of fracture mechanics principles is necessary in order to properly carry out the evaluation. This is particularly the case for demonstrating compliance with the ductile crack growth stability criterion. This paper discusses robust implementation of EMA calculations and identifies recommended changes to RG 1.161 and ASME Code, Section XI, nonmandatory Appendix K.

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

2011 ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Evaluation Criteria ◽  
Nuclear Regulatory Commission ◽  
Pressurized Thermal Shock ◽  
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.

Study on Applicability of Master Curve Methodology Using Miniature C(T) Specimen to a Reactor Pressure Vessel With Low Upper Shelf Energy

2018 ◽  
Author(s):  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto ◽  
Masato Oshikiri ◽  
Kazuya Tsutsumi ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Master Curve ◽  
Energy Use ◽  
Reactor Pressure Vessel ◽  
Test Method ◽  
Shelf Energy ◽  
Irradiation Embrittlement ◽  
Reactor Pressure

Miniature compact tension (Mini-C(T)) specimen can be an effective tool by utilizing together with Master Curve (MC) methodology for fracture toughness evaluation of irradiated reactor pressure vessel (RPV) steels. Recently, Mini-C(T) specimen has been incorporated into the Japanese standard test method related to MC methodology, JEAC4216-2015 and several studies were found focusing on applicability of Mini-C(T) specimen to irradiated RPV materials. However, there exist some other issues to be resolved considering application to irradiated materials. One of them is violation against the limitation criteria for ductile crack growth (DCG) specified in the standards. In general, upper shelf energy (USE) of RPV materials tends to decrease as well as shift in Charpy transition temperature due to neutron irradiation embrittlement. It may cause reduction in resistance of material against DCG and this leads to the problem peculiar to low USE materials such that the limitation for DCG might be dominant rather than that for KJclimit. Therefore, it is possible to fail to obtain valid KJc data even within valid temperature range of MC methodology, i.e. −50°C ≤ T-To ≤ 50°C, for low USE materials using Mini-C(T) specimens. In this study, the RPV steel with USE lower than 68J was made simulating reduction of USE due to neutron irradiation. Fracture toughness tests were performed using Mini-C(T) specimens as well as the standard 1T-C(T) specimens. Based on the test results, the validity for DCG limitation was also evaluated for each datum by post-test observation of fracture surface. Additionally, effectiveness of added side grooves and double thickness of specimen was examined as a countermeasure for Mini-C(T) specimen.

Fracture Toughness of Highly Irradiated Pressure Vessel Steels in the Upper Shelf Temperature

2006 ◽  
Author(s):  
Hiroshi Matsuzawa ◽  
Toru Osaki
Keyword(s):  
Fracture Toughness ◽  
Chemical Composition ◽  
Reference Material ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Charpy Impact ◽  
Pressure Vessels ◽  
Shelf Energy ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Nine Reactor Pressure Vessel (RPV) Steels and four RPV weld were irradiated up to 1.2 × 1024n/m2 fast neutron fluence (E>1MeV), and their fracture toughness and Charpy impact energy were measured. As chemical compositions, such as Cu, are known to affect the fracture toughness reduction due to neutron exposure, the above steels were fabricated by changing chemical composition widely to cover the chemical composition of the RPV materials of the operating Japanese nuclear power plants. 2.7 mm thick compact specimens were used to measure the upper shelf fracture toughness of highly irradiated materials, and their Charpy upper shelf energy was also measured. By correlating Charpy upper shelf energy to fracture toughness, the upper shelf fracture toughness evaluation formulae for highly irradiated reactor pressure vessel steels were developed. Both compact and V-notched Charpy impact specimens were irradiated in a test reactor. The fast neutron flux above 1MeV was about 5 × 1016n/(m2s). Charpy impact specimens made of Japanese PWR reference material containing 0.09w% Cu were irradiated simultaneously. The upper shelf energy of the reference material up to the medium fluence level showed little difference in the reduction of upper shelf energy to that which had been in the operating plant and which was irradiated to the same fluence. The developed correlation formulae have been adopted in the Japan Electric Association Code as new formulae to predict the fracture toughness in the upper shelf region of reactor pressure vessels. They will be applied to time limited ageing analysis of low upper shelf reactor pressure vessels in Japan, on a concrete technical basis in very high fluence regions.

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.

Accounting for the Effects of Inservice Inspection Flaw Depth Sizing Errors in Developing Flaw Size Distribution Tables

2009 ◽  
Author(s):  
S. R. Gosselin ◽  
F. A. Simonen
Keyword(s):  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Expert Elicitation ◽  
Flaw Size ◽  
Physical Models ◽  
Asme Code ◽  
Valid Comparison ◽  
Vessel Material ◽  
Reactor Pressure

Probabilistic fracture mechanics studies have addressed reactor pressure vessels that have high levels of material embrittlement. These calculations have used flaw size and density distributions determined from precise and optimized laboratory measurements made and validated with destructive methods as well as from physical models and expert elicitation. The experimental data were obtained from reactor vessel material samples removed from cancelled plants (Shoreham and the Pressure Vessel Research Users Facility (PVRUF)). Consequently, utilities may need to compare the numbers and sizes of reactor pressure vessel flaws identified by the plant’s inservice inspection program to the numbers and sizes of flaws assumed in prior failure probability calculations. This paper describes a method to determine whether the flaws in a particular reactor pressure vessel are consistent with the assumptions regarding the number and sizes of flaws used in other analyses. The approach recognizes that ASME Code Section XI examinations suffer from limitations in terms of sizing errors for very small flaws. Direct comparisons of a vessel specific flaw distribution with other documented flaw distributions would lead to pessimistic conclusions. This paper provides a method for a valid comparison that accounts for flaw sizing errors present in ASME Code Section XI examinations.

A Study of Axial Compression in the ASME B&PV Code for Pipe Supports and Restraints

2016 ◽  
Author(s):  
Phillip E. Wiseman ◽  
Zara Z. Hoch
Keyword(s):  
Axial Compression ◽  
Pressure Vessel ◽  
Slenderness Ratio ◽  
Elastic Buckling ◽  
Allowable Stress ◽  
Linear Elastic ◽  
Division 1 ◽  
Asme Code ◽  
American Society ◽  
Allowable Stress Design

Axial compression allowable stress for pipe supports and restraints based on linear elastic analysis is detailed in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NF. The axial compression design by analysis equations within NF-3300 are replicated from the American Institute of Steel Construction (AISC) using the Allowable Stress Design (ASD) Method which were first published in the ASME Code in 1973. Although the ASME Boiler and Pressure Vessel Code is an international code, these equations are not familiar to many users outside the American Industry. For those unfamiliar with the allowable stress equations, the equations do not simply address the elastic buckling of a support or restraint which may occur when the slenderness ratio of the pipe support or restraint is relatively large, however, the allowable stress equations address each aspect of stability which encompasses the phenomena of elastic buckling and yielding of a pipe support or restraint. As a result, discussion of the axial compression allowable stresses provides much insight of how the equations have evolved over the last forty years and how they could be refined.

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.

Radiation Effects in Boiling Water Reactor Pressure Vessel Steels

1980 ◽  
Vol 102 (2) ◽  
pp. 177-186
Author(s):  
J. N. Kass ◽  
A. J. Giannuzzi ◽  
D. A. Hughes
Keyword(s):  
Pressure Vessel ◽  
Radiation Effects ◽  
Statistical Evaluation ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Shelf Energy ◽  
Water Reactor ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Effect of neutron irradiation on notch toughness properties of Boiling Water Reactor pressure vessel steels was determined. Samples from several heats of plate, weld metal, and forgings were irradiated to three different fluence levels and tested. A statistical evaluation of the data was conducted to determine regression analysis mean decreases in upper shelf energy and increases in transition temperature versus fluence.

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