Nuclear Regulatory Commission Staff View of Progress in the Nondestructive Testing of Cast Austenitic Steel Components

2014 ◽  
Author(s):  
Stephen E. Cumblidge
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Degradation Mechanism ◽  
Austenitic Steels ◽  
Inspection Method ◽  
The Past ◽  
Asme Code

Welds in cast austenitic steels (CASS) are very challenging to inspect using the current American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements. Supplement 9 of ASME Boiler and Pressure Vessel Code Section XI, Appendix VIII is still in the course of preparation, requiring inspectors to use ASME Code Section XI, Appendix III, which provides prescriptive ultrasonic testing (UT) requirements that are significantly less rigorous than UT techniques that have been demonstrated under Appendix VIII. The inability of licensees to demonstrate that the welds in CASS components meet ASME Code requirements has been an ongoing area of concern for the NRC staff. The lack of a reliable inspection method for welds in CASS materials has led to hundreds of relief requests over the past four decades. While no degradation mechanism has been found in CASS components to date, there is no guarantee that a new degradation mechanism affecting CASS welds will not emerge as nuclear power plants go beyond forty years of operation. Licenses need qualified procedures and personnel for the inspection of welds in CASS materials in order to put licensees into compliance with ASME Code, meet federal regulations, reduce the number of needed relief requests, and ensure the structural integrity of their welds. Over the past decade there have been significant developments in nondestructive examination (NDE) technology. The use of encoded phased array techniques using low frequency ultrasound has been shown to be able to reliably find flaws greater than 30% through wall in CASS materials with a variety of microstructures. Additionally, an improved understanding of the fracture mechanics of CASS components is being developed that shows the flaw sizes that can be tolerated in CASS components. These advances in NDE techniques and fracture mechanics theory are converging on a path to allow for qualifications of procedures and personnel for the ultrasonic inspections of welds in CASS components. Recent developments in ASME Code includes Code Case N-824, which provides guidance on the examination of CASS materials based on the advances in NDE technology and an improved understanding of the NDE techniques capable of finding flaws in CASS components as well as Code Case N-838 for flaw tolerance evaluations of CASS piping components. Finally, work on ASME Code Section XI Supplement 9 is progressing, with several important issues still to be addressed. The NRC staff sees a clear path forward and is working to ensure that qualified inspections of welds in CASS materials will be possible in the future.

Structural Integrity Analysis of Nuclear Power Plant Pressure Vessel Penetration Nozzle Repaired

2016 ◽  
Vol 853 ◽  
pp. 346-350
Author(s):  
Lin Wei Ma ◽  
Jia Sheng He ◽  
An Qing Shu ◽  
Xiao Tao Zheng ◽  
Yan Wang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Degradation Mechanism ◽  
Design Requirement ◽  
Original Design ◽  
Asme Code ◽  
Integrity Analysis

Primary water stress corrosion cracking (PWSCC) has been observed in CRDM nozzles, BMI nozzles and other penetration nozzles. The industry has used the repair method of replacement of nozzles fabricated of Alloy 690. After the replacement of the nozzle, the structural integrity analysis of new nozzle and welds should be performed to ensure the pressure boundary compliance with the original design requirement. In this paper, the pressurizer top head instrument nozzle of PWR nuclear power plant is evaluated as a typical pressure vessel penetration nozzle. The results showed that the repaired nozzle satisfies the ASME Code design requirement and the crack growth of the postulated flaw in 40 years of the nuclear plant life is acceptable.PWSCC degradation mechanism has been observed in CRDM nozzles, BMI nozzles and other penetration nozzles [1]. In some nuclear power plants built in China earlier, such as DAYABAY nuclear power plant and QINSHAN nuclear power plant, PWSCC degradation mechanism has been found in CRDM nozzle welds which manufactured of Alloy 600 and welded of Alloy 82/182[2]. The repair of the degraded nozzles is the popular choice for the nuclear power plant owners. After the replacement of the nozzle, the structural integrity analysis of new nozzle and welds should be performed to ensure the pressure boundary compliance with the original design requirement. In this paper, the pressurizer top head nozzle of PWR nuclear power plant is evaluated as a typical pressure vessel penetration nozzle. Stress intensities were conservatively determined for pressure and applicable thermal transients and compared to the allowable values of the ASME Code, Section III. Thermal stress of the transients was obtained from 3D finite element model (FEM). Residual stress of J-groove weld was obtained from 2D FEM analysis and used for fracture mechanics analysis. All of the analysis showed that the repaired nozzle satisfies the ASME Code design requirement and the crack growth of the postulated flaw in 40 years of the nuclear plant life is acceptable.

Strategies for Treating Weld Residual Stresses in Probabilistic Fracture Mechanics Codes

2011 ◽  
Author(s):  
Frederick W. Brust ◽  
R. E. Kurth ◽  
D. J. Shim ◽  
David Rudland
Keyword(s):  
Fracture Mechanics ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Degradation Mechanism ◽  
Welding Process ◽  
Corrosion Cracking ◽  
Uncertainty Sources

Risk based treatment of degradation and fracture in nuclear power plants has emerged as an important topic in recent years. One degradation mechanism of concern is stress corrosion cracking. Stress corrosion cracking is strongly driven by the weld residual stresses (WRS) which develop in nozzles and piping from the welding process. The weld residual stresses can have a large uncertainty associated with them. This uncertainty is caused by many sources including material property variations of base and welds metal, weld sequencing, weld repairs, weld process method, and heat inputs. Moreover, often mitigation procedures are used to correct a problem in an existing plant, which also leads to uncertainty in the WRS fields. The WRS fields are often input to probabilistic codes from weld modeling analyses. Thus another source of uncertainty is represented by the accuracy of the predictions compared with a limited set of measurements. Within the framework of a probabilistic degradation and fracture mechanics code these uncertainties must all be accounted for properly. Here we summarize several possibilities for properly accounting for the uncertainty inherent in the WRS fields. Several examples are shown which illustrate ranges where these treatments work well and ranges where improvement is needed. In addition, we propose a new method for consideration. This method consists of including the uncertainty sources within the WRS fields and tabulating them within tables which are then sampled during the probabilistic realization. Several variations of this process are also discussed. Several examples illustrating the procedures are presented.

Proposed Improvements of the ASME B&PV Code and Future Directions of CIWG

2013 ◽  
Author(s):  
Jinquan Yan ◽  
Yinbiao He ◽  
Gang Li ◽  
Hao Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Third Generation ◽  
Future Directions ◽  
The Past ◽  
Development Direction ◽  
Asme Code ◽  
Set Up ◽  
Near Future ◽  
First Time

The ASME B&PV Code, Section III, is being used as the design acceptance criteria in the construction of China’s third generation AP1000 nuclear power plants. This is the first time that the ASME Code was fully accepted in Chinese nuclear power industry. In the past 6 years, a few improvements of the Code were found to be necessary to satisfy the various requirements originated from these new power plant (NPP) constructions. These improvements are originated from a) the stress-strain curves needed in elastic-plastic analysis, b) the environmental fatigue issue, c) the perplexity generated from the examination requirements after hydrostatic test and d) the safe end welding problems. In this paper, the necessities of these proposed improvements on the ASME B&PV code are further explained and discussed case by case. Hopefully, through these efforts, the near future development direction and assignment of the ASME B&PV-III China International Working Group can be set up.

Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

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.

Constraint-Dependent J-R Curves of a Dissimilar Metal Welded Joint for Connecting Pipe-Nozzle of Nuclear Pressure Vessel

2015 ◽  
Vol 137 (2) ◽  
Author(s):  
J. Wang ◽  
G. Z. Wang ◽  
F. Z. Xuan ◽  
S. T. Tu
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Welded Joint ◽  
Pressure Vessels ◽  
Stress Constraint ◽  
Dissimilar Metal ◽  
Local Strength ◽  
T Stress

In this paper, the J-R curves of two cracks (A508 HAZ crack 2 and A508/Alloy52Mb interface crack 3) located at the weakest region in an Alloy52M dissimilar metal welded joint (DMWJ) for connecting pipe-nozzle of nuclear pressure vessel have been measured by using single edge-notched bend (SENB) specimens with different crack depths a/W (different constraint). Based on the modified T-stress constraint parameter τ*, the equations of constraint-dependent J-R curves for the crack 2 and crack 3 were obtained. The predicted J-R curves using different constraint equations derived from the three pairs of crack growth amount all agree with the experimental J-R curves. The results show that the modified T-stress approach for obtaining constraint-dependent J-R curves of homogeneous materials can also be used for the DMWJs with highly heterogeneous mechanical properties (local strength mismatches) in nuclear power plants. The use of the constraint-dependent J-R curves may increase the accuracy of structural integrity design and assessment for the DMWJs of nuclear pressure vessels.

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL

2017 ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments

Probabilistic fracture mechanics (PFM) has been recognized as 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. In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency (JAEA) to evaluate the through-wall cracking frequencies of Japanese 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 non-ductile fracture. On the other hand, unlike deterministic analysis codes, the verification of PFM analysis codes is not easy. A series of activities has been performed to verify the applicability of PASCAL. In this study, as a part of the verification activities, a working group was established in Japan, 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.

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.

Technical Basis for Cases N-629 and N-631 as an Alternative for RTNDT Reference Temperature

2007 ◽  
Author(s):  
J. G. Merkle ◽  
K. K. Yoon ◽  
W. A. VanDerSluys ◽  
W. Server
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Threshold Level ◽  
Nuclear Regulatory Commission ◽  
New Approach ◽  
The Us ◽  
Asme Code ◽  
Regulatory Commission ◽  
Technical Basis

ASME Code Cases N-629/N-631, published in 1999, provided an important new approach to allow material specific, measured fracture toughness curves for ferritic steels in the code applications. This has enabled some of the nuclear power plants whose reactor pressure vessel materials reached a certain threshold level based on overly conservative rules to use an alternative RTNDT to justify continued operation of their plants. These code cases have been approved by the US Nuclear Regulatory Commission and these have been proposed to be codified in Appendix A and Appendix G of the ASME Boiler and Pressure Vessel Code. This paper summarizes the basis of this approach for the record.

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