Optimization for Nuclear Class 1 Opening Reinforcement

2021 ◽  
Author(s):  
Wolf Reinhardt
Keyword(s):  
Limit Analysis ◽  
Optimum Amount ◽  
Limit Pressure ◽  
Class 1 ◽  
Asme Code ◽  
Opening Up

Abstract An important element in evolving Section III of the ASME Code is the re-examination of Code rules to identify the potential for efficiencies. The present paper looks at the rules for reinforcement of openings. The present Code rules require the area of material removed to create the opening (up to the design thickness) to be added around the opening with certain limits on the distance from the opening. Past studies have suggested that the present rules may add more material than needed to maintain the strength of the vessel in some cases. The optimum amount of reinforcement is therefore postulated to be determined by the criterion that the limit pressure should not be reduced excessively relative to the limit pressure of the vessel in the absence of the opening. Limit analysis is performed to derive possible rules on the amount and distribution of around an opening. Options for alternative rules for reinforcement of openings, and restrictions on the present Code rules, are proposed.

Comparison Between ASME Code-Case N-761, NUREG/CR-6909 and Stainless Steel Component Fatigue Test Results

2014 ◽  
Author(s):  
H. T. Harrison ◽  
Robert Gurdal
Keyword(s):  
Stainless Steel ◽  
Fatigue Test ◽  
Fatigue Tests ◽  
Test Series ◽  
Test Results ◽  
Actual Number ◽  
Steel Component ◽  
Class 1 ◽  
Asme Code ◽  
Number Of Cycles

For Class 1 components, the consideration of the environmental effects on fatigue has been suggested to be evaluated through two different methodologies: either NUREG/CR-6909 from March 2007 or ASME-Code Case N-761 from August 2010. The purpose of this technical paper is to compare these two methods. In addition, the equations from Revision 1 of the NUREG/CR-6909 will be evaluated. For these comparisons, two stainless steel component fatigue test series with documented results are considered. These two fatigue test series are completely different from each other (applied cyclic displacements vs. insurge/outsurge types of transients). Therefore, they are producing an appropriate foundation for these comparisons. In general, the severities of the two methods are compared, where the severity is defined as the actual number of cycles from the fatigue tests, including an evaluation of the scatter, divided by the number of design cycles from the two methods. Also, how stable the methods are is being evaluated through the calculation of the coefficient of variation for each method.

2004 Edition of Japanese Fitness-for-Service Code for Nuclear Power Plants

2006 ◽  
Author(s):  
Koichi Kashima ◽  
Tomonori Nomura ◽  
Koji Koyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Industrial Safety ◽  
National Rule ◽  
Class 1 ◽  
Asme Code ◽  
Service Inspection ◽  
Core Shroud

JSME (Japan Society of Mechanical Engineers) published the first edition of a FFS (Fitness-for-Service) Code for nuclear power plants in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection. Individual inspection rules were prescribed for specific structures, such as the Core Shroud and Shroud Support for BWR plants, in consideration of aging degradation by Stress Corrosion Cracking (SCC). Furthermore, JSME established the third edition of the FFS Code in December 2004, which was published in April 2005, and it included requirements on repair and replacement methods and extended the scope of specific inspection rules for structures other than the BWR Core Shroud and Shroud Support. Along with the efforts of the JSME on the development of the FFS Code, Nuclear and Industrial Safety Agency, the Japanese regulatory agency approved and endorsed the 2000 and 2002 editions of the FFS Code as the national rule, which has been in effect since October 2003. The endorsement for the 2004 edition of the FFS Code is now in the review process.

Fatigue-Strength-Reduction Factors for Welds in Pressure Vessels and Piping

2000 ◽  
Vol 122 (3) ◽  
pp. 297-304 ◽  
Author(s):  
Carl E. Jaske
Keyword(s):  
Fatigue Strength ◽  
Weld Joint ◽  
Pressure Vessels ◽  
Strength Reduction ◽  
Current Status ◽  
Published Data ◽  
Weld Joints ◽  
Class 1 ◽  
Asme Code ◽  
Reduction Factors

Fatigue-strength-reduction factors (FSRFs) are used in the design of pressure vessels and piping subjected to cyclic loading. This paper reviews the background and basis of FSRFs that are used in the ASME Boiler and Pressure Vessel Code, focusing on weld joints in Class 1 nuclear pressure vessels and piping. The ASME Code definition of FSRF is presented. Use of the stress concentration factor (SCF) and stress indices are discussed. The types of welds used in ASME Code construction are reviewed. The effects of joint configuration, welding process, cyclic plasticity, dissimilar metal joints, residual stress, post-weld heat treatment, the nondestructive inspection performed, and metallurgical factors are discussed. The current status of weld FSRFs, including their development and application, are presented. Typical fatigue data for weldments are presented and compared with the ASME Code fatigue curves and used to illustrate the development of FSRF values from experimental information. Finally, a generic procedure for determining FSRFs is proposed and future work is recommended. The five objectives of this study were as follows: 1) to clarify the current procedures for determining values of fatigue-strength-reduction factors (FSRFs); 2) to collect relevant published data on weld-joint FSRFs; 3) to interpret existing data on weld-joint FSRFs; 4) to facilitate the development of a future database of FSRFs for weld joints; and 5) to facilitate the development of a standard procedure for determining the values of FSRFs for weld joints. The main focus is on weld joints in Class 1 nuclear pressure vessels and piping. [S0094-9930(00)02703-7]

Development of Analytical Evaluation Procedures and Acceptance Criteria for Pipe Wall Thinning in ASME Code Section XI

2004 ◽  
Author(s):  
Kunio Hasegawa ◽  
Gery M. Wilkowski ◽  
Lee F. Goyette ◽  
Douglas A. Scarth
Keyword(s):  
Pipe Wall ◽  
Evaluation Methodology ◽  
Current Status ◽  
Evaluation Procedure ◽  
Analytical Evaluation ◽  
Acceptance Criteria ◽  
New Class ◽  
Wall Thinning ◽  
Class 1 ◽  
Asme Code

As the worldwide fleet of nuclear power plants ages, the need to address wall thinning in pressure boundary materials becomes more acute. The 2001 ASME Code Case N-597-1, “Requirements for Analytical Evaluation of Pipe Wall Thinning,” provides procedures and criteria for the evaluation of wall thinning that are based on Construction Code design concepts. These procedures and criteria have proven useful for Code Class 2 and 3 piping; but, they provide relatively little flexibility for Class 1 applications. Recent full-scale experiments conducted in Japan and Korea on thinned piping have supported the development of a more contemporary failure strength evaluation methodology applicable to Class 1 piping. The ASME B&PV Code Section XI Working Group on Pipe Flaw Evaluation has undertaken the codification of new Class 1 evaluation methodology, together with the existing Code Case N-597-1 rules for Class 2 and 3 piping, as a non-mandatory Appendix to Section XI. This paper describes the current status of the development of the proposed new Class 1 piping acceptance criteria, along with a brief review of the current Code Case N-597-1 evaluation procedure in general.

Round Robin Calculations of Collapse Loads—A Torispherical Pressure Vessel Head With a Conical Transition

1997 ◽  
Vol 119 (4) ◽  
pp. 503-509 ◽  
Author(s):  
Y. Yamamoto ◽  
S. Asada ◽  
A. Okamoto
Keyword(s):  
Limit Analysis ◽  
Pressure Vessel ◽  
Standard Procedure ◽  
Round Robin ◽  
Evaluation Procedure ◽  
Analysis Method ◽  
Calculation Methods ◽  
Secondary Stress ◽  
Asme Code ◽  
Limit Analysis Method

Round robin calculations of collapse loads for a pressure vessel were made by 16 teams in Japan. The model is composed of a cylinder and a torispherical head with a conical transition. The structure is an example in which the stress classifications specified in the ASME Code are not strictly applicable. The calculations were performed to clarify the issue of the evaluation procedure using the limit analysis method specified in the ASME Code, Sect. III, and to check the sensitivity of calculation models and programs. It is found that the stress in the knuckle region has certain characteristics of secondary stress, yet still dominates the collapse of the vessel. Using the limit analysis to prove the validity of stress classifications is recommended. The sensitivity of the calculation methods is not so significant. Therefore, it is concluded that the limit analysis can be used as a standard procedure in regulations.

Implementation of ASME Code, Section XI, Code Case N-770, on Alternative Examination Requirements for Class 1 Butt Welds Fabricated With Alloy 82/182

2011 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Michael T. Anderson
Keyword(s):  
Stress Corrosion Cracking ◽  
Nuclear Regulatory Commission ◽  
Butt Welds ◽  
Division 1 ◽  
Class 1 ◽  
Final Rule ◽  
Asme Code ◽  
Areas Of Concern ◽  
Regulatory Commission ◽  
The U.S

In May 2010, the U.S. Nuclear Regulatory Commission (NRC) issued a proposed notice of rulemaking (75 FR 24324) [1] that includes a new section to its rules to require licensees to implement ASME Code Case N–770, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without the Application of Listed Mitigation Activities, Section XI, Division 1,” [2] with 15 conditions. Code Case N-770 contains baseline and inservice inspection (ISI) requirements for unmitigated Alloy 82/182 butt welds and preservice and ISI requirements for mitigated Alloy 82/182 butt welds. The NRC stated that application of ASME Code Case N-770 is necessary because the inspections currently required by the ASME Code, Section XI, were not written to address stress corrosion cracking of Alloy 82/182 butt welds, and the safety consequences of inadequate inspections can be significant. The NRC expects to issue the final rule incorporating this Code Case into its regulations toward the middle of 2011. This paper discusses the new examination requirements, the conditions that NRC proposed to impose, and potential areas of concern with implementation of the new Code Case.

Influence of Head Thickness on Yield Pressure for Cylindrical Pressure Vessels

1974 ◽  
Vol 96 (2) ◽  
pp. 113-120 ◽  
Author(s):  
Andre´ Biron ◽  
Jean Veillon
Keyword(s):  
Limit Analysis ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Additional Material ◽  
Asme Code

Results are presented for the limit analysis of pressure vessel heads of torispherical and ellipsoidal shapes in order to evaluate the influence of different head thicknesses for a given cylinder thickness. Comparison is made with presently used configurations as recommended by the ASME Code. It is found in particular that increasing the knuckle thickness of a torispherical head would provide a significant increase in yield pressure without excessive additional material.

Comparison and Harmonization of French RCCM and ASME Code

2010 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Development Process ◽  
State Of The Art ◽  
Strong Support ◽  
Design Stage ◽  
Industrial History ◽  
Global Quality ◽  
Class 1 ◽  
International Regulations ◽  
Asme Code ◽  
New Ideas

The objectives of this paper is to discuss technical harmonization of Nuclear Codes and Standards, based on French long experience in Codes and Standards used for design-fabrication and operation of nuclear components (mainly pressure retaining components). After a long period of use of ASME Section III code, during the Westinghouse licensing process, AFCEN (AREVA, EDF and the major manufacturers) decided to develop their own AFCEN French Codes. The 1st version has been issued in 1980 and the last one in 2007, completed by annual addendum. During more than 20 years the 2 Codes, RCCM and ASME Section III, have leave separately, with different constraints like industrial history, localisation of fabrication, more new plants in France than in USA, different R&D programs to support Code improvement… Recently a detailed review of differences for class 1 vessel has showed under a “general global quality equivalence”, a lot of differences in the Code development process, in the Code organization, in the scopes, in the State of the Art fulfillment, in ageing consideration at the design stage, in relation with national or international regulations, in term of standards used or complementary specification needs… The harmonization of Codes and Standards is possible under an important effort to move toward new ideas, more international rules and with a strong support of national safety authorities.

Environmental Fatigue Evaluation for Primary Components of APR1400 Reactor Coolant System

2010 ◽  
Author(s):  
J. M. Kim ◽  
K. W. Kim ◽  
K. S. Yoon ◽  
S. H. Park ◽  
I. Y. Kim ◽  
...  
Keyword(s):  
Environmental Effects ◽  
Integral Method ◽  
Power Reactor ◽  
Light Water Reactor ◽  
Outlet Nozzle ◽  
Design Life ◽  
Class 1 ◽  
Asme Code ◽  
Fatigue Evaluation ◽  
Coolant System

USNRC Regulatory Guide (RG) 1.207 provides a guideline for evaluating fatigue analyses due to the environmental effects on the new light water reactor (LWR). The environmental correction factor (Fen) is used to incorporate the LWR environmental effect into fatigue analyses of ASME Class 1 components. In this paper, the environmental fatigue evaluation is applied to some primary components with 60 year design life of Advanced Power Reactor (APR1400). The materials sampled from Class 1 components are the low alloy steel for the reactor vessel (RV) outlet nozzle and the carbon steel for the hot leg which are attached to the outlet nozzle. The simplified method, time-based integral method and strain-based integral method are used to compute the Fen values. The calculated fatigue usage factors including the environmental effects are compared with those obtained using the current ASME Code rules. As the calculated cumulative fatigue usage factor considering environmental effects (CUFen) is below 1.0, there is no concern for the RV outlet nozzle to implement design for environmental fatigue effects.

Adaptation of BPVC Section XI Requirements: The Operator’s and the Regulator’s Perspective

2009 ◽  
Author(s):  
P. Babics ◽  
S. Ratkai ◽  
D. Szabo ◽  
P. Trampus
Keyword(s):  
Structural Integrity ◽  
Management Program ◽  
Regulatory Regime ◽  
Substantial Part ◽  
Efficient Operation ◽  
Normative Documents ◽  
Integrity Assessment ◽  
Class 1 ◽  
Asme Code ◽  
Cost Efficient

The owner of Paks NPP, Hungary’s nuclear generating facility, is aiming at adjusting the ISI program to meet ASME Code requirements. The objective is to achieve an internationally acceptable level in structural integrity assessment of long-lived and passive components, and to create the basis for a proper ageing management program for the operations period beyond design life of the units. Apart from this, it would allow to extend the current four-year inspection interval for Class 1 components up to an eight-year one, which would contribute to a more cost-efficient operation and maintenance. Hungarian nuclear regulatory regime gives an opportunity for this because the nuclear safety regulation does not determine explicitly the applicable codes neither for the design nor for the ISI. First, the basic regulatory principles related to ASME adaptation will be summarized. They focus on aspects of maintaining the current licensing basis as well as on the necessity to demonstrate the compliance with Section III requirements. The substantial part of the work is the construction review of selected Class 1 and 2 components. Then, the results of comparison of the current ISI program, mainly based on Russian normative documents, and the Section XI based one will be shown. These comparative studies have justified the feasibility of the project. The licensing of the ASME based ISI program is under way, and the regulator’s position will be presented as well.

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