Risk-Informed Extension of the Reactor Vessel Nozzle Inservice Inspection Interval

2011 ◽  
Author(s):  
Stephen M. Parker ◽  
Nathan A. Palm ◽  
Paul R. Stevenson ◽  
Bruce A. Bishop
Keyword(s):  
Pressure Vessel ◽  
Reactor Vessel ◽  
Significant Cost ◽  
Pilot Studies ◽  
Failure Frequency ◽  
Core Damage ◽  
Inspection Interval ◽  
American Society ◽  
Specific Implementation ◽  
Technical Basis

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code specifies a 10-year interval between reactor vessel (RV) nozzle weld inspections. The industry has expended significant cost and man-rem exposure performing inspections that have found no service-induced flaws in ASME Section XI Category B-F or B-J RV nozzle welds that do not contain Alloy 82/182. Furthermore, many plants have implemented a 20-year inspection interval for the RV shell-to-shell and shell-to-nozzle welds in accordance with WCAP-16168-NP-A, Revision 2. For many of these plants, continuing to inspect the RV nozzle welds on a 10-year interval presents a significant hardship without a corresponding increase in safety from performing the inspections. This paper will provide a summary of the technical basis and methodology developed by Westinghouse for extending the Section XI inspection interval from the current 10 years to 20 years for Category B-F and B-J RV nozzle-to-safe-end and safe-end-to-pipe welds that are not fabricated with Alloy 82/182 materials. Bounding change-in-failure-frequency values have been calculated for use in plant-specific implementation of the extended inspection interval. Plant-specific pilot studies have been performed and the results show that the change in risk associated with extending the interval from 10 to 20 years after the initial 10-year inservice inspection (ISI) satisfies the guidelines specified in Regulatory Guide 1.174 for an acceptably low change in risk for core damage frequency (CDF) and large early release frequency (LERF). Further, the pilot-plant results show that the effect of the extended inspection interval on the plant’s risk-informed inservice inspection (RI-ISI) program for piping, if any, would also be acceptable.

Risk-Informed Pressurized Water Reactor Vessel Inspection Interval Extension

2004 ◽  
Author(s):  
Cheryl L. Boggess ◽  
Bruce A. Bishop ◽  
Nathan A. Palm ◽  
Owen F. Hedden
Keyword(s):  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pilot Studies ◽  
Water Reactor ◽  
Insignificant Change ◽  
Inspection Interval ◽  
Interval Extension

The methodology discussed in this paper provides a risk informed basis for decreasing the frequency of inspection for the Pressurized Water Reactor (PWR) reactor pressure vessel (RPV). The decrease in frequency is based on extending the interval between inspections from the current interval of 10 years to 20 years. Results of pilot studies on typical designs of PWR vessels show that the change in risk associated with extending the inspection interval by more than 10 years is within the guidelines specified in U.S. Regulatory Guide 1.174 for insignificant change in risk. The current requirements for inspection of reactor vessel pressure-containing welds have been in effect since the 1989 Edition of American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, supplemented by U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.150, June 1981. The manner in which these examinations are conducted has recently been augmented by Appendix VIII of Section XI, 1996 Addenda, as implemented by the NRC in amendment to 10CFR50.55a effective November 22, 1999. This paper summarizes the insignificant change in risk results for the PWR pilot-plant studies, including the effects of fatigue crack growth and in-service inspection of postulated surface-breaking flaws. These results demonstrate that the proposed RPV inspection interval extension is a viable option for the industry.

Technical Basis for Proposed ASME Section XI Code Case on Beyond Design Bases Earthquake

2016 ◽  
Author(s):  
Paul R. Donavin ◽  
Ramiz Gilada ◽  
Harry Gustin ◽  
Truong Vo ◽  
Raymond Pace
Keyword(s):  
High Risk ◽  
Pressure Vessel ◽  
Seismic Loading ◽  
Seismic Events ◽  
Allowable Stress ◽  
Reactor Coolant ◽  
A Value ◽  
Design Basis ◽  
American Society ◽  
Technical Basis

This paper will provide the bases for the requirements in the Beyond Design Basis Events (BDBE) evaluation performed in accordance with the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (BPVC), Section XI, Code Case (CC). The CC provides rules to facilitate the affected components Return to Service (RTS) after a BDBE magnitude Earthquake. The paper describes the bases for the examination of Reactor Coolant Pressure Boundary (RCPB) Structures, Systems and Components (SSCs) for determining the actual seismic loading and magnitude. The implications of the examination data based on design allowable stress values. The pipe loads determined shall be used to calculate stresses on the pumps and valves. The paper also describes the methodology for seismic events lasting longer than 100 cycles. The Cumulative Usage Factor (U) due to the event is calculated from individual cycles as; U = U1 + U2 + U3 + ... + Un. Aftershocks are accounted for in the methodology. Fatigue usage from the event that increases the total U to a value greater than 0.8 shall be included as high risk location(s) in the ASME BPVC, Section XI, Inservice Inspection (ISI) Program.

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.

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.

Overview of Revisions to the ASME Boiler and Pressure Vessel Code Section VIII Division 3 for the 2017 Edition and Near Future

2017 ◽  
Author(s):  
Daniel Peters ◽  
Gregory Mital ◽  
Adam P. Maslowski
Keyword(s):  
High Pressure ◽  
Pressure Vessel ◽  
Central Area ◽  
Local Strain ◽  
Pressure Vessels ◽  
Conformity Assessment ◽  
Inspection Planning ◽  
Section Viii ◽  
American Society ◽  
Rules Changes

This paper provides an overview of the significant revisions pending for the upcoming 2017 edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section VIII Division 3, Alternative Rules for Construction of High Pressure Vessels, as well as potential changes to future editions under consideration of the Subgroup on High Pressure Vessels (SG-HPV). Changes to the 2017 edition include the removal of material information used in the construction of composite reinforced pressure vessels (CRPV); this information has been consolidated to the newly-developed Appendix 10 of ASME BPVC Section X, Fiber-Reinforced Plastic Pressure Vessels. Similarly, the development of the ASME CA-1, Conformity Assessment Requirements standard necessitated removal of associated conformity assessment information from Section VIII Division 3. Additionally, requirements for the assembly of pressure vessels at a location other than that listed on the Certificate of Authorization have been clarified with the definitions of “field” and “intermediate” sites. Furthermore, certain design related issues have been addressed and incorporated into the current edition, including changes to the fracture mechanics rules, changes to wires stress limits in wire-wound vessels, and clarification on bolting and end closure requirements. Finally, the removal of Appendix B, Suggested Practice Regarding Post-Construction Requalification for High Pressure Vessels, will be discussed, including a short discussion of the new appendix incorporated into the updated edition of ASME PCC-3, Inspection Planning Using Risk Based Methods. Additionally, this paper discusses some areas in Section VIII Division 3 under consideration for improvement. One such area involves consolidation of material models presented in the book into a central area for easier reference. Another is the clarification of local strain limit analysis and the intended number and types of evaluations needed for the non-linear finite element analyses. The requirements for test locations in prolongations on forgings are also being examined as well as other material that can be used in testing for vessel construction. Finally, a discussion is presented on an ongoing debate regarding “occasional loads” and “abnormal loads”, their current evaluation, and proposed changes to design margins regarding these loads.

Design criteria for exposed hydro penstocks

1978 ◽  
Vol 5 (3) ◽  
pp. 340-351 ◽  
Author(s):  
J. L. Gordon
Keyword(s):  
Quality Control ◽  
Pressure Vessel ◽  
High Strength Steel ◽  
High Strength ◽  
Design Criteria ◽  
Similar Work ◽  
Wide Range ◽  
Different Types ◽  
Control Procedures ◽  
American Society

At present there are no national codes for the design of exposed hydro-electric penstocks. Thus an engineer must either make reference to other national codes for similar work, such as the American Society of Mechanical Engineers boiler and pressure vessel code or the American Water Works Association Standard for steel water piping, or he must write his own code and is then faced with the decision of having to select design criteria that must cover a wide range of steels; different operating and waterhammer conditions; a wide range of quality control procedures used in manufacture and erection of the penstock; and different types of penstocks, isostatic where the stresses can be calculated with precision, and hyperstatic where the stress calculation is more imprecise. This paper discusses design criteria, factors of safety, and corresponding quality control procedures that can be used for either isostatic or hyperstatic penstocks using mild, intermediate, or high strength steel for penstocks supplying reaction of impulse turbines.

Background and Technical Basis for the Evaluation of Flaws in PWR Reactor Vessel Head Penetration Regions

2004 ◽  
Author(s):  
Warren Bamford ◽  
Guy De Boo
Keyword(s):  
Reactor Vessel ◽  
Flaw Size ◽  
Acceptance Criteria ◽  
Head Penetration ◽  
Industry Group ◽  
Surface Flaws ◽  
Critical Flaw ◽  
Technical Basis ◽  
Outside Diameter

Acceptance criteria have been developed for indications found during inspection of reactor vessel in upper head penetrations. These criteria were originally developed for inside surface flaws, as part of an industry program coordinated by NUMARC (now NEI) in 1992. These criteria were not inserted into Section XI at the time, because inspections were not required for these regions. In developing the enclosed acceptance criteria, the approach used by the industry group was similar to that used in other portions of Section XI, in that an industry consensus was reached using input from the operating utility technical staff, each of the three PWR vendors, and representatives of the NRC staff. The criteria developed are applicable to all PWR plant designs. The discovery of leaks at Oconee, ANO-1, and several other plants, have led to the imposition of inspection requirements for head penetration regions, and therefore the need to develop criteria for indications in all portions of the tubes. This would include indications on the inside diameter of the tube, as well as on the outside diameter of the tube below the attachment weld, and flaws in the attachment weld itself. The criteria presented herein are limits on flaw sizes which are acceptable. The criteria are to be applied to inspection results. It should be noted that determination of the period of future service during which the criteria are satisfied is plant-specific and dependent on flaw geometry and loading conditions. It has been previously demonstrated by each of the owners groups that the penetrations are very tolerant of flaws. It was concluded that complete fracture of the penetration would not occur unless very large through-wall flaws were present; therefore, protection against leakage during service is the priority. The approach used here is more conservative than that used in Section XI applications where the acceptable flaw size is calculated by putting a margin on the critical flaw size. In this case, the critical flaw size is far too large to allow a practical application of this approach, so protection against leakage is the key element used to define the acceptance criteria. Also, the use of flaw acceptance standards tables is not allowed for this region, for penetrations which are susceptible to stress corrosion cracking. The acceptance criteria apply to all flaw types regardless of orientation and shape. The same approach is used by Section XI, where flaws are characterized according to established rules and their future predicted size is then compared with the acceptance criteria.

New ASME Rules for Spent Nuclear Fuel and High Level Radioactive Material and Waste Storage Containments

2004 ◽  
Author(s):  
Ralph S. Hill ◽  
Gerald M. Foster
Keyword(s):  
Nuclear Fuel ◽  
Pressure Vessel ◽  
Spent Nuclear Fuel ◽  
The Body ◽  
Radioactive Material ◽  
Comprehensive Understanding ◽  
Waste Storage ◽  
American Society ◽  
High Level

In 2004, a new Code Case, N-717, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) will be published. The new Code Case provides rules for construction of containments used for storage of spent nuclear fuel and high level radioactive material and waste. Some time after publication, CC N-717 will be incorporated into the body of the Code. This paper provides an informative insight to the Code Case so that Owners, regulators, designers, and fabricators have a more comprehensive understanding.

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