ML020390041
| ML020390041 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 03/05/2002 |
| From: | Gramm R NRC/NRR/DLPM/LPD4 |
| To: | Cottle W South Texas |
| References | |
| TAC MB1277, TAC MB1278 | |
| Download: ML020390041 (13) | |
Text
March 5, 2002 Mr. William T. Cottle President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station P. O. Box 289 Wadsworth, TX 77483
SUBJECT:
APPROVAL OF RELIEF REQUEST FOR APPLICATION OF RISK-INFORMED INSERVICE INSPECTION PROGRAM FOR AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE CLASS 1 AND 2 PIPING FOR SOUTH TEXAS PROJECT, UNITS 1 AND 2 (TAC NOS. MB1277 AND MB1278)
Dear Mr. Cottle:
By letter dated February 27, 2001, you requested approval of an alternative risk-informed inservice inspection (RI-ISI) program for American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (ASME Code) Class 1 and 2 piping welds for South Texas Project (STP), Units 1 and 2. The letter included an enclosure describing the proposed program. Additional information was provided in your letter dated January 10, 2002. By letter dated September 11, 2000, you received approval for an RI-ISI plan for ASME Code Class 1 welds (excluding socket welds), which is not affected by the RI-ISI program and will be retained.
The RI-ISI program was developed in accordance with Electric Power Research Institute Topical Report TR-112657, Revision B-A, using the Nuclear Energy Institute template methodology. Based on the enclosed safety evaluation, the NRC staff concluded that the proposed RI-ISI program is an acceptable alternative to the requirements of Section XI of the ASME Code for ISI. Therefore, your request for relief is authorized pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative provides an acceptable level of quality and safety. The relief is authorized for the second 10-year ISI interval for STP, Units 1 and 2.
Sincerely,
/RA/
Robert A. Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499
Enclosure:
Safety Evaluation cc w/encls: See next page
March 5, 2002 Mr. William T. Cottle President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station P. O. Box 289 Wadsworth, TX 77483
SUBJECT:
APPROVAL OF RELIEF REQUEST FOR APPLICATION OF RISK-INFORMED INSERVICE INSPECTION PROGRAM FOR AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE CLASS 1 AND 2 PIPING FOR SOUTH TEXAS PROJECT, UNITS 1 AND 2 (TAC NOS. MB1277 AND MB1278)
Dear Mr. Cottle:
By letter dated February 27, 2001, you requested approval of an alternative risk-informed inservice inspection (RI-ISI) program for American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (ASME Code) Class 1 and 2 piping welds for South Texas Project (STP), Units 1 and 2. The letter included an enclosure describing the proposed program. Additional information was provided in your letter dated January 10, 2002. By letter dated September 11, 2000, you received approval for an RI-ISI plan for ASME Code Class 1 welds (excluding socket welds), which is not affected by the RI-ISI program and will be retained.
The RI-ISI program was developed in accordance with Electric Power Research Institute Topical Report TR-112657, Revision B-A, using the Nuclear Energy Institute template methodology. Based on the enclosed safety evaluation, the NRC staff concluded that the proposed RI-ISI program is an acceptable alternative to the requirements of Section XI of the ASME Code for ISI. Therefore, your request for relief is authorized pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative provides an acceptable level of quality and safety. The relief is authorized for the second 10-year ISI interval for STP, Units 1 and 2.
Sincerely,
/RA/
Robert Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499 DISTRIBUTION:
PUBLIC SAli
Enclosure:
Safety Evaluation PDIV-1 r/f SDinsmore RidsNrrDlpmLpdiv (S.Richards) MRubin cc w/encls: See next page RidsNrrDlpmLpdiv (R.Gramm) TChan RidsNrrPMJDonohew SMorris, EDO RidsNrrPMMThadani GHill (4)
RidsAcrsAcnwMailCenter RidsOgcRp RidsRgn4MailCenter(D. Graves, L. Hurley, D. Bujol)
ACCESSION NO: ML020390041
- EMCB and PSAB Memo dated February 4, 2002 OFFICE PDIV-2/PM PDIV-1/LA EMCB/SC SPSB/SC OGC PDIV-1/SC NAME JDonohew:as MMcAllister TChan* MRubin* RHoefling RGramm DATE 02/21/02 03/05/02 02/04/2002 02/04/2002 02/20/02 03/05/02 OFFICIAL RECORD COPY
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RISK-INFORMED INSERVICE INSPECTION PROGRAM STP NUCLEAR OPERATING COMPANY, ET AL.
SOUTH TEXAS PROJECT, UNITS 1 AND 2 DOCKET NOS. 50-498 AND 50-499
1.0 INTRODUCTION
By application dated February 27, 2001 (Reference 1), South Texas Project (STP) Nuclear Operating Company (the licensee) proposed a risk-informed inservice inspection (RI-ISI) program as an alternative to a portion of their current inservice inspection (ISI) program for STP, Units 1 and 2. The scope of the RI-ISI program is limited to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (the ASME Code) Class 1 and 2 piping (Categories B-F, B-J, C-F-1, and C-F-2 welds) only. Additional information was provided in a letter from the licensee dated January 10, 2002 (Reference 2). The licensees letter dated January 10, 2002, was in response to the Nuclear Regulatory Commission (NRC) staffs request for additional information dated December 10, 2001.
The licensees RI-ISI program was developed in accordance with the methodology contained in the Electric Power Research Institute (EPRI) Topical Report (TR) EPRI TR-112657, Revision B-A (Reference 3), which was previously reviewed and approved by the NRC staff.
Reference 3 contains the letter issued by the NRC staff on October 28, 1999, that approved the TR. The RI-ISI program proposed by the licensee is an alternative pursuant to 10 CFR 50.55a(a)(3)(i).
2.0 BACKGROUND
2.1 Applicable Requirements Pursuant to 10 CFR 50.55a(g), the ISI of the ASME Code Class 1, 2, and 3 components must be performed in accordance with Section XI of the ASME Code, Rules for Inservice Inspection of Nuclear Power Plant Components, and applicable addenda, except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i). The regulation, 10 CFR 50.55a(a)(3), states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the proposed alternatives would provide an acceptable level of quality and safety, or if the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements set forth in the ASME Code, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that ISI of components conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.
STP Unit 1 began its second 10-year interval on September 25, 2000, and STP Unit 2 began its second 10-year interval on October 19, 2000. The applicable edition of the ASME Code,Section XI for both STP, Units 1 and 2 is the 1989 edition, no addenda.
2.2 Summary of Proposed Approach The licensee has proposed to use a RI-ISI program for ASME Code Class 1 socket-welded piping (Examination Category B-J welds) and Class 2 piping welds (Examination Categories C-F-1 and C-F-2), as an alternative to the ASME Code,Section XI requirements. The ASME Code requires in part that for each successive 10-year ISI interval, 25 percent of Category B-J socket welds for the ASME Code Class 1 non-exempt piping be selected for surface examination, based on existing stress analyses and cumulative usage factors. For Category C-F welds, 7.5 percent of non-exempt welds are selected for volumetric and/or surface examination. The application follows the NRC staff-approved RI-ISI process and methodology delineated in Reference 3.
The number of inspection locations is significantly reduced by assessing piping failure potential and piping failure consequences, and performing probabilistic risk assessment (PRA) and safety significance ranking of piping segments. However, the RI-ISI program retains the fundamental requirements of the ASME Code, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Thus, ISI program requirements of other non-related portions of the ASME Code,Section XI are unaffected.
The licensee stated that the augmented examinations to address pipe cracks in stagnant borated water systems documented in NRC Bulletin 79-17, "Pipe Cracks in Stagnant Borated Water Systems at PWR Plants," dated July 26, 1979, have been subsumed into the RI-ISI program as allowed by Reference 3. The augmented ISI program for flow accelerated corrosion (FAC) implemented in response to NRC Bulletin 89-08, Erosion/Corrosion - Induced Pipe Wall Thinning, is not changed by the RI-ISI program. The licensee also indicated that the augmented inspection program for high energy No Break Zone piping is not affected by this RI-ISI program. Other remaining augmented ISI programs are either unaffected or modified in accordance with the guidance of Reference 3.
According to the information provided in Reference 1, STP, Units 1 and 2 are currently at the start of the first period of their second 10-year interval. The licensee stated that 100 percent of the RI-ISI examinations will be performed during the second interval. The licensee further stated that examinations shall be performed during the interval such that the period examination percentage requirements of ASME Section XI, paragraphs IWB-2412 and IWC-2412, are met.
The implementation of a RI-ISI program for piping should be initiated at the start of a plants 10-year ISI interval consistent with the requirements of the ASME Code and Addenda committed to by the licensee in accordance with 10 CFR 50.55a. However, the implementation may begin at any point in an existing interval, as long as the examinations are scheduled and distributed consistent with the ASME Code requirements (e.g., the minimum examinations completed at the end of the three inspection periods under ASME Code Program B should be 16 percent, 50 percent, and 100 percent, respectively, and the maximum examinations credited at the end of the respective periods should be 34 percent, 67 percent, and 100 percent, respectively).
It is also the NRC staffs view that the inspections for the RI-ISI program and for the balance of the ISI program should be on the same interval start and end dates. This can be accomplished by either implementing the RI-ISI program at the beginning of the interval, or merging the RI-ISI program into the ISI program for the balance of the inspections if the RI-ISI program is to begin during an existing ISI interval. One reason for this view is that it eliminates the problem of having different ASME Codes of record for the RI-ISI program and for the balance of the ISI program. A potential problem, with using two different interval start dates, and hence two different ASME Codes of record, would be having two sets of repair/replacement rules depending upon which program identified the need for repair (e.g., a weld inspection versus a pressure test). According to the information provided in Reference 1, the licensee will merge the RI-ISI program into the existing ISI program so that the 10-year interval start and end dates will not be impacted.
3.0 EVALUATION Pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff has reviewed and evaluated the licensees proposed RI-ISI program, including those portions related to the applicable methodology and processes contained in Reference 3, based on guidance and acceptance criteria provided in Regulatory Guides 1.174 (Reference 4) and 1.178 (Reference 5), and in Standard Review Plan Chapter 3.9.8 (Reference 6).
3.1 Proposed Changes to the ISI Program The scope of the licensees proposed RI-ISI program is limited to ASME Code Class 1 and Class 2 piping welds for the following examination categories: B-J for pressure retaining socket-welded piping, C-F-1 for pressure retaining welds in austenitic stainless steel or high alloy piping, and C-F-2 for pressure retaining welds in carbon or low alloy steel piping. The RI-ISI program is proposed as an alternative to the existing ISI requirements of the ASME Code,Section XI. A general description of the proposed changes to the ISI program is provided in Sections 3 and 5 of Attachment 1 to Reference 1.
During the course of its review, the NRC staff verified that the proposed RI-ISI program is consistent with the guidelines contained in Reference 3, which states that industry and plant-specific piping failure information, if any, is to be utilized to identify piping degradation mechanisms and failure modes, and consequence evaluations are to be performed using PRAs to establish piping segment safety ranking for determining new inspection locations. Thus, the NRC staff concludes that the licensees application of the Reference 3 approach is an acceptable alternative to the current STP, Units 1 and 2 piping ISI requirements with regards to
the number, locations, and methods of inspections, and provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(a)(3)(i).
3.2 Engineering Analysis In accordance with the guidance provided in References 4 and 5, an engineering analysis of the proposed changes is required using a combination of traditional engineering analyses and supporting insights from the PRA. The licensee elaborated as to how the engineering analyses conducted for the STP, Units 1 and 2 RI-ISI program ensure that the proposed changes are consistent with the principles of defense-in-depth. This is accomplished by evaluating a locations susceptibility to a particular degradation mechanism and then performing an independent assessment of the consequence of a failure at that location. No changes to the evaluation of design-basis accidents in the final safety analysis report are being made in the RI-ISI process. Therefore, sufficient safety margins will be maintained.
The licensees RI-ISI program at STP, Units 1 and 2 is limited to ASME Code Class 1 socket-welded piping and Class 2 piping welds. The licensee stated in Reference 1 that other non-related portions of the ASME Code will be unaffected by this program. Piping systems defined by the scope of the RI-ISI program were divided into piping segments. Pipe segments are defined as lengths of pipe whose failure leads to similar consequences and are exposed to the same degradation mechanisms. That is, some lengths of pipe whose failure would lead to the same consequences may be split into two or more segments when two or more regions are exposed to different degradation mechanisms.
In Reference 1, the licensee stated that failure potential categories were generated utilizing industry failure history, plant-specific failure history, and other relevant information using the guidance provided in Reference 3. The degradation mechanisms identified in the submittal include thermal fatigue, including thermal stratification, cycling and striping (TASCS), and thermal transients; and FAC. The licensee stated in Section 2.2 of Attachment 1 to Reference 1, that the augmented inspection program for FAC is relied upon to manage this mechanism, and is not changed by the RI-ISI program.
In Section 3 of Attachment 1 to Reference 1, the licensee described a deviation to the EPRI RI-ISI methodology for assessing the potential for TASCS that was implemented by the licensee. In Reference 2, the licensee stated that the methodology for assessing TASCS in the STP, Units 1 and 2 RI-ISI submittal is identical to the Materials Reliability Project (MRP) methodology in EPRI TR-1000701, "Interim Thermal Fatigue Management Guideline (MRP-24)," January 2001. The NRC staff has reviewed the guidance for evaluating TASCS in EPRI TR-1000701 and finds it to be acceptable. The licensee further stated that it will update the RI-ISI program based on the final EPRI MRP guidance if warranted.
Additionally, the licensee stated that the consequences of pressure boundary failure were evaluated and ranked based on their impact on core damage and containment performance (isolation, bypass and large early release), and that the impact due to both direct and indirect effects was considered using guidance provided in Reference 3. The licensee reported no deviations from the consequence evaluation methodology in Reference 3. The licensee further stated that shutdown operation and external events are included in the analysis. Based on the above discussion, the NRC staff finds that the consequence evaluation performed for this application to be acceptable.
3.3 PRA The licensee evaluated the consequences of pipe rupture for Reference 1 using the Level 2 Probabilistic Safety Assessment and Individual Plant Examination, dated August 1992, supplemented by the PRA model, STP_1997 (hereafter STP_1997 PRA). This is the same version of the PRA that was used to support the licensee's December 30, 1999, RI-ISI submittal for Class 1 welds (excluding socket welds) that was approved by the NRC staff on September 11, 2000 (Reference 7). In Reference 2, the licensee clarified that the STP_1997 PRA reflected the design and operation of the plants up to December 31, 1997, and that it was approved for use for STP, Units 1 and 2 in March 1999, and the analyses to support Reference 1 were prepared between June and September 2000. The licensee further clarified that a new PRA model, referred to as STP_1999, was approved for use by STP in October 2001.
Reference 4 includes the NRC staff expectation that the engineering analyses conducted to justify a proposed change, should be based on the as-built and as-operated and maintained plant. Therefore the PRA used to support a risk-informed submittal should reflect the current characteristics of the plant, or the licensee should review modifications not reflected in the PRA and conclude, if applicable, that the modifications would have a negligible impact on the RI-ISI results if included in the PRA. The NRC staff has not yet defined current in terms of a fixed time between the last PRA update and the preparation of a submittal based on that update.
The time period between December 1997, and June of 2000, is a longer time period than previous RI-ISI submittals have reported. However, the RI-ISI methodology is relatively insensitive to variations in the PRA results. Additionally, monitoring a RI-ISI program requires feedback of new information to ensure the appropriate categorization of the safety significance of piping locations. New information that is significant requires frequent adjustment, and the risk ranking is to be reviewed and adjusted on an ASME period basis as a minimum. Because of the relative insensitivity to variations in the PRA results and the periodic up-date requirements, the NRC staff finds that the PRA accepted for use to support the December 30, 1999, submittal is also acceptable for use to support Reference 1.
The degradation category and the consequence category were combined according to the approved methodology described in Reference 3 to categorize the risk significance of each segment. The risk significance of each segment is used to determine the number of weld inspections required in each segment.
As required by Section 3.7 of Reference 3, the licensee evaluated the change in risk expected from replacing the current ISI program for Class 1 socket-welded piping and Class 2 piping welds with the RI-ISI program. The calculations do not include changes in inspections locations for the remaining Class 1 welds (excluding socket welds) approved by the NRC staff in Reference 7. The calculations estimated the change in risk due to removing locations and adding locations to the inspection program. The expected change in risk was quantitatively evaluated using the Simplified Risk Quantification Method described in Section 3.7.2 of Reference 3. For high consequence category segments, the licensee used the conditional core damage probability (CCDP) and conditional large early release probability (CLERP) based on the highest estimated CCDP and CLERP. For medium consequence category segments, bounding estimates of CCDP and CLERP were used.
The licensee performed its bounding analysis with, and without, taking credit for an increased probability of detection (POD). In Reference 1, the licensee estimated the aggregate changes in core damage frequency (CDF) and large early release frequency (LERF), as listed in the following table:
Change in CDF Change in LERF without improved with improved without improved with improved POD POD POD POD Unit 1 -4.0E-11/year -1.2E-10/year -4.0E-12/year -1.2E-11/year Unit 2 -1.0E-11/year -9.0E-11/year -1.0E-12/year -9.0E-12/year Including an improved POD will decrease the risk of the RI-ISI program but not change the risk of the current program. Therefore, if the change in risk is negative without the improved POD, it will be a larger negative value with the improved POD.
The NRC staff finds the licensees process to evaluate and bound the potential change in risk is reasonable because it accounts for the change in the number and location of elements inspected, recognizes the difference in degradation mechanism related to failure likelihood, and considers the effects of enhanced inspection. As noted in the NRC staff safety evaluation (Reference 7) for the licensee's submittal of December 30, 1999, discussed on the previous page, the estimated changes in risk due to implementing the RI-ISI program on Class 1 welds (excluding socket welds) met all the acceptance guideline values in Reference 3. In Reference 1, the licensee reported that all system level and aggregate estimates of the changes in CDF and LERF are less than zero, indicating a reduction in risk due to implementing the RI-ISI program on Class 1 socket-welded piping and Class 2 piping welds. Because all the estimated changes in risk in Reference 1 are negative, the estimated cumulative impact due to implementing the RI-ISI program on all Class 1 welds (including socket welds) and all Class 2 piping welds will also meet all the acceptance guideline values in Reference 3. The NRC staff finds that re-distributing the welds to be inspected with consideration of the safety-significance of the segments provides assurance that segments whose failures have a significant impact on plant risk receive an acceptable and often improved level of inspection. Therefore, the NRC staff concludes that the implementation of the RI-ISI program as described in Reference 1 will have a small impact on risk consistent with the guidelines of Reference 4.
3.4 Integrated Decisionmaking As described in Reference 1, an integrated approach is utilized in defining the proposed RI-ISI program by considering, in concert, the traditional engineering analysis, risk evaluation, and the implementation and performance monitoring of piping under the program. This is consistent with the guidelines of Reference 5.
The selection of pipe segments to be inspected using the results of the risk category rankings and other operational considerations is described in Section 3.5 of Attachment 1 of Reference 1. Tables 3.5-1 and 3.5-2 provide the number of locations and inspections by risk category for the various systems in STP, Units 1 and 2, respectively. Tables 5-2-1 and 5-2-2
compares the number of inspections required under the existing ASME Code,Section XI ISI program with the alternative RI-ISI program for STP, Units 1 and 2, respectively. The risk impact analysis results for each system are provided in Tables 3.6-1 and 3.6-2 for STP, Units 1 and 2, respectively. The licensee used the methodology described in Reference 3 to guide the selection of examination elements within high and medium risk-ranked piping segments. The methodology described in Reference 3 requires that existing augmented programs, other than thermal fatigue and intergranular stress-corrosion cracking Category A piping welds, which the RI-ISI program subsumes, be maintained. Reference 3 describes targeted examination volumes (typically associated with welds) and methods of examination based on the type(s) of degradation expected. The NRC staff has reviewed these guidelines and has determined that, if implemented as described, the RI-ISI examinations should result in improved detection of service-related degradations over those currently required by ASME Code,Section XI.
The NRC staff finds that the location selection process is acceptable since it is consistent with the process approved for Reference 3, takes into account defense-in-depth, and includes coverage of systems subjected to degradation mechanisms in addition to those covered by augmented inspection programs.
The objective of the ISI required by ASME Code,Section XI is to identify conditions (i.e., flaw indications) that are precursors to leaks and ruptures in the pressure boundary that may impact plant safety. Therefore, the RI-ISI program should meet this objective if found to be acceptable for use. Further, since the risk-informed program is based on inspection for cause, element selection should target specific degradation mechanisms. The inspection for cause approach involves identification of specific damage mechanisms that are likely to be operative, the location where they may be operative, and appropriate examination methods and volumes specific to address the damage mechanisms.
Chapter 4 of Reference 3 provides guidelines for the areas and/or volumes to be inspected, as well as the examination method, acceptance standard, and evaluation standard for each degradation mechanism. Based on review of the cited portion of Reference 3, the NRC staff concludes that the examination methods for the proposed RI-ISI program are appropriate since they are selected based on specific degradation mechanisms, pipe sizes, and materials of concern.
3.5 Implementation and Monitoring Implementation and performance monitoring strategies require careful consideration by the licensee and are addressed in Element 3 of References 5 and 6. The objective of Element 3 is to assess the performance of the affected piping systems under the proposed RI-ISI program by implementing monitoring strategies that confirm the assumptions and analyses used in the development of the RI-ISI program. To approve an alternative pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff must conclude that implementation of the RI-ISI program, including inspection scope, examination methods, and methods of evaluation of examination results, must provide an acceptable level of quality and safety.
The licensee stated that, upon approval of the RI-ISI program, procedures that comply with the guidelines in Reference 3 will be prepared to implement and monitor the RI-ISI program. The licensee confirmed that the applicable portions of the ASME Code, such as inspection methods,
acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements would be retained.
The licensee stated in Section 4 of Attachment 4 to Reference 1 that the RI-ISI program is a living program and its implementation will require feedback of new, relevant information to ensure the appropriate identification of safety significant piping locations. The submittal also states that, as a minimum, risk ranking of piping segments will be reviewed and adjusted on an ASME Code period basis, and that significant changes may require more frequent adjustment as directed by NRC bulletin or generic letter requirements, or by industry and plant-specific feedback.
In response to the NRC staffs request for further clarification, the licensee stated in Reference 2 that the ISI program will be updated and submitted to the NRC consistent with regulatory requirements in effect at the time such update is required (currently every 10 years).
The licensee stated that this may again take the form of a relief request to implement an updated RI-ISI program depending on future regulatory requirements. Reference 2 also stated that the RI-ISI program will be resubmitted to the NRC prior to the end of any 10-year ISI interval if there is some deviation from the RI-ISI methodology described in Reference 1, or if industry experience determines that there is a need for significant revision to the program as described in Reference 1.
The licensee presented the criteria, in Reference 1, for engineering evaluations and additional examinations if unacceptable flaws or relevant conditions are found during examinations. The licensee further stated in Reference 1 that the evaluation will include whether other elements in the segment or segments are subject to the same root cause conditions. In Reference 2, the licensee clarified that additional examinations will be performed on these elements up to a number equivalent to the number of elements required to be inspected on the segment or segments scheduled for the STP Unit 1 October 2001 refueling outage. Reference 2 also stated that elements selected for additional examinations will be selected based on the root cause or damage mechanism, and will include high risk-significant, as well as medium risk-significant elements (if needed), to reach the required number of additional elements.
The proposed periodic reporting requirements meet existing ASME Code requirements and applicable regulations and, therefore, are acceptable. The NRC staff finds that the proposed process for RI-ISI program updates meets the guidelines of Reference 4 which provide that risk-informed applications should include performance monitoring and feedback provisions; therefore, the licensees proposed process for program updates is acceptable.
4.0 CONCLUSION
S In accordance with 10 CFR 50.55a(a)(3)(i), proposed alternatives to regulatory requirements may be used when authorized by the NRC when the applicant demonstrates that the alternative provides an acceptable level of quality and safety. In this case, the licensee's proposed alternative is to use the risk-informed process described in the NRC-approved Reference 3.
The NRC staff finds that the results of the different elements of the engineering analysis are considered in an integrated decisionmaking process. The impact of the proposed change in the ISI program is founded on the adequacy of the engineering analysis and acceptable change in plant risk in accordance with the guidelines in References 4 and 5.
The STP, Units 1 and 2 methodology also considers implementation and performance monitoring strategies. Inspection strategies ensure that failure mechanisms of concern have been addressed and there is adequate assurance of detecting damage before structural integrity is affected. The risk significance of piping segments is taken into account in defining the inspection scope for the RI-ISI program.
System pressure tests and visual examination of piping structural elements will continue to be performed on all Class 1, 2, and 3 systems in accordance with the ASME Code,Section XI program. The RI-ISI program applies the same performance measurement strategies as existing ASME Code requirements and, in addition, increases the inspection volumes at weld locations that are exposed to thermal fatigue.
The STP, Units 1 and 2 methodology provides for conducting an engineering analysis of the proposed changes using a combination of engineering analysis with supporting insights from a PRA. Defense-in-depth and quality are not degraded in that the methodology provides reasonable confidence that any reduction in existing inspections will not lead to degraded piping performance when compared to existing performance levels. Inspections are focused on locations with active degradation mechanisms, as well as selected locations that monitor the performance of system piping. As discussed in Section 3.2 above, the licensee will address any NRC staff concern, if applicable, as a result of a separate, ongoing review on the generic report MRP-24 regarding alternative TASCS screening criteria.
The NRC staffs review of the licensees proposed RI-ISI program concludes that the program is an acceptable alternative to the current ISI program, which is based on ASME Code,Section XI, requirements for Class 1 socket-welded piping and Class 2 piping welds. In Section 3.1 above, the NRC staff concluded that the licensees proposed RI-ISI program, as described in its application and supplemental responses to the NRC staff, will provide an acceptable level of quality and safety pursuant to 10 CFR 50.55a(a)(3)(i) with regard to the number of inspections, locations of inspections, and methods of inspections. Therefore, the licensees request for relief is authorized pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the request provides an acceptable level of quality and safety. This safety evaluation authorizes application of the proposed RI-ISI program during the second 10-year ISI interval for STP, Units 1 and 2.
In Reference 7, STP, Units 1 and 2 received approval from the NRC for an RI-ISI plan for ASME Class 1 welds (excluding socket welds), which is not affected by this RI-ISI program and will be retained.
5.0 REFERENCES
- 1. Letter, dated February 27, 2001, T.J. Jordan to U. S. Nuclear Regulatory Commission, containing Relief Request for Application of an Alternative to the ASME Boiler and Pressure Vessel Code Section XI Examination Requirements for Class 1 Socket-Welded Piping and Class 2 Piping Welds (RR-ENG-2-23), South Texas Project, Units 1 and 2.
- 2. Letter, dated January 10, 2002, T.J. Jordan to U. S. Nuclear Regulatory Commission, containing Response to Request for Additional Information Regarding Risk-Informed
Inservice Inspection Application for Section XI Examination Requirements for Class 1 and 2 Piping Welds (RR-ENG-2-23), South Texas Project, Units 1 and 2.
- 3. EPRI TR-112657, Revision B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure, January 2000.
- 4. NRC Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, July 1998.
- 5. NRC Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping, September 1998.
- 6. NRC NUREG-0800, Chapter 3.9.8, Standard Review Plan for Trial Use for the Review of Risk-Informed Inservice Inspection of Piping, September 1998.
- 7. Letter, Robert A. Gramm (U.S. Nuclear Regulatory Commission) to William T. Cottle (STP Nuclear Operating Company), South Texas Project, Units 1 and 2 - Request for Relief from ASME Code Requirements for the Second 10-year Interval Inservice Inspection Program Based on Risk-Informed Alternative Approach (Relief Request RR-ENG-2-16), dated September 11, 2000.
Principal Contributors: Syed Ali Stephen Dinsmore Date: March 5, 2002
South Texas, Units 1 & 2 cc:
Mr. Cornelius F. OKeefe A. H. Gutterman, Esq.
Senior Resident Inspector Morgan, Lewis & Bockius U.S. Nuclear Regulatory Commission 1111 Pennsylvania Avenue, NW P. O. Box 910 Washington, DC 20004 Bay City, TX 77414 Mr. J. J. Sheppard, Vice President A. Ramirez/C. M. Canady Engineering & Technical Services City of Austin STP Nuclear Operating Company Electric Utility Department P. O. Box 289 721 Barton Springs Road Wadsworth, TX 77483 Austin, TX 78704 S. M. Head, Manager, Licensing Mr. M. T. Hardt Nuclear Quality & Licensing Department Mr. W. C. Gunst STP Nuclear Operating Company City Public Service Board P. O. Box 289, Mail Code: N5014 P. O. Box 1771 Wadsworth, TX 77483 San Antonio, TX 78296 Office of the Governor Mr. C. A. Johnson/R. P. Powers ATTN: John Howard, Director AEP - Central Power and Light Company Environmental and Natural P. O. Box 289 Resources Policy Mail Code: N5022 P. O. Box 12428 Wadsworth, TX 77483 Austin, TX 78711 INPO Jon C. Wood Records Center Matthews & Branscomb 700 Galleria Parkway 112 East Pecan, Suite 1100 Atlanta, GA 30339-3064 San Antonio, TX 78205 Regional Administrator, Region IV Arthur C. Tate, Director U.S. Nuclear Regulatory Commission Division of Compliance & Inspection 611 Ryan Plaza Drive, Suite 400 Bureau of Radiation Control Arlington, TX 76011 Texas Department of Health 1100 West 49th Street D. G. Tees/R. L. Balcom Austin, TX 78756 Houston Lighting & Power Co.
P. O. Box 1700 Jim Calloway Houston, TX 77251 Public Utility Commission of Texas Electric Industry Analysis Judge, Matagorda County P. O. Box 13326 Matagorda County Courthouse Austin, TX 78711-3326 1700 Seventh Street Bay City, TX 77414 February 2002