ML19176A008
| ML19176A008 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/28/2019 |
| From: | Robert Pascarelli Plant Licensing Branch IV |
| To: | Welsch J Pacific Gas & Electric Co |
| Singal B, NRR/DORL/LPL4, 301-415-3016 | |
| References | |
| EPID L-2018-LLR-0138 | |
| Download: ML19176A008 (10) | |
Text
Enclosure June 28, 2019 Mr. James M. Welsch Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Nuclear Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424
SUBJECT:
DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 - REQUEST FOR RELIEF FROM THE INSERVICE INSPECTION PROGRAM REQUIREMENTS OF ASME CODE, SECTION XI, FOR THE STEAM GENERATOR AND PRESSURIZER NOZZLE INSIDE RADIUS SECTIONS (EPID L-2018-LLR-0138)
Dear Mr. Welsch:
By letter dated November 7, 2018 (Agencywide Documents Access and Management System Accession No. ML18311A391), Pacific Gas and Electric Company (the licensee) submitted Relief Request NDE-SG-PZR-IRS for relief from the inservice inspection (ISI) program requirements of Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for the fourth ISI interval, applicable to certain ASME Code Class 1 component welds.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR)
Section 50.55a(z)(2), the licensee requested relief on the basis that performing the ASME Code required examination for steam generator and pressurizer nozzle inside radius sections for Diablo Canyon Nuclear Power Plant (Diablo Canyon), Units 1 and 2 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Based on the enclosed safety evaluation, the U.S. Nuclear Regulatory Commission (NRC) staff has determined that the proposed alternative provides reasonable assurance that structural integrity and leak tightness of the subject components will be maintained during the fourth ISI interval at Diablo Canyon, Units 1 and 2. The staff finds that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in quality and safety. Accordingly, the staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the staff authorizes the use of Relief Request NDE-SG-PZR-IRS for the fourth 10-year ISI interval at Diablo Canyon, Units 1 and 2. The fourth ISI interval began on May 7, 2015, for Unit 1 and March 13, 2016, for Unit 2 and is currently scheduled to end on the license expiration dates of November 2, 2024, for Unit 1 and August 26, 2025, for Unit 2.
J. Welsch All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions concerning this matter, please contact the Project Manager, Mr. Balwant K. Singal at 301-415-3016 or via e-mail at Balwant.Singal@nrc.gov.
Sincerely,
/RA/
Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323
Enclosure:
Safety Evaluation cc: Listserv
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST NDE-SG-PZR-IRS RELIEF FROM ASME CODE, SECTION XI, INSERVICE INSPECTION REQUIREMENTS FOR CERTAIN CLASS 1 COMPONENT WELDS DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
By letter dated November 7, 2018 (Agencywide Documents Access and Management System Accession No. ML18311A391), Pacific Gas and Electric Company (PG&E, the licensee) submitted Relief Request NDE-SG-PZR-IRS for relief from the inservice inspection (ISI) program requirements of Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code) for the fourth ISI interval, applicable to certain ASME Code Class 1 component welds.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR)
Section 50.55a(z)(2), Hardship without a compensating increase in quality and safety, the licensee requested relief on the basis that performing the ASME Code required examination for steam generator (SG) and pressurizer nozzle inside radius sections for Diablo Canyon Nuclear Power Plant (Diablo Canyon), Units 1 and 2, would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements in 10 CFR 50.55a throughout the service life of a boiling-or pressurized-water reactor. The exception is the design and access provisions and preservice examination requirements set forth in Section XI of editions and addenda of the ASME Code that become effective subsequent to editions specified in 10 CFR 50.55a(g)(2) and 10 CFR 50.55a(g)(3), which are incorporated by reference in paragraph 10 CFR 50.55a(a)(1)(ii) to the extent practical within the limitations of design, geometry, and materials of construction of the components.
Volumetric examination of the volume depicted in Figure IWB-2500-7(a) through (d), as applicable, once during the inspection interval for all SG nozzle inside radius sections and all pressurizer nozzle inside radius sections, with acceptance standard IWB-3512, is required by the 1998 Edition as referenced in 10 CFR 50.55a(b)(2)(xxi)(A).
Pursuant to 10 CFR 50.55a(b)(2)(xxi)(A), [a] visual examination with magnification that has a resolution sensitivity to resolve 0.044 inch (1.1 mm [millimeter]) lower case characters without an ascender or descender (e.g., a, e, n, v), utilizing the allowable flaw length criteria in Table IWB-3512-1, 1997 Addenda through the latest edition and addenda incorporated by reference in [10 CFR 50.55a(a)(1)(ii)], with a limiting assumption on the flaw aspect ratio (i.e., a/l=0.5 [(where a is the crack depth and l is the crack length)]), may be performed instead of the ultrasonic examination.
Section 50.55a(z), Alternatives to codes and standards requirements, of 10 CFR states, in part, that alternatives to the requirements of paragraphs (b) through (h) addressed in 10 CFR 50.55a may be used, when authorized by the U.S. Nuclear Regulatory Commission (NRC), if the licensee demonstrates that (1) the proposed alternatives would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC staff to grant the relief requested by the licensee.
3.0 TECHNICAL EVALUATION
3.1 Licensees Relief Request NDE-SG-PZR-IRS 3.1.1 ASME Code Components Affected This relief request for Diablo Canyon, Units 1 and 2, addresses the ASME Code Class 1 SG and pressurizer Nozzle Inside Radius Sections; ASME Section XI, Table IWB-2500-1, Category B-D, Items B3.120 and B3.140 (Inspection Program B)1 of the 1998 Edition, as specified in 10 CFR 50.55a(b)(2)(xxi)(A).
3.1.2 Applicable Code Edition and Addenda The Code edition and addenda applicable to the Diablo Canyon, Units 1 and 2, for the fourth ISI interval is the ASME Code,Section XI, 2007 Edition through 2008 Addenda. Use of Table IWB-2500-1, Category B-D, Items B3.120 and B3.140 (Inspection Program B) of the 1998 Edition is required by 10 CFR 50.55a(b)(2)(xxi)(A).
1 Per 10 CFR 50.55a(b)(2)(xxi)(A), the provisions of IWB 2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels and, Items B3.40 and B4.60 (Inspection Program A) and Items B3.120 and B3.140 (Inspection Program B) of the 1998 Edition must be applied when using the 1999 Addenda through the latest Edition and addenda incorporated by reference in 10 CFR 50.55a(a)(1)(ii).
3.1.3 Applicable Code Requirement In its letter dated November 7, 2018, the licensee states, in part that:
Volumetric examination of the volume depicted in [ASME Code, Section XI]
Figure IWB-2500-7(a) through (d), as applicable, once during the inspection interval for all steam generator nozzle inside radius sections and all pressurizer nozzle inside radius sections, with acceptance standard IWB-3512 is required by the 1998 edition as referenced in 10 CFR 50.55a(b)(2)(xxi)(A).
Per 10 CFR 50.55a(b)(2)(xxi)(A), a visual examination with magnification that has a resolution sensitivity to resolve 0.044 inch (1.1 mm) lower case characters without an ascender or descender (e.g., a, e, n, v), utilizing the allowable flaw length criteria in [ASME Code, Section XI] Table IWB-3512-1, 1997 Addenda through the latest edition and addenda incorporated by reference in 10 CFR 50.55a(a)(1)(ii), with a limiting assumption on the flaw aspect ratio (i.e., all=0.5), may be performed instead of the ultrasonic examination.
3.1.4 Licensees Reason for Request In its letter dated November 7, 2018, the licensee states, in part that:
ASME Code Case N-619[2], which eliminated the ASME Code requirement for nozzle inside radius examinations on Class 1 steam generators and pressurizers, was approved by ASME on February 15, 1999. Also, at that time, the ASME Section XI eliminated the requirement to examine steam generator and pressurizer nozzle inside radius sections in all editions and addenda of Table IWB-2500-1, Category B-D, after the 1998 edition. No relevant indications had been found at U.S. pressurized water reactor (PWR) facilities at that time, which remains the case today.
All Class 1 steam generator and pressurizer nozzle inside radius sections are covered with corrosion resistant, welded stainless steel cladding which is not credited as part of the pressure retaining boundary. During visual examinations conducted per 10 CFR 50.55a(b)(2)(xxi)(A), the actual pressure boundary is thus not visible due to the presence of the protective cladding.
3.1.5 Proposed Alternative The licensee proposed not to perform the inside radius section examinations on the SG and pressurizer nozzles at Diablo Canyon, Units 1 and 2, as currently specified in 10 CFR 50.55a(b)(2)(xxi)(A) during the current fourth ISI interval. In addition, the licensee proposed not to perform volumetric examination for the inside radius section of the SG and pressurizer nozzles in accordance with the ASME Code,Section XI, Table IWB-2500-1.
Furthermore, the licensee stated that the cumulative industry experience has demonstrated that neither the volumetric nor the visual examinations have indicated any evidence of degradation that warrants personnel exposure to the industrial safety hazard and radiation dose inherent in performing these examinations.
2 Alternative Requirements for Nozzle Inner Radius Inspections for Class 1 Pressurizer and Steam Generator Nozzles,Section XI, Division 1 3.1.6 Licensees Basis for Relief Request 3.1.6.1 SG Nozzles Inside Radius Sections Licensees Inspection of SG Nozzles Inside Radius Sections The licensee stated that the Diablo Canyon SG nozzles are SA-508 Grade 3 Class 2 low alloy steel, integrally forged with the channel head and then clad with weld-deposited stainless steel.
As addressed in Table 1, Head Weld and Pipe Weld Distances from Edge of Inside Radius Examination Volume (estimated based on vendor drawings), of the letter dated November 7, 2018, nozzle-to-pipe welds are located far from the edge of the inside radius examination volume. The licensee further stated that the heat affected zone (HAZ) of the nozzle-to-pipe welds would not affect the scope of the inside radius examinations. The SG nozzle inside radius sections are challenging to examine ultrasonically due to the varying external geometry and surface contours around the nozzle circumference. In past inspection intervals, the licensee obtained NRCs approval of relief from conducting volumetric examination of SG nozzle inside radius sections due to the continuously variable external geometry and surface contours making ultrasonic examination impractical. The licensee stated that the provision for alternate visual inspection in 10 CFR 50.55a(b)(2)(xxi)(A) was intended to make these SG examinations physically practicable by using a non-volumetric technique.
The new Unit 1 SGs were installed during the 15th refueling outage in 2009. The Unit 2 SGs were installed during the 14th refueling outage in 2008. Extensive construction and preservice inspections were conducted prior to service, including the nozzle inside radius sections, with no flaws identified.
Furthermore, the licensee stated that visual examinations of the SG nozzles entail radiation exposures to the personnel due to multiple entries in the SG. The SG bowl entries put examination personnel at risk of physical injury both during passage through the manways and while handling the camera system components inside the confines of the SG bowl. Based on the third interval experience, the cumulative radiation exposure required to examine all SG nozzles during the fourth interval would be approximately 800 millirem for each unit (i.e., 1600 millirem for both units). Additional dose would be incurred for data recording and incidental support. These totals do not include the dose required for installing nozzle dams, which also support other activities.
Despite the attendant difficulties, required visual examinations of SG and pressurizer nozzle inside radius sections were performed during subsequent outages in the third inspection interval with no recordable indications and no degradation detected. The licensee stated that visual examinations of the nozzles performed during the third ISI interval revealed no degradation nor any recordable indications. Therefore, there is no active aging degradation in the radius portion of the SG nozzles at Diablo Canyon, Units 1 and 2.
Inspection of Original SGs - Operating Experience (OE)
In its letter dated November 7, 2018, the licensee states, in part that:
The original steam generators had nozzle dam rings welded to the stainless steel cladding after construction was complete. This welding caused local fissures in the cladding near the nozzle inside radius sections that was determined to be arrested in the cladding. Those local fissures did not propagate into the base material. In the new steam generators, nozzle dam rings were incorporated during vessel fabrication and did not cause conditions that were observed in the original steam generators. The radius of the nozzles in the original steam generators and the current steam generators are approximately the same. In summary, no service-related degradation was identified in the nozzle inside radius sections of the original steam generators. This OE was consistent with the overall industry experience.
3.1.6.2 Licensees Inspection of pressurizer nozzles inside radius sections In its letter dated November 7, 2018, the licensee states, in part that:
The Unit 1 pressurizer nozzles are SA-216 Grade WCC carbon steel, integrally cast with the vessel heads and then clad with stainless steel. The Unit 2 pressurizer nozzles are SA-508 Class 2 carbon steel [and consistent with the design addressed in ASME Code] Section XI Figure IWB-2500-7(b), welded into the vessel heads and then clad with stainless. The spray and surge nozzles in both units each have stainless steel thermal sleeves.
As addressed in Table 1 of letter dated November 7, 2018, nozzle-to-pipe welds and the vessel head weld (Unit 2) are located far from the edge of the inside radius examination volume.
Therefore, the HAZ of the welds would not affect the scope of the inside radius examinations.
Ultrasonic examinations of the pressurizer nozzle inside radius sections are not practical to examine visually due to difficulty in achieving suitable accessibility for inspecting the area.
Additionally, any inspections in the pressurizers would result in an increase in additional radiation exposure to the personnel. The ultrasonic examinations specified in the 1998 Edition of the ASME Code,Section XI, were performed in the third ISI interval at Diablo Canyon, Units 1 and 2, and these examinations resulted in additional radiation exposure to the inspection team.
Both the Diablo Canyon, Units 1 and 2, pressurizers are original equipment having entered service in 1985 and 1986, respectively. Preservice inspections and followup ISIs conducted since that time through the third inspection interval have not detected any flaw conditions or degradation in the nozzle inside radius sections.
In its letter dated November 7, 2018, the licensee further states that:
PG&E performed a detailed review of the recorded industry experience for steam generator and pressurizer nozzle inside radius sections at U.S. PWR facilities.
This review did not identify any detected flaw indications and did not find the presence of any degradation mechanism. Therefore, PG&E has not performed any postulated flaw growth calculations for the nozzle inside radius sections.
3.1.7 Duration of Proposed Relief Request The proposed relief would apply for the duration of the current fourth ISI interval through the termination of the operating licenses. For Unit 1, the fourth interval started on May 7, 2015, and the license expires on November 2, 2024. For Unit 2, the fourth interval started on March 13, 2016, and the license expires in August 26, 2025.
3.2
NRC Staff Evaluation
The NRC staff reviewed the licensees justification for deleting the visual and volumetric examinations of the SGs and pressurizers inside radius examinations during the fourth ISI interval at Diablo Canyon, Units 1 and 2. Based on the information provided in its letter dated November 7, 2018, the NRC staff determined the following:
- 1.
All Class 1 SG and pressurizer nozzle inside radius sections are covered with corrosion resistant, welded stainless steel cladding. To date, there has been no evidence of any aging degradation nor any relevant indications on the stainless-steel cladding of the nozzles.
- 2.
The nozzle-to-pipe welds and vessel head weld for Unit 2 are located away from the all the SG and pressurizer nozzle inside radius sections. Thus, the HAZ associated with these welds do not affect the inside radius sections.
- 3.
Volumetric examinations (ultrasonic examinations) of the SG and pressurizer nozzle inside radius sections are not practical due to difficulty in achieving suitable accessibility and/or the contour of the nozzle surface area. The NRC has approved the licensee previously submitted relief for substituting visual examinations in lieu of ultrasonic examinations of the SG and pressurizer nozzle inside radius sections. Substitution of visual examinations in lieu of ultrasonic examinations of the SG and pressurizer nozzle inside radius sections was allowed in accordance with the provisions addressed in 10 CFR 50.55a(b)(2)(xxi)(A).
- 4.
The industry inspection results to date show no active aging degradation nor any reportable indications in the inside radius sections of nozzles in the SG and pressurizer.
- 5.
Based on Item 4 above, the ASME Committee issued ASME Code Case N-619 on February 15, 1999, which eliminated the ASME Code requirement for nozzle inside radius examinations on Class 1 SGs and pressurizers.
- 6.
The ASME Code eliminated the requirement to examine SG and pressurizer nozzle inside radius sections in Table IWB-2500-1, Category B-D in all editions and addenda after the 1998 Edition.
- 7.
Ultrasonic examinations and visual examinations would entail additional radiation exposure to personnel and cause hardship or unusual difficulty without a compensating increase in the level of quality and safety.
- 8.
With regard to the postulated flaw growth calculations for the nozzle inside radius sections, the NRC staff determined that flaw growth calculations are not required for this situation because of the following: (1) no active degradation in the inner radius of the nozzles was identified for the past 30 years of operation; (2) presence of stainless cladding on the inner radius surface of the nozzle provides extra protection from any aging degradation; (3) no cracking to date, was identified on the stainless cladded surface in PWRs due to presence of low oxygen in the reactor coolant system water; and (4) Diablo Canyon, Unit 1, is expected to cease operations in year 2024 and Diablo Canyon, Unit 2, in the year 2025. Therefore, these units will be in service for only 5 and 6 years, respectively. With few years of service left, degradation of the stainless steel cladded surface is highly unlikely. Based on the above reasons, the NRC staff did not determine the need for flaw growth calculations for Diablo Canyon, Units 1 and 2.
Based on the above (technical bases discussed in Items 1 through 8), the NRC staff noted that to date, no active aging degradation and no recordable indications on the inside radius of the nozzles associated with the SG and pressurizer have been observed. Therefore, the staff concludes that there is reasonable assurance that the structural integrity of the nozzles would be maintained during the fourth ISI interval at Diablo Canyon, Units 1 and 2.
4.0 CONCLUSION
S As set forth above, the NRC staff determined that the proposed alternative provides reasonable assurance that structural integrity and leak tightness of the subject components will be maintained during fourth ISI interval at Diablo Canyon, Units 1 and 2. The NRC staff finds that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the staff authorizes the use of Relief Request NDE-SG-PZR-IRS for the fourth 10-year ISI interval at Diablo Canyon, Units 1 and 2.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Ganesh Cheruvenki, NRR/DMLR/MPHB Date: June 28, 2019
- SE received via email dated June 12, 2019 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DMLR/MPHB/BC NRR/DORL/LPL4/BC NAME BSingal PBlechman ABuford(A)
RPascarelli DATE 06/28/19 06/27/19 06/12/19*
06/28/19