ML14295A532

Catawba Nuclear Station Units 1 and 2: Proposed Relief Request 14-CN-001, American Society of Mechanical Engineers (ASME) Section Ix Volumetric Examination Requirements (TAC Nos. MF3527 and MF3528)

ML14295A532
Person / Time
Site:	Catawba
Issue date:	10/30/2014
From:	Pascarelli R J Plant Licensing Branch II
To:	Henderson K Duke Energy Carolinas
Miller G E
References
TAC MF3527, TAC MF3528
Download: ML14295A532 (10)

Text

UNITED STATES REGu( o.,. ... NUCLEAR REGULATORY COMMISSION

(") 0 ;;: '</.

_..o **** ... Mr. Kelvin Henderson Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 297 45 WASHINGTON, D.C. 20555-0001 October 30, 2014

SUBJECT:

CATAWBA NUCLEAR STATION UNITS 1 AND 2: PROPOSED RELIEF REQUEST 14-CN-001, AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI VOLUMETRIC EXAMINATION REQUIREMENTS (TAC NOS. MF3527 AND MF3528)

Dear Mr. Henderson:

By letter dated February 20, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 14052A387),

as supplemented by letter dated September 11, 2914 (ADAMS Accession No. ML 14265A044),

Duke Energy Carolinas, LLC requested relief from the volumetric examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for the third 10-year inservice inspection (lSI) interval for Catawba Nuclear Station (Catawba),

Units 1 and 2. Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(g),

the licensee requested relief from the volumetric examination requirements.

The third 1 0-year lSI interval began on June 29, 2005 for Catawba, Unit 1, and October 15, 2005, for Catawba, Unit 2. The U.S. Nuclear-Regulatory Commission (NRC) staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, the examinations exams were performed to the extent practical and provide reasonable assurance of structural integrity of the subject areas. Therefore, the NRC staff grants relief as requested in 14-CN-003.

All other ASME Code,Section XI, requirements, for which relief was not specifically requested and authorized herein by the NRC staff, remain applicable, including the third-party review by the Authorized Nuclear In-service Inspector.

K. Henderson If you have any questions, please contact the Project Manager, Ed Miller at 301-415-2481 or via e-mail at Ed.Miller@nrc.gov.

Docket Nos. 50-413 and 50-414

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ Sincerely, Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

.* UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 14-CN-001 THIRD 10-YEAR INSERVICE INSPECTION INTERVAL DUKE ENERGYCAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414

1.0 INTRODUCTION

By letter dated February 20, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 14052A387),

as supplemented by letter dated September 11, 2014 (ADAMS Accession No. ML 14265A044),

Duke Energy Carolinas, LLC (the licensee) requested relief from the volumetric examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for the third 10-year in service inspection (lSI) interval for Catawba Nuclear Station (Catawba),

Units 1 and 2. The third 1 0-year lSI interval began on June 29, 2005 for Catawba, Unit 1, and October 15, 2005, for Catawba, Unit 2. This lSI interval ended on July 15, 2014 for Catawba, Unit 1, and is currently scheduled to end August 19, 2016, for Catawba, Unit 2. 2.0 REGULATORY EVALUATION The ASME Code Class 1, 2, and 3 components for the lSI is performed in accordance with Section XI, as required by Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(g),

except where specific relief has been granted by the U.S. Nuclear Regulatory Commission (NRC) in accordance with 10 CFR 50.55a(g)(6)(i).

Pursuant to 10 CFR 50.55a(g)(4),

ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components,"

to the extent practical within the limitations of design, geometry, and materials of construction of the components.

The regulations require that inservice examination of components and system pressure tests conducted during the first 1 0-year lSI interval and subsequent lSI intervals comply with the Enclosure requirements in the latest edition and addenda of Section XI 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.

The Code of Record for the Catawba, Units 1 and 2, third 1 0-year lSI Intervals is the 1998 Edition through the 2000 Addenda of the ASME Code. The third 1 0-year lSI Interval.

for Catawba, Unit 1, started on June 29, 2005, and is currently scheduled to end on August 19, 2015. The third 1 0-year lSI Interval for Unit 2 started on October 15, 2005, and is currently scheduled to end on August 19, 2015. With respect to examination

coverage, "essentially 100 percent" is clarified by ASME Code Case N-460, "Alternative Examination for Class 1 and Class 2 Welds," to be greater than 90 percent coverage of the examination volume, or surface area, as applicable.

ASME Code Case N-460 has been approved for use by the NRC in Regulatory Guide 1.147, Revision 16, "lnservice Inspection Code Case Acceptability."

3.0 PROPOSED ALTERNATE FOR CATAWBA UNITS 1 AND 2 3.1 Code Requirements ASME Code,Section XI, 1998 Edition through the 2000 Addenda, Table IWB-2500-1, Category B A, "Pressure Retaining Welds in Reactor Vessel,"

Item No. 81.40, "Head-to-Flange Weld," requires volumetric examination of essentially 100 percent of the weld length as depicted in Figure IWB-2500-5.

Subsubarticle IWA-2230, "Volumetric Examination,"

of Section XI of the Code identifies volumetric examination techniques that may be employed for the examinations specified in Table IWB-2500-1.

The licensee elected to use ultrasonic examination techniques for . examination of the welds for which relief was requested.

For ultrasonic examination techniques, Paragraph IWA-2232, "Ultrasonic Examination,"

of Section XI of the ASME Code specifies that ultrasonic examination shall be conducted in accordance with Appendix I, "Ultrasonic Examinations,"

For the single reactor vessel weld included in 14-CN-001 for Catawba, Unit 1, Subsubarticle 1-2110, "Reactor Vessels,"

of Appendix I to Section XI of the ASME Code specifies that ultrasonic examination of the reactor vessel-to-flange weld shall be conducted in accordance with Article 4 of Section V of the ASME Code, as supplemented by Table 1-2000-1, except that alternative examination beam angles may be used. Therefore, the licensee's examination for this weld was described in RR No. 14-CN-001 as using scanning requirements in ASME Code,Section V, Article 4, T-441.1.2(a),

T-441.1.3, T-441.1.4, T-441.1.5 and T-441.1.6.

Use of the provisions of* Section Vindicates that the requirements of Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems,"

to Section XI of the ASME Code were not utilized for the reactor vessel weld examination.

ASME Code,Section XI, 1998 Edition through the 2000 Addenda, Table IWB-2500-1, Category 8-D, "Full Penetration Welded Nozzles in Vessels -Inspection Program 8," Item No. 83.11 0," Nozzle-to-Vessel Welds" (Pressurizer) requires volumetric examination of 100 percent of the weld and adjacent base material as depicted in Figures IWB-2500-?(a) through (d). The licensee's submittal indicates that Figure IWB-2500-?(b) is applicable for the pressurizer nozzle configurations included in RR No. 14-CN-001.

Since the licensee elected to use ultrasonic examination techniques for examination of the subject nozzle welds, Appendix I to Section XI of the Code also applies to these welds. For the eight pressurizer nozzle welds included in 14-CN-001, Subsubarticle I 2120, "Other Vessels,"

of Appendix I to Section XI of the ASME Code specifies that ultrasonic examination of all other vessels greater than 2 inches in thickness shall be conducted in accordance with Article 4 of Section V, as supplemented by Table I 2000-1. Therefore, the licensee's nozzle examinations were described in RR No. 14-CN-001 as using manual scanning with conventional methods in accordance with ASME Code,Section V, Article 4, T-441.1.2(a),

T-441.1.3, T-441.1.4, T-441.1.5 and T-441.1.6, indicating that the requirements of Appendix VIII to Section XI of the Code were not utilized for the pressurizer nozzle examinations.

3.2 Component Identification

1. ASME Code,Section XI, Table IW8-2500-1, Examination Category 8-A, "Pressure Retaining Welds in Reactor Vessel,"

Item No. 81.40,"Head-to-Fiange Weld." Components:

Catawba, Unit 1, Reactor Vessel Weld: Head-to-Flange Weld Summary No. C1.81.40.0001, Weld No. 1-RPV-WOB

• 2. ASME Code,Section XI, Table IW8-2500-1, Examination Category 8-D, "Full Penetration Welded Nozzles in Vessels-Inspection Program 8," Item No. 83.110, "Nozzle-to-Vessel Welds." Catawba.

Unit 1 Components:

Welds: Pressurizer Surge Nozzle and Pressurizer Safety/Relief Valve Nozzles Pressurizer Surge Nozzle-to-Head Circumferential Weld Summary No. C1.83.110.0001, Weld No. 1-PZR-W1 Pressurizer Safety Valve Nozzle-to-Head Circumferential Welds Summary No. C1.83.110.0004, Weld No. 1-PZRW4A; Summary No. C1.B3.110.0005, Weld No. 1-PZR-W4B; and Summary No. C1.B3.110.0006, Weld No. 1-PZR-W4C

Catawba, Unit 2 Components:

Welds: Pressurizer Surge Nozzle and Pressurizer Safety/Relief Valve Nozzles Pressurizer Surge Nozzle-to-Head Circumferential Weld Summary No. C2.83.110.0001, Weld No. 2PZR-W1 Pressurizer Safety Valve Nozzle-to-Head Circumferential Welds Summary No. C2.83.110.0003, Weld No. 2PZR-W3; Summary No. C2.83.110.0004, Weld No. 2PZR-W4A; and Summary No. C2.83.110.0005, Weld No. 2PZR-W48 3.3 Licensee's Relief Request (as stated) 1. Relief is requested from the requirement of Table IWB-2500-1, Examination Category B-A, Item B1.40, to examine the volume A-B-C-D specified in Figure IWB-2500-5, and from the requirement of footnote (2) to examine "essentially 100% of the weld length" for the Catawba, Unit 1, Reactor Vessel Head to Flange Weld (Summary No. C1.B1.40.0001, Weld No. 1-RPV-W08).

2. Relief is requested from the requirement of Table IWB-2500-1, Examination*

Category B-0, Item B3.11 0, to examine the volume A-B-C-0-E-F-G-H specified in Figure IWB-2500-?(b) for the Catawba, Units 1 and 2, Category B-0, Item B3.110, welds listed in 1 [of the licensee's February 20, 2014 submittal].

3.4 Licensee's Basis for Requesting Relief (as stated) The welds identified in this request were examined to the maximum extent practicable utilizing approved examination techniques and equipment in accordance with requirements of the ASME Code,Section XI, IWB-2500, Table IWB-2500-1, Examination Category B-'-A, Item B1.40, and Examination Category B-0, Item B3.110, as applicable.

No substitution alternative for these welds is available that would provide better coverage.

Radiography (RT) is not a desired option in lieu of ultrasonic examination because RT is limited in the ability to detect service induced flaws. Use of other manual or automated ultrasonic techniques, whether conventional or phased array, were considered, but would not increase coverage due to the limitation created by the component configuration.

The use of any other currently available technique would incur the same physical scanning limitations.

System leakage tests are performed each refueling outage in accordance with the ASME Code,Section XI, Table IWB-2500-1; Examination Category B-P, during which a visual examination

• (VT-2) is performed to detect evidence of leakage.

These tests and the VT-2 visual examinations provide additional assurance of pressure boundary integrity.

In addition to the system leakage tests and VT -2 visual examinations, Reactor Building Containment Sump monitoring provides additional assurance that, in the event that leakage did occur through any of these welds, leakage would be detected.

Based on the acceptable results for the coverage obtained by the volumetric examination, system leakage testing and VT-2 examinations, and leakage monitoring, it is the license's position that the combination of these examinations and tests provides reasonable assurance of the continued structural and leak tight integrity of these welds. 3.5 Licensee's Proposed Alternative Examination (as stated) Because the specified requirements of the ASME Code,Section XI, cannot be met without redesigning the components, and because there are no alternatives available that can be used to achieve greater volumetric coverage for the welds, compliance with the requirements specified in Section 3.0 of [the licensee's February 20, 2014] request is determined to be impractical and relief is requested in accordance with 10 CFR 50.55a(g)(5)(iii). 4.0 NRC STAFF EVALUATION 4.1 Catawba, Unit 1, Reactor Vessel' Head-to-Flange Weld ASME Code,Section XI, 1998 Edition through the 2000 Addenda, Table IWB-2500-1, Category B-A, "Pressure Retaining Welds in Reactor Vessel,"

Item No. 81.40, "Head-to-Flange Weld," requires volumetric examination of essentially 100 percent of the weld length. The design configuration of Catawba, Unit 1, reactor vessel head-to-flange weld precludes an ultrasonic examination of essentially 100 percent of the required weld length. The component design configuration limits ultrasonic test examination coverage of the weld to 82.2 percent coverage (Table 1 of licensee's February 20, 2014, submittal).

In order to examine the weld in accordance with ASME Code requirements, the reactor vessel head would require extensive design modifications.

In their request, the licensee indicated that the limited examination coverage for this weld was caused by the configuration of the head ring-to-flange weld that does not allow meaningful interrogation from the flange side. The licensee also indicated that there are three permanent lifting lugs that do not allow scanning from the head ring side in the vicinity of the lugs. In order to all of the required volume for this weld and meet the ASME Code requirements, the head ring-to-flange weld and the permanent lifting lugs would have to be redesigned or replaced to allow scanning from both sides of the weld to facilitate access for UT search units, which is impractical.

Previous examinations of the reactor vessel head-to-flange weld were completed during the Catawba, Unit 1, second 1 0-year lSI interval.

The percentage of examination coverage obtained in the Third 10-Year lSI Interval (82.2 percent coverage) was similar to the coverage previously reported for the second 1 0-year lSI interval.

No relevant indications were noted or recorded during the examination.

Additionally, as indicated in the licensee's submittal, the weld for which relief was requested receives a VT-2 examination in conjunction with the Class I System Leakage Test conducted during each refueling outage in accordance with ASME Code requirements to compliment the limited examination coverage.

Consequently, the reactor vessel has undergone several VT-2 examinations during past refueling outages.

The periodic VT-2 examinations are supplemented by Reactor Building Containment Sump monitoring that provides additional assurance of leakage detection through the head-to-flange weld. The VT-2 examinations of the reactor vessel head area, which will continue to be performed during each refueling outage in

• conjunction with the ASME Code, Class 1, System Leakage Test, and the leakage monitoring will provide reasonable assurance of the leak tightness of the head-to-flange weld because the system leakage test will provide for detection of flaws when they are small and can be repaired prior to the reactor vessel head-to-flange weld losing its ability to perform its* intended function.

Based on the foregoing, the NRC staff determined that the ASME Code requirements are

• impractical for the Catawba, Unit 1, reactor vessel head-to-flange weld. The NRC staff has, based on the volumetric coverage obtained for the Catawba, Unit 1, head-to-flange weld, determined that it is reasonable to conclude that if significant service-induced degradation had occurred in the head-to-flange weld, evidence of degradation would have been detected in the portion that was examined.

Therefore, the NRC staff concludes that the examinations performed provide reasonable assurance of structural integrity of the head-to-flange weld of the reactor vessel. 4.2 Catawba, Units 1 and 2, Pressurizer Surge Nozzle and Safety/Relief Valve Head Circumferential Welds ASME Code,Section XI, 1998 Edition through the 2000 Addenda, Table IWB-2500-1, Category B D, "Full Penetration Welds of Nozzles in Vessels,"

Item No. 83.11 Vessel Welds," requires volumetric examination of 100 percent of the weld and adjacent base material.

The design configuration of pressurizer surge and safety/relief valve nozzle-to-vessel welds precludes ultrasonic examination of essentially 100 percent of the required volumes.

The component design configurations limits ultrasonic examination coverage of these welds to approximately 80 percent coverage (Table 1 of licensee's February 20, 2014 submittal).

In order to examine the welds in accordance with ASME Code requirements, the pressurizer would require extensive design modifications.

The licensee noted in their request that the impracticality for the pressurizer surge nozzle for both Catawba units is caused by the weld taper configuration of the nozzle-to-head that does not allow meaningful interrogation from the nozzle side. In addition, Heater Wells limit access from the head side of the weld. In order to scan all of the required volume for this weld, the head-to-nozzle weld would have to be redesigned or replaced and the heater wells redesigned and relocated to allow scanning from both sides of the weld, which is impractical.

The licensee noted in their request that the impracticality for the pressurizer safety/relief valve

• nozzles for both Catawba units was caused by the weld taper configuration of the head area that does not allow meaningful interrogation from the nozzle side. In order to scan all of the required volume for these welds, the head-to-nozzle welds would have to be redesigned or replaced to allow scanning from both sides of the weld, which is impractical.

Previous examinations of the pressurizer surge and safety/relief valve nozzles were completed during the Catawba, Units 1 and 2, Second 1 0-year lSI intervals.

In a request for additional information, the NRC staff noted a significant increase (more than 38%) in examination coverage for the CNS Unit 2 pressurizer surge nozzle weld (Weld No. 2 PZR-W1) compared to the coverage achieved in the prior examination performed in 2001 for the Second 1 0-Year lSI interval.

In their September 11, 2014, response, the licensee indicated that their ultrasonic procedure was revised in 2005. The revised procedure allowed the use of smaller transducers, as well as lower 35-degree and 45-degree transducer beam angles. These changes led to a considerable increase in the cumulative scan coverage.

The licensee also indicated that the revised procedure is qualified and acceptable based on ASME Code requirements.

The NRC staff finds the licensee's responses to be acceptable.

The percentage of examination coverage amounts obtained in the third 10-year lSI interval (approximately 80 percent coverage) for all other nozzles included in the licensee's request were similar to the coverages previously reported.

For all nozzles included in RR No. 14-CN-001, no relevant indications were noted or recorded during the examinations. In addition, as indicated in the licensee's submittal, the welds for which relief was requested receive a VT-2 examination in conjunction with the Class I System Leakage Test conducted during each refueling outage in accordance with ASME Code requirements to compliment the limited examination coverage.

Consequently, the pressurizer surge and safety/relief valve nozzles have undergone several VT-2 examinations during past refueling outages.

The periodic VT -2 examinations are supplemented by Reactor Building Containment Sump monitoring that provides additional assurance of leakage detection through these welds. The VT-2 examinations of the nozzle areas, which will continue to be performed during each refueling outage in conjunction with the ASME Code, Class 1, System Leakage Test, and the leakage monitoring will provide reasonable assurance of the leak tightness of the nozzle-to-vessel welds because the system leakage test will provide for detection of flaws when they are small and can be repaired prior to the pressurizer surge and safety/relief valve nozzles losing their ability to perform their intended functions.

Based on the foregoing, the NRC staff determined that the ASME Code requirements are impractical for the pressurizer surge and safety/relief valve nozzle-to-vessel welds. Based on the volumetric coverage obtained for the nozzle-to-vessel welds, the NRC staff has determined that it is reasonable to conclude that if significant service-induced degradation had occurred in the nozzle-to-vessel welds, evidence of degradation would have been detected in the portions that were examined.

Therefore, the NRC staff concludes that the examinations performed provide reasonable assurance of structural integrity of the nozzle-to-vessel welds of the pressurizer surge and safety/relief valve nozzles for Catawba, Units 1 and 2.

5.0 CONCLUSION

As set forth above, the NRC staff has determined that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

Furthermore, the staff concluded that the examinations performed to the extent practical provide reasonable assurance of structural integrity of the subject reactor vessel, pressurizer surge* *nozzles, and pressurizer safety/relief valve nozzles.

Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i).

Therefore, the NRC staff grants relief for the subject examinations of the reactor vessel, pressurizer surge nozzles, and pressurizer safety/relief valve nozzles contained in 14-CN-001 for the Catawba, Units 1 and 2, third 1 0-year lSI intervals.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject request for relief remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.

Principal Contributors:

G. Stevens, RES J. Poehler, NRR Date of issuance:

October 30, 2014 If you have any questions, please contact the Project Manager, Ed Miller at 301-415-2481 or via e-mail at Ed.Miller@nrc.gov.

Docket Nos. 50-413 and 50-414

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ DISTRIBUTION:

PUBLIC LPL2-1 R/F RidsNrrDoriDpr Resource RidsNrrDorllpl2-1 Resource RidsNrrDeEvib Resource RidsAcrsAcnw_MaiiCTR Resource RidsNrrLASFigueroa Resource RidsNrrPMCatawba Resource (hard copy) RidsRgn2MaiiCenter Resource ADAMS Accession No. ML 14295A532 Sincerely,

/RAJ Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

• via memorandum OFFICE DORLILPL2-1

/PM DORLILPL2-1

/LA DE/EVIB/BC DORLILPL2-1/BC NAME GEM iller SFigueroa SRosenberg*

RPascarelli (SRohrer for) DATE 10/23/14 10/23/14 10/15/14 10/30/14 OFFICIAL RECORD COPY