ML20211A979

Safety Evaluation Granting Licensee Request for Approval of Proposed Relief from Volumetric Exam Requirements of ASME B&PV Code,Section Xi,For Plant,Unit 2

ML20211A979
Person / Time
Site:	Catawba
Issue date:	08/20/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20211A969	List: ML20211A964 ML20211A979
References
NUDOCS 9908240186
Download: ML20211A979 (4)

Text

.

y u%q

[ k UNITED STATLS g j

't NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055H001

...~

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PLAN REQUEST FOR REllEF NO. 99 01 CATAWBA NUCLEAR STATION. UNIT 2 DUKE ENERGY CORPORATION DOCKET NUMBER: 50-414

1.0 INTRODUCTION

The inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by Title 10 of the Code of Federal Reaulations (10 CFR) Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulties 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) 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 Inservice 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 ten-year interval and subsequent intemals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) on the date twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. Accordingly, the applicable edition of the ASME Code,Section XI, for Catawba Nuclear Station, Unit 2, during the second 10-year inservice inspection (ISI) interval, is the 1989 Edition. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.

Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section al of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the deterrrtination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving 9908240186 990820 _

PDR ADOCK 05000414 p PDR _

-2 due consideration to the burden upon the licensee that could result if the requirements were imposed, in a letter dated January 18,1999, Duke Energy Corporation (the licensee) requested relief from the volumetric examination requirement of the ASME Code,Section XI, during the end-of-Cycle-9 outage of Catawba Nuclear Station, Unit 2. The licensee stated that the volumetric examination was restricted at the Pressurizer nozzles and at the Residual Heat Removat Heat Exchanger (RHR HX) nozzles due to configuration. Following are details of the staff's review of this request.

2.0 QLSCUSSION 2.1 System / Components for Which Relief is Requested item Number Component Examined B03.120.002 Pressurizer Spray nozzle inner Radius B03.120.006 Pressurizer Safety nozzle Inner Radius C01.010.007 RHR HX (2A) Shell to Flange Weld CO2.021.006 RHR HX (2A) Inlet nozzle to Shell Weld CO2.021.007 RHR HX (2A) Outlet nozzle to Shell Weld 2.2 Code Requirements ASME Section XI 1989 Edition, Table IWB-2500-1, Examination Category B-D, Full Penetration Welds of Nozzles in Vessels, item number B3.120; Table IWC-2500-1, Examination j Category C-A, Pressure Retaining Welds in Pressure Vessels, item number C1.10; and  !

Examination Category C-B, Pressure Retaining Nozzle Welds in Vessels, item number C2.21 )

require a volumetric examination of essentially 100 percent of the weld volume. Duke Energy Corporation has adopted Code Case N-460 which defines " essentially 100%" as greater than ,

90 percent coverage.

2.3 Licensee's Request for Relief The licensee requested relief from meeting the volumetric examination requirement of the ASME Code,Section XI,1989 Edition for the following components:

4 pressurizer spray and safely nozzle inner radius in Figure IWB-2500-7 (b) j residual heat removal heat exchanger 2A shell to flange weld in Figure IWC-2500-1(a) residual heat removal heat exchanger 2A inlet and outlet nozzle to shell welds in Figure IWC-2500-4 l

l

, During the ultrasonic examination of the pressurizer spray nozzle inner radius 2PZR-W2 (ltem No. B03.120.002) cnd pressurizer safety nozzle inner radius 2PZR-W4C (item number B03.120.006) the Code-required coverage of the examination volume could nct be obtained. Limitations caused by the ratio of the nozzle outer diameter (OD) to the vessel thickness resulted in 66.14 percent coverage of the required volume. Nozzle inner radii were examined with the ultrasonic method to the extent practical from the vessel wall.

During the radiographic examination of the residual heat removal heat exchanger 2A shell-to-flange weld 2ARHRHX-5-9 (item no. C01.010.007), the Code-required volumetric coverage of the examination volume could not be obtained. The radiographic examination coverage was limited to 62 percent due to the shell-to-flange configuration and internal divider plates.

During the ultrasonic examination of the residual heat removal heat exchanger 2A inlet nozzle to shell weld 2ARHRHX-5-A (item number CO2.021.006), and residual heat removal neat exchanger 2A outlet nozzle-to-shell weld 2ARHRHX-5-B (item number C02.021.007), the Code-required coverage of the examination volume could not be obtained since the welds could not be scanned from both sides due to configuration. The maximum volumetric examination coverage obtained for these welds was limited to 62 percent.

No additional examinations have been planned by the licensee during the current interval for the inner radii 2PZR-W2 and W4C of the pressurizer spray and the safety nozzles and the welds 2ARHRHX-5-A, B and 9 on the RHR heat exchanger.

3.0 EVALUATION The staff has evaluated the limitations to volumetric examination of the inner radii of the pressurizer spray nozzle and the safety nozzle and of the welds on the RHR heat exchanger.

The staff has determined that it is impractical to obtain the Code-required volumetric examination coverage due to the configuration of the component being examined.

Class 1 and 2 components are designed with welded joints such as nozzle-to-shell, shell-to-flange which obstruct all or part of the required examination.

The licensee has scanned the inner radii and the welds on the above components using ultrasonic transducers of multiple angles, extended Vee paths and other techniques to obtain maximum volumetric coverage. The licensee's best-effort examination has resulted in volumetric coverage ranging from 62 percent to 66 percent. To obtain the Code-required examination coverage, modification and/or replacement of the component would be required.

Moreover, the extent of examinations already performed on the subject welds and the inner radii would detect generic degradation if it existed, and therefore, does provide reasonable assurance of structuralintegrity of the components.

4.0 CONCLUSION

The staff has determined that the Code requirement to perform essentially 100 percent volumetric examination of the welds and the inner radii in the licensee's Relief Request 99-01 is impractical due to component configuration, it would be a significant burden on the licensee to redesign and/or replace components in order to comply with the Code requirements. Moreover, the completed examinations provide reasonable assurance of structuralintegrity of the

components. The relief granted is authorized by law, will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration

. to the burden upon the licensee that could result if the requirements were imposed on the.

facility. Therefore, the request for relief is authorized pursuant to 10 CFR 50.55a(g)(6)(i) for Catawba Nuclear Station, Unit 2, during the second ten-year inspection interval.

Principal Contributor: Prakash Patnaik Date: August 20,1999 I

1 f

l l

a. l