ML20134P366

SER Accepting Test & Technical Evaluations Performed for Reactor Vessel Shell Welds,Per 10CFR50.55a(g)(6)(ii)(A)(5)

ML20134P366
Person / Time
Site:	Hatch
Issue date:	02/21/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20134P348	List: ML20134P346 ML20134P366
References
NUDOCS 9702250388
Download: ML20134P366 (4)

Text

, .

@ utg p 1 UNITED STATES g j f

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20566-0001 o

+9 . . . . . ,o I

2 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE AUGMENTED EXAMINATION OF THE REACTOR VESSEL

GEORGIA POWER COMPANY. ET AL.

EDWIN I. HATCH NUCLEAR POWER PLANT. UNIT 2 DOCKET NO. 50-366

1.0 INTRODUCTION

The Technical Specifications for Edwin I. Hatch Nuclear Plant, Unit 2, state that 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 (ASME 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)(1). Section 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. 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 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 10-year interval and subsequent intervals 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) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for the Edwin I. Hatch Nuclear Plant, Unit 2, third 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 50.5.5a(b) subject to the limitations and modifications listed therein and subject to Commission approval.

Enclosure 9702250388 970221 DR ADOCK0500g6

-. ~ . . - - . . - - . . . -. .-- .. - - . . - . _ . - - . - - .

, , a i

i l l Pursuant to 10 CFR 50.55a(g)(5), if a licensee determines that conformance d

with an examination requirement of Section XI 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 i Code requirement. After evaluation of the determination, pursuant to i 10 CFR 50.55a(g)(6)(1), 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 due consideration to the burden upon the licensee that could result if the requirements were imposed.

l Additionally, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), licensees that make a

. determination that they are unable to completely satisfy the requirements for

! the augmented reactor vessel shell weld examination specified in 10 CFR 50.55a(g)(6)(ii)(A) shall submit information to the Commission to support the determination and shall propose an alternative to the examination requirements that would provide an acceptable level of quality and safety. Licensees may use the proposed alternative when authorized by the Director of the Office of Nuclear Reactor Regulation.

In its letter dated October 29, 1996, Georgia Power Company (licensee),

submitted to the NRC its results of the augmented examination of the reactor vessel shell welds conducted in accordance with 10 CFR 50.55a(g)(6)(ii)(A) for Hatch Unit 2 during the fall 1995 outage. The licensee requested relief from examining essentially 100 percent of one circumferential shell weld in the reactor vessel due to physical limitations.

2.0 DISCUSSION Examination Requirement:

ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item l Numbers Bl.10 requires volumetric examination of essentially 100% of weld length of the reactor vessel circumferential and longitudinal shell welds as detailed in Figures IWB-2500-1 and IWB-2500-2.

Section 50.55a(g)(6)(ii)(A)(2) states that for the purpose of this augmented examination, " essentially 100%" as used in Table IWB-2500-1 means more than 90 percent of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or part geometry.

< d i

4 Licensee's Code Relief Request:

? Hatch Unit 2 reactor pressure vessel (RPV) upper shell to upper intermediate shell circumferential weld identified as weld 2C-2, received an 83.7%

, volumetric examination coverage in lieu of essentially 100% as required by the l Code due to interference from six stabilizer brackets on the reactor vessel  ;

j adjacent to the subject weld.

Licensee's Basis for Relief:

i The licensee conducted a study of the volumetric examination of reactor vessel shell welds performed from inside diameter (ID) surface and the outside i diameter (00) surface to determine the extent of volumetric coverage of welds.

~

The licensee concluded that the examination performed from the OD surface provided significantly greater volumetric coverage for most welds than that  :

conducted from the ID surface. Furthermore, it was determined that all longitudinal welds and three out of four circumferential welds would meet the Code-required examination volume for the reactor vessel shell welds specified in Item B1.10 of Examination Category B-A. The only circumferential weld that could not be examined to the Code-required volume was the upper shell to the intermediate shell weld identified as weld 2C-2, due to interference from from six stabilizer brackets adjacent to the weld. However, the percentage that was examined amounted to 83.7%. Hence, it was prudent to perform an OD examination rather than an ID examination of the reactor vessel shell welds in regard to the augmented examination of the subject welds.

Licensee's Alternative Examination:

The licensee performed a combination of automated and manual ultrasonic examination of reactor vessel shell welds specified in Item Bl.10 of Examination Category B-A from the external surface of the vessel and performed a technical evaluation of the integrity of shell welds.  ;

3.0 EVALUATION The licensee conducted the augmented examination of the RPV shell welds specified in Item Bl.10 of Examination Category B-A to the maximum extent possible by examining from outside surface of the vessel and substantiated the unexamined volume of circumferential weld 2C-2 with a technical evaluation of weld integrity by the BWRVIP. The examination of the weld was limited due to interference from six stabilizer brackets permanently attached on the 00 surface of the vessel. If the examination was performed from inside the vessel, the weld 2C-2 would meet the volumetric requirement of the Code but the other critical welds would not meet the percentage requirements. Having conducted an extensive 00 examination using automated and manual techniques, if the licensee were to perform another examination from inside the vessel for weld 2C-2 to satisfy the Code requirement, a significant burden would be placed on the licensee in terms of procurement of ultrasonic equipment, man-rem dose, and lost time without a compensating increase in safety. The staff

believes that if a pattern of degradation does exist in the subject weld, the volumetric examination of 83.7 percent of the weld would have detected this.

Furthermore, the likelihood of a defect existing in the unexamined volume is extremely small. Therefore, the tests already conducted and the evaluation performed by the licensee provides an acceptable level of quality and safety.

4.0 CONCLUSION

The staff has reviewed and evaluated the licensee's request for an alternative to the examination requirement of reactor vessel shell weld 2C-2, pursuant to the provisions of 10 CFR 50.55a(g)(6)(ii)(A)(5). Due to physical interference of permanent attachments on the vessel, the requirement of the Code for the subject weld is impractical for the examination already performed and places significant burden on the licensee if the Code requirements were imposed. The tests and technical evaluations performed for the reactor vessel shell welds provide an acceptable level of quality and safety and therefore, the licensee's alternative is authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5).

Principal Contributor: P. Patnaik Date: February 21, 1997 l