ML20206M463

Safety Evaluation Supporting Alternative Proposed by Licensee to Perform Ultrasonic Exam on Inner Surface of Nozzle to safe-end Weld

ML20206M463
Person / Time
Site:	Millstone
Issue date:	05/11/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20206M454	List: ML20206M450 ML20206M463
References
NUDOCS 9905140267
Download: ML20206M463 (3)

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. SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF REQUEST FOR RELIEF NO. IR-9. REVISION 2 FOR NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION. UNIT NO. 3 DOCKET NO. 50-423  :

1.0 INTRODUCTION

The Technical Specifications for Millstone Unit 3, 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 (B&PV)

Code and applicable addenda as required by Title 10 of the Code of Federal Reaulations (10 >

CFR), Part 50, Section 50.55a. Pursuant to 10 CFR 50.55a(a)(3), altematives to the l requirements of 10 CFR 50.55a(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 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) shall meet the requirements, except the design and access provisions and the pre-service 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 i 10-year inservice inspection (ISI) 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.

In a letter dated December 21,1998, Northeast Nuclear Energy Company (the licensee, NNECO), submitted " Request for Relief No. IR-9, Revision 2, for the Millstone Nuclear Power Station, Unit No. 3." The licensee sought relief from the requirements'of ASME Code,  !

Section XI for eight nozzle to safe-end welds on the reactor vessel of Millstone Nuclear Power l

Station, Unit No. 3. Specifically, the ASME Code requires that surface examination is  :

performed on the outside diameter (OD) of the vessel nozzle to safe-end welds while NNECO I is' proposing to perform full volumetric examination on those welds in order to reduce radiation  !

exposure to plant personnel.  !

ENCLOSURE 9905140267 990511 gDR ADOCK 05000423  !

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e 2.0 EVALUATION The information provided by NNECO in support of its request for relief from Code requirements has been evaluated and the basis for disposition is documented below. The Code of record i for the Millstone Nuclear Power Station, Unit No. 3, first 10-year ISI interval, which began April 23,1986, is the 1983 Edition through Summer 1983 Addenda of Section XI of the ASME {

1 Boiler and Pressure Vessel Code. j i

I 2.1 Reauest for Relief for the Nozzle to Safe-End W' elds Code Reauirement: Table IWB-2500-1, Examination Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles, item No. B5.10 of Section XI of the ASME B&PV Code,1983 Edition, with Addenda through the Summer of 1983, requires that a volumetric and surface examination be performed on nozzle to safe-end welds greater than 4-inch nominal pipe size (nps).

System / Components for Which Relief is reauested:

Eight reactor vessel nozzle to safe-end welds greater than 4-inch nps.

1 Licensee's Proposed Altemative: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing the ASME Code-required surface examination of the reactor vessel nozzle to safe-end welds greater than 4-inch nps. The licensee stated: l A. The subject welds received a full volume volumetric examination from the ID surface in cccordance with ASME Section XI and Section IWA 2240.

This examination con:isted of 40 degree refracted longitudinal wave search units manipulated and scanned in four directions to obtain full volumetric coverage of the weld and % inch of base material on each side of the OD weld fusion lines. This altemative volumetric examination was demonstrated prior to performing the examinations. The demonstration included the ability to detect OD flaws with the actual procedure and instrument used in the proposed examination.

B. In-service System Leakage Tests will be performed per Code Case N 498-1 as granted by NRC correspondence dated January 18,1995.

Licensee' Basis for Proposed Altemative (as stated):

The Millstone Unit 3 reactor vesselis a four-loop PWR. There are eight nozzle to safe-end welds for item B5.10, in the in-service inspection program. The purpose of this relief is to perform a full volumetric ultrasonic examination in lieu of a surface examination in order to reduce exposure. The reactor vessel nozzles have a permanent mirror type insulation installed which covers the OD surface of these welds. The exposure received from the removal and reinstallation of this insulation and performing the surface examination is estimated to be 2.5 man-rem per weld. The ultrasonic examination of these

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welds will be performed during the third inspection period from the ID surface using the remote immersion method.- NNECO believes that any failure of the weld will be induced from the ID surface and that the ultrasonic examination from

' this surface shall be sufficient to detect any indications.

Licensee's Justification for Grantina Relief (as stated): j NNECO believes that the proposed ultrasonic examination will prove to be adequate to detect flaws in the listed items. NNECO believes that the ultrasonic exam will assist in reducing the exposure rates, without losing the ability to detect flaws in the reactor vessel nozzle to safe-end and safe-end to pipe welds.

2.2 Staff Evaluation:

' The Code requires surface examination of the outer surface of the reactor vessel nozzle to safe- l end welds. The licensee stated that the reactor vessel nozzles have a permanent mirror type j insulation installed over the OD surface of these welds. The radiation exposure that the plant I personnel would receive during the removal of the insulation and performing surface examination is estimated to be 2.5 man-rem per weld. As an altemative, the licensee proposes j to perform ultrasonic examination that will be performed from the ID surfaces of the weld using a j remote immersion method. This method will detect defects that may be present on the ID l

surfaces of the welds where most likely a flaw can be initiated. The staff finds the licensee has  ;

provided an acceptable alternative to the requirements of the ASME Code,Section XI, Table  ;

IWB-2500 i

3.0 CONCLUSION

The NRC staff evaluated the licensee's submittal and concluded that pursuant to I 10 CFR 50.55a(a)(3)(ii) that the Code requirement to perform a surface examination on the OD of the vessel nozzle to safe-end welds would result in a hardship upon the licensee and that the proposed alternative to the requirements of the ASME Code,Section XI is authorized. The staff finds that performing surface examinations as required by the Code would result in hardship upon the licensee because the licensee would have to remove the insulation around the reactor nozzle to safe-end welds. Removing the insulatic in order to perform the Code required examination would result in additional radiation exposure of plant personnel without a compensating increase in the level of quality and safety.' As an alternative, the licensee proposes to perform ultrasonic examination on the inner surface of the nozzle to safe-end weld.

The staff finds this alternative to be acceptable. The staff, therefore, concludes that authorization of the licensee alternative would provide an acceptable level of quality and safety.

Pursuant to 10 CFR 50.55a(a)(3)(ii), the alternative is authorized. The relief granted is authorized by law and 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.

Principal Contributor: G. Georgiev Date: May 11, 1999