ML20198P872

Safety Evaluation Accepting Request for Alternative Insp Requirements for Augmented Exam of Reactor Presure Vessels for Zion Nuclear Power Station,Units 1 & 2

ML20198P872
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	01/09/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198P861	List: ML20198P859 ML20198P872
References
NUDOCS 9801220276
Download: ML20198P872 (4)

Text

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. 'p 4 UNITE 3 STATES j NUCLEAR RESULATORY C$MMISSION WASHINGTON, D.C. 3000HOO1

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%, . . . . . f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE SECOND TEN YEAR INTERVAL INSERVICE INSPECTION PLAN PROPOSED ALTERNATIVE TO REACTOR VESSEL AUGMENTED EXAMINATION COMMONWEALTH EDISON COMPANY Zl,ON NUCLEAR STATION. UNITS 1 AND 2 DOCKET NUMBERS: 50 295 AND 50-304

1. INTRODUCTION The Technical Specifbations for Zion Nuclear Station, Units 1 and 2, state that the inservice inspection and testing 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 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(l).

10 CFR 50.55a(a)(3) states that altamatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed attematives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulties whhout 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 ccmponents (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 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) on the date twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of the ASME Code,Section XI, for Zion Nuclear Station, Units 1 and 2, during the second ten-year inservice inspection (ISI) Interval, was the 1980 Edition through Winter 1981 Addenda. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by refereace in 10 CFR 50.55a(b) subject to Gie 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 XI of the ASME Code is not practical for its facility, information shall be submitted to the Corr. mission in support of that determination and a request made for relief from the ASME Code requirement. After evaiuation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose attemative requirements that are determined to be authorized by law, will not endanger life, property, or the 9001220276 990109 PDR

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4 2-common defense and security, and are otherwise in the public interest, giving due consideration i to the burden upon the licenses that could result if the requirements were imposed, j Pursuant to 10 CFR 50.55a(g)(6)(ii)(A) the Commission revoked all previous reliefs granted to 5 licensees for the extent of volumetric examinations of reactor vessel she'l welds, as specified in Section XI, Division 1 of the ASME Boiler and Pressure Vessel Code. The Commission further required that all licensees augment their reacter vessel examination by implementing once, as  ;

part of the inservice inspec' ion interval in effect on September 8,1992, the item B1.10 requirements (examine essentially 100% of the volume of each shell weld) of the 1989 Edition of the ASME Code.

Under 10 CFR 50.55t(g)(6)(ii)(A)(4), licensees may satisfy the requirements for augmented examination of the reactor vessel by performing the ASME Section XI reactor vessel shell wold.

examinations scheduled for implementation during inservice inspection intervals in effect on r

September 8,1992. As a result, the licensee is required to submit both an altemative to 10 CFR 50.55a(g)(6)(ii)(A) and a request for relief per 10 CFR 50.55a(g)(5)(iii), or a proposed attemative per 10 CFR 50.55a(3), for the same wolds when the licensee obtains less than the required coverage (essentially 100%) during the examinations.

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 attemative to the examination requirements that would provide an acceptable level of quality and safety. The licensee may use l the proposed attemative when authorized by the Director of the Office of Nuclear Reactor Regulation, in a (5).Authorization to Use Partial Insp Coverage Requested|letter dated August 13,1997]], Commonwealth Edison Company (Comed or the licensee),

submitted to the staff its attemative to the augmented examination of the reactor vessel shell welds conducted pursuant to 10 CFR 50.55a(g)(6)(ii)(A) for Zion Nuclear Power Station, Units 1 and 2, during the second ten-year interval. The licensee's proposed attemative to examination of

" essentially 100%" of two non-beltline welds in each reactor vessel is a best-effort examination resulting in limited examination coverage of the welds that provide an acceptable level of quality and safety. The staff has reviewed and evaluated the licensee's proposed attemative and the supporting information, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) for Zion Nuclear Power Station, Units 1 and 2.

2.0 Q@CUSSION COMPONENT IDENTIFICATION:

, Code Class: 1

Reference:

Table IWB-25001 Examination Category: B-A p item Numbers: B1.11 and 81.12 l.

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Description:

Limited Volumetric Examination of  ;

Reactor Pressure Vessel Shell Wolds i Units: Zion Nuclear Power Station, Units 1 and 2 l l

Component Numbers: Circumferential RPV Wold #2, Wold #5 i

(for both units)

EXAMINATION REQUIREMENT:

10 CFR 50.55a(g)(6)(ii)(A)(2) states that all hoensees shall augment their reactor vessel examinations by implementing the examination requirements for Reacbr Pressure Vessel (RPV) <

shell welds spcukd la item B1.10 of Examination Category B-A, " Pressure Retaining Wolds in Reactor Vessel," in Table IWB-2500-1 of Subsection lWB of the 1989 Edition of Section XI, Division I, of the ASME Boiler and Pressure Vessel Code, subject to the conditions specified in 50,55a(g)(6)(ii)(A)(3) and (4). For the purpose of this augmented examination, essentially 100%

as used in Table IWB-25001 means more than 90% of the examination volume for each weld.

Additionally,10 CFR 50.55a(g)(6)(ii)(A)(5) requires licensees that are unable to completely satisfy ,

the augmented RPV shell weld examination requirement to submit information to the U.S.

Nuclear Regulatory Commission to support the determination, and propose an altemative to the  ;

examination requirements that would provide an acceptable level of quality and safety.

LICENSEE'S PROPOSED ALTERNATIVE: (As stated)

" Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), Zion Station requests approval of partial  ;

coverage (UT coverage of less than or equal to 90% of the required examination volume)

J on Unit 1 and 2 welds #2 and #5 as an ahomative to the rules of 10 CFR 50.55a(g)(6)(ii)(A)(2), on the basis that the proposed altemative would provide an acceptable level of quality and safety."

LICENSEE'S BASIS FOR REQUESTING ALTERNATIVE:

in each unit, four circumferential welds (Nos. 2, 3, 4, and 5) and four longitudinal welds (Nos. 7, 8,9, and 10) were examined to satisfy the augmented RPV exso,ination requirement. All recordable indications in both units were categorized as subsurface planar flaws which were found to be acceptable under the ASME Code,Section XI,1980 Edition through Wint6r 1981 l Addenda.

All four longitudinal welds and two circumferential welds (Weld Nos. 7, 8, 9,10, 3 and 4) in the beltline region of each reactor vessel were 100% examined by ultrasonics. However, two lc circumferential welds (Weld Nos. 2 and 5) outside the beltline in each reactor vessel received l

- ultrasonic coverage less than 90% due to part geomet:y and interference with other components.

One of the circumferential welds (Weld No. 2) located at the same elevation of the inlet and

- outlet nozzles received a coverage of approximately 74% in each unit due to tapering of nozzle blend radius. The other circumferential weld (Weld No. 5) between the lower shell course and

' the reactor vessel bottom head received a coverage of approximately 81% due to interference from six core support lugs in each unit. The results obtained from volumetric examination of Wolds #2 and #5 adequately represented the structural condition of the subject welds since more than half of each weld length examined, confirmed absence of rejectable service induced flaws.

Further, all six beltline welds in each reactor vessel which are exposed to higher neutron fluence than Welds #2 and #5, received 100% volumetric examination coverage with no rejectable

service induced flaws.

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4, 3.0 EVALUATION The staff has evaluated the attematives proposed by the k 'nsee for the volumetric examination of the abovementioned reactor vessel shell welds in regarc w the following factors:

• physical constraints at each weld that limit required examination coverage;

• maximum extent of volumetric coverage obtained with the existing constraints;

• supplementing inner diameter examination with examination from outside;

• results of augmented vessel examination;

• detect presence of degradation mechanism, if any, from the examination; and

• effect of neutron irradiation on the subject welds.

The licensee performed a best-effor1 examination of the above welds in both reactor vessels from the inside surface. In each reactor vessel, the volumetric coverages of the Upper Shell Weld (Weld #2) was 74% and the Lower Shell Course To Bottom Head Weld (Weld #5) was 81%. The reduced examination coverages of Weld #2 and #5 wore attributed to geometric configuration of Weld #2 and interference from adjacent companents at Weld 35. All other welds specified in item B1.10 of Examination Category B-A, of Units 1 and 2, reactor vessel met the Code examination requirement. The licensee states that in both units of Zion there is insufficient clearance between the reactor vessel wall and concrete which prevents an examination from the outside surface of the vessel to supplement the examination coverage of the subject welds.

The staff noted that all recordable ind5ations found during augmented examination were Code acceptable and further determined that the Upper Shell Weld (Weld #2) and the Lower Shell Course To Bottom Head Weld (Weld #5), being located outside the vessel beltline region, would not be exposed to critical neutron fluence level to adversely affect their fracture toughness. The staff also determined that unacceptable flaws caused due to any degradation mechanism, if present, would have been detected with reasonable confidence with volumetric examination coverages of 74% and 81% for the subject welds. Therefore, the licensee's proposed attemativa provides an acceptable level of quality and safety.

4.0 CONCLUSION

Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the staff evaluated the licensee's proposed attemative for examining reactor vessel shell welds specified in item B1.10 of Examination Category B-A,

" Pressure Retaining Welds in Reactor Vessel,"in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of the ASME Code,Section XI and concluded that the Ecensee has maximized examination coverage for the reactor vessel welds and that service-induced degradation, if present, would have been detected. Thus, the licensee's proposed attemative to augmented reactor vessel examinction provides an acceptable level of quality and safety. Therefore, the licensee's proposed attemative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for Zion Nuclear Power Station, Units 1 and 2, for the augmented reactor vessel examinations that were performed per the requirement in 10 CFR 50.55a(g)(6)(ii)(A)(2) during the third inspection period of the second inservice inspection interval.

Principal contributor: P. Patnaik Date: January 9, 1998