ML20198R995

Safety Evaluation Authorizing Request for Approval of Alternative to Exam Requirement of Reactor Vessel Shell Weld,Per 10CFR50.55a(g)(6)(ii)(A)(5)

ML20198R995
Person / Time
Site:	Catawba
Issue date:	10/30/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198R987	List: ML20198R983 ML20198R995
References
NUDOCS 9711130415
Download: ML20198R995 (5)

Text

. - . . - . - - -. -- _ _ - -

zpwhq p *4 UNITED STATES

" }

4 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PLAN 4

REQUESTS FOR RELIEF NO. 97-01 AND NO. 97-02 DUKE ENERGY CORPORATION CATAWBA NUCLEAR STATION. UNITS 1 AND 2 D.O 3KET NOS. 50-413 AND 50-414

1.0 INTRODUCTION

4 The Technical Specifications for Catawba 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 (ASME Code) and applicable Addenda as required by Title 10 of the Cgde 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).

Section 50.55a(a)(3) states that attematives 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 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 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 l 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) on

, the date 12 months prior to the start of the 120-montn interval, subject to the limitations and modifications listed therein. The applicable edition of the ASME Code,Section XI, for Catawba Nuclear Station, Units 1 and 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.

Enclosure 9711130415 971030 PDR ADOCK 05000413 P PDR

i j- ..

p

- 2-i-

l 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 Commission in support of that determination and a request i

made for relief from the ASME Code requirement.. After evaluation of the determination, j pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and 'may impose '

l altemative requirements that are determined to be authorized by law, will not endanger life,-

i - property, or the common defense and security, and are othenvise in the public interest, giving j- due consideration to the burden upon the licensee that could result if the requirements were

! - imposed on the facility. '

i i

Pursuant to 10 CFR 50.55a(g)(6)(li)(A), the Commission revoked all previous reliefs granted to

[ licensees for the extent of volumetric examinations of reactor vessel shell welds, as specified in e Section XI, Division 1, of the ASME Code'. The Commission further required that alllicensees e

augment their reactor vessel examination by implementing, as part of the ISI interval iri effect i i

on September 8,1992, the Item B1,10 requiremer:ts (examine essentially 100% of the volume -

of each shell weld) of the 1989 Edition of the ASME Code.

Under 10 CFR 50.53a(g)(6)(li)(A)(4), licensees may satisfy the augmented requirements by performing the ASME Section XI reactor vessel shell weld examinations scheduled for ,

implementation during ISI intowals in effect on 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 '

pursuant to 10 CFR 50.55a(g)(5)(iii), or a proposed attemative pursuant to 10 CFR 50.55a(3), ,

for the same welds 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), a licensee who makes a determination that it is 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 altamative to the examination requirements that would provide an acceptable level of quality and safety. The licensee may

. use the proposed attemative when authorized by the Director of the Office of Nuclear Reactor -

- Regulation.

In a letter dated February 18,1997, and a subsequent response on September 2,1997, to the staff's request for additional information, Duke Energy Corporation (DEC, the licensee) submitted to the NRC its attematives to the augmented examination of the reactor vessel shell welds to be conducted pursuant to 10 CFR 50.55a(g)(6)(ii)(A) for Catawba, Units 1 and 2, during the second 10-ye_r interval. The licensee's attemative plan includes examination of all reactor vessel shell welds to the maximum extent practical. The staff has reviewed and evaluated the licensee's request for attemative, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) and the supporting information for Catawba, Units 1 and 2.

h,_ _ _ __ m .-wi- --,a-i-me-y .% +..%,4 . -.,.. ,,,, c. ., . _4- , _-, , , . . ,, . _ , . - -.--y. ,-,.- .. -s_i.p 7

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

8 w

p - 4 I

i - 2.0 DISCUSSION i- - Examination Requirement:

Section 50.55t.(g)(6)(ii)(A)(2) states that 'all licensees shall augment their reactor vessel examinations by implementing the examination requirements for reactor pressure vessel (RPV) l '

shell welds specified in 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, l Division I, of the ASME Code, subject to the conditions specifM in 10 CFR -

i 50.55a(g)(6)(li)(A)(3) and (4). For the purpose of this augmented examination, essentially 100% as used in Table IWB-2500-1 means more than 90% of the examination volume for eacn i weld. Additionally,10 CFR 50.55a(g)(6)(ii)(A)(5) requires licensees that are unable to i completely satisfy the augmented RPV shell weld examination requirement to submit 4 --information to the NRC to support the determination, and propose an altemative to the examination requirements that would provide an acceptable level of quality and safety, i

! Licensee's Request For Relief:

f Reactor vessel shell welds specified in item B1.10 of Examination Category B-A of the

! ASME Code,Section XI,' 1989 Edition that did not receive " essentially 100%" examination.

Unit 1 Reactor Vessel Head-to-Shell Circumferential Weld ID Number item Number 1RPV-WO3 B01.001.001 Shell-to-Nozzle Belt Circumferential Weld ID Number item Number 1RPV-WOG B01.001.004 Unit 2 Reactor Vessel Lower Head-to-Shell Circumferential Weld ID Number item Number 2RPV-101-141 B01.011.001 Lower Shell Longitudinal Seams ID.fstmket item Number 2RPV-101-142A B01.012.007 2RPV-101-142B B01.012.008 2RPV-101-142C B01.012/09

~

4 Licensee's Basis For Requesting Alternative:

During the ultrasonic examination of the welds, the minimum 90% coverage requirement of ASME Section XI,1980 Edition through Winter 1981 Addenda, clarified by Code Case N-460, could not be obtained due to part geometry and physical barriers. A combination of multiple angles and UT techniques was used to obtain the maximum coverage possible. Nthough the coverage requirements of ASME Section XI could not be met, the amount of coverage obtained for these examinations provides an acceptable level of quality and integrity. Based on these evaluations, the limited coverage will in no way endanger the health and safety of the general public.

These welds were examined to the maximum extent practical in accordance with ASME Section V, Adicle 4,1980 Edition with Winter 1981 Addenda and the additional requirements of Regulatory Guide 1.150.

No additional examinations will be required.

Licensee's Attemate Examinations:

The licensee states that the use of radiography as an attemate volumetric examination method ,

is not practical due to component thickness and configurations. Other restrictions making radiography impractical, are physical barriers prohibiting access for placement of source, film, number bands, etc.

The licensee will contir.ue to use the most current ultrasonic techniques available for future examination of the welds for which relief is being requested. The licensee believes that the limited examination is the best available and will continue to perform an ultrasonic examiration of all reactor vessel welds to the maximum extent practical in accordanc with the requirements of ASME Section V, Article 4,198g Edition and Regulatory Guide 1.150, Revisien 1, Appendix A.

3.0 EVALUATION Then staff has avaluated the attematives proposed by the licensee for the volumetric examination of the above-mentioned reactor vessel shell welds in regard to the following

- factors, o Physical constraints at each weld that limits the examination coverage o- Maximum extent of volumetric coverage obtained with the existing constraints o Supplementing inner diameter examination with examination from outside o Results of previous vessel examinations o Detect presence of degradation mechanism, if any, from the examination o Effect of neutron irradiation on the subject welds as of the secor:d 10-year inspection interval

c .a

~.

> l

) x i --

The licensee performed a best-effort examination of the above welds in both reactor vessels 3

from the inside surface._ in the Unit i reactor vessel, a volumetric coverage of 44% and 48% for the lower head-to-shell weld and the shell-to-nozzle belt weld respectively, was obtained. . The -

limited scan was due to geometric configuration of the welds. All other welds specified in item B1.10 of Examination Category B A, of Unit 1 reactor vessel met the Code examination requirement. In the Unit 2 reactor vessel, the lower head-to-shell weld received a volumetric coverage of 57% and three lower shell longitudinal seams received volumetric coverage of 81%, each due to physical obstructions from the core guide lugs. All other welds specified in item 81.10 of Examination Category B-A, of Unit 2 reactor vessel met the Code examination requirement. In response to staffs request for additional information dated April 17,1997, the licensee stated that in both units there is insufficient clearance between the reactor vessel wall and concrete to perform an examination from the outside surface for the shell-to-nozzle belt weld in Unit 1 and three lower shell longitudinal seams in Unit 2. However, he volumetric examination coverage of the lower head-to-shell weld in both units can be li.c Nwd by an outer diameter examination with high man-rem penalty.

The licensee's response to the staffs request for additional information states that the preservice inspection of the welds for which an attemative is being requested did not indicate any recordable flaw in the unexamined volume of each weld. Furthermore, the shell-to-nozzle belt weld and the lower head to-shell weld, being located outside the vessel beltline region, have not experienced neutron embrittlement, which would have adversely affected the fracture toughness of the welds and the heat-affected zones. With the number of transients below the design basis, the probability of origination and growth of a flaw to unacceptable dimensions is extremely small within the second 10-year interval. Moreover, the extent of volumetric coverage should have detected the presence of such a flaw. The three lower shelllongitudinal welds of Unit 2, located within the vessel beltline region received volumetric examination coverage of 81%. The staff has determined that flaws of unacceptable dimensions caused due to any degradation mechanism, if present, would have been detected with reasonable confidence with such volumetric coverage. Therefore, the licensee's proposed attemative provides an acceptable level of quality and safety,

4.0 CONCLUSION

The staff has reviewed the licensee's submittals and concludes that the licensee has maximized examination coverage for the reactor vessel welds and that service-induced degradation, if -

present, would have been detected. Compliance with the code results in hardship or unusual difficulty without a compensating increase in the level of quality and safety in that the licensee's proposed alternative contained in Requests for Relief No. 97-01 and 97e? provides an acceptable level of quality and safety. Therefore, the licensee's proposed attemative is authorized pursuant to the provisions of 10 CFR 50.55a(g)(6)(ii)(A)(5) and 50.55a(3)(ii) for  !

Catawba, Units 1 and 2, during the second 10-year interval. '

l Principal Contributor: Prakash Patnaik I

-- Date: October 30, 1997 I

l l