ML20058L611

Safety Evaluation Approving Second ten-year Interval Inservice Insp Request for Relief Re Use of IWA-5250 Requirements Listed in 1992 Edition of ASME Code

ML20058L611
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/13/1993
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20058L610	List: ML20058L605 ML20058L611
References
NUDOCS 9312170106
Download: ML20058L611 (4)

Text

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ENCLOSURE

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20566 0001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE SECOND TEN-YEAR INTERVAL INSERVICE INSPECTION RE00EST FOR RELIEF NO. lWA-5250 ENTERGY OPERATIONS. INC.

ARKANSAS NUCLEAR ONE. UNIT 2 DOCKET NO. 50-363

1.0 INTRODUCTION

Technical Specification 4.0.5 for Arkansas Nuclear One, Unit 2 (ANO-2), states that 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)(1). 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 compensatir.g 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 second 10-year interval comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by 1 reference in 10 CFR 50.55a(b) on the date 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  !

ANO-2, Second 10-Year Inservice Inspection (ISI) Interval is the 1986 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 XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission ft N 0$0$o$6B.

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i in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 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. In a letter i dated August 21, 1992, the licensee, Entergy Operations, Inc., submitted a relief request to use the requirements found in Paragraph IWA-5250(a)(2) of the 1992 Edition of the Code in lieu of the requirements of the 1986 Code.

2.0 EVALUATION AND CONCLUSION The staff, with technical assistance from its contractor, the Idaho National Engineering Laboratory (INEL), has evaluated the information provided by the licensee in support of its proposed alternative. Based on the information provided, the staff adopts the contractor's evaluation and conclusion -

presented in the attached Technical Evaluation Summary. The staff concludes that the licensee's proposed alternative to use Paragraph IWA-5250(a)(2) of the 1992 Edition of the Code will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(1), the licensee's proposed alternative is authorized.

Principal Contributor:

Thomas McLellan, EMCB 1

Date: December 13, 1993

Attachment:

Technical Evaluation Summary l

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ATTACHMENT i

TECHNICAL EVALUATION

SUMMARY

OF THE SECOND TEN-YEAR INTERVAL INSERVICE INSPECTION RE0 VEST FOR RELIEF NO. IWA-5250 ENTERGY OPERATIONS. INC.

l ARKANSAS NUCLEAR ONE. UNIT 2 DOCKET NO. 50-368

1.0 INTRODUCTION

In a letter dated August 21, 1992, the licensee, Entergy Operations, Inc.,

submitted Request for Relief No. IWA-5250, requesting to use the requirements found in Paragraph IWA-5250(a)(2) of the 1992 Edition of the Code in lieu of the requirements of the 1986 Code. The Idaho National Engineering Laboratory *

(INEL) staff has evaluated the information provided by licensee in support of the licensee's proposed alternative contained in Request for Relief No.

IWA-5250 as follows.

2.0 EVALUATION Reouest for Relief No. IWA-5250. Paraaraoh IWA-5250(a)(2). Corrective Measures for leakaae at Bolted Connections Code Requirement: Paragraph IWA-5250(a)(2) states that the source of leakages detected during the conduct of a system pressure test shall be located and evaluated by the Owner. If the leakage occurs at a bolted connection, the bolting shall be removed, visually examined (VT-3) for corrosion, and o evaluated in accordance with IWA-3100.

Licensee's Code Relief Reauest: Relief is requested to use the corrective measures for leakage at bolted connections contained in IWA-5250(a)(2) of the -

1992 Code.

Licensee's Basis for Reauestina Relief: The licensee states that Subsection IWA-5250(a)(2) was added to the Code with the 1986 Edition and then revised in the 1990 Addenda to the 1989 Edition. This requirement has been maintained in the recently published 1992 Edition of the Code. The remainder of Section IWA-5250 has not been changed with subsequent editions and addenda.

A literal interpretation of the wording of the 1986 Edition requires complete removal of all bolts for evaluation if leakage occurs at a bolted connection during system pressure testing. Operating under such an interpretation could lead to unnecessary cooldown of the plant (i.e., should a minor leak be made worse), unnecessary radiation exposure of plant maintenance and inspection personnel, and significant delays in plant startup following a refueling outage should non-degraded bolts be required to be removed. The revised a wording in the 1992 Edition provides relief to the current restrictions in P

. i' l that it limits complete disassembly of a halted connection to cases where degradation of a single bolt has been demonstrated by VT-3 examination.

Licensee's Proposed Alternative Examination: Corrective measures for leakage at bolted connections will be performed to the requirements of Paragraph IWA-5250(a)(2) of the 1992 Edition of the ASME Code as follows:

If leakage occurs at a bolted connection, one of the bolts shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100. The bolt selected shall be the one closest to the source of the leakage.

When the removed bolt has evidence of degradation, all remaining bolting i in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3300.

Evaluation: The correct Wii on, of the Code for the ANO-2, second 10-year ISI interval is the 1986 Edition. The licensee is requesting to use a portion of ,

a subsequent edition of the Code that is not referenced in 10 CFR 50.55a(b).

The 1986 Edition of the Code requires that for leakage at a bolted connection, all bolting be removed and VT-3 examined. These requirements were revised in the 1990 Addenda to the 1989 Code to require the removal and VT-3 examination of only of one bolt, located closest to the leak. If degradation is evident, ,

the remaining bolts are removed and VT-3 examined.

The approach taken in the 1990 Addenda (through the 1992 Edition) of the Code  ;

eliminates the need to completely disassemble bolted connections to determine  !

if degradation has occurred as a result of the leakage. Instead, only one bolt closest to the leakage is evaluated to assess potential degradation. In this manner, the evaluation is focused at the most probable degradation location, and degradation, if present, will be detected. If degradation is evident, the remaining bolts are removed and VT-3 examined. Accordingly, the proposed alternative will provide an acceptable level of quality and safety.

3.0 CONCLUSION

We have reviewed the licensee's submittal and have concluded that the proposed alternative to use Paragraph IWA-5250(a)(2) of the 1992 Edition of the Code will provide an acceptable level of quality and safety and therefore, pursuant to 10 CFR 50.55a(a)(3)(1), the licensee's proposed alternative should be authorized.

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