ML20217E349

Safety Evaluation Authorizing Licensee Proposed Alternative for Current Interval Insp Program Plan

ML20217E349
Person / Time
Site:	Summer
Issue date:	09/22/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20217E334	List: ML20217E330 ML20217E349
References
NUDOCS 9710060464
Download: ML20217E349 (7)

Text

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

a Ct00 1 9

g* k UNITED STATES

• ]'

NUCLEAR REGULATORY COMMISSION WASHINGTON. O.C. soseHooi k . . . . . /j ENCLOSURE 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ,

OF THE  !

SECOND 10 YEAR INTERVAL INSERVICE IFSPECTION PR(XiRAM PLAN INTERIM REQUEST FOR 1ELIEF SOUTH CAROLINA CTRIC & GAS CO.

VIRGIL C. SUMME R NUCLEAR STATION. UNIT 1 DOCKE1 NUMBER: 50 395

1.0 INTRODUCTION

The Technical Specifications (TS) for V. C. Summer Nuclear Station. Unit 1.

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 10 CFR 50.55a(g) except where specific written relief has been 50.55a(g)(6)(1), 10 CFR granted50.55a(a) by(the Commission

3) states pursuanttotothe that alternatives 10 CFR requirements of paragraph (g) may be used, when authorized by the NRC. if (1) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with tie specified requirements would result in hardship or unusual difficultly 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 requiremerts, except the design and access provisions and the Code.Section XI. " Rules pre service for Inservice examination requirements, Inspection setPower of Nuclear forth inPlant the ASME Components." to the extent practical within the limitations of design.

and materials of construction of the components. The regulations geometry, require t hat inservice examination of components and system pre.w re tests conducted during the first 10-year interval and subsecuent it tervcis co iy with the requirements in the latest edition and aodenca of SecHr. XI o@f tar 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 V.

C. Summer Nuclear Station. Unit 1. second 10-year inservice inspection (ISI) intervai is the 1989 Edition.

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 Comission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation ' the determination, pursuant to 10 CFR 50.55a(g)(6)(ih the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law, will not

u 18sM Pe h G PDR

O 4

g 2

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 dated February 25. 1997 South Carolina Electric & Gas Co.

(licensee) submitted its request to use Code Case N 566 in lieu of the Code recuirements. However. Code Case N 566 is currently under review by the staff anc has not been approved for use by reference in Regulatory Guide 1.147, inservice inspection Code Case Acceptabi1ity. ASME Sectton XI. 01viston 1 nor authorized as an alternative pursuant to 10 CFR 50.55a(a) P :(1). Because of a-1997 fall refuelin information (RAI) gregarding outage. in response to an NRC request for additional the use of Code Case N 566, the licensee submitted an interim alternative to the requirements of lWA-5250(a)(2) in a letter dated July 30, 1997, for V. C. Summer Nuclear Station. Unit 1.

2.0 EVALUATION

{ The staff, with technical assistance from its contractor, the Idaba National Engineering and Environmental Laboratory (INEEL). has evaluated the

' information provided by the licensee in support of its interim alternative to the requirements of IWA 5250(a)(2) in a letter dated July 30, 1997, for V. C.

a Summer Nuclear Station. Unit 1. Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the j enclosed Technical Letter Report (TLR),

! Interim Re 5250(a)(2),quest for Relief The Code.Section XI. !WA-i requires that ifIWA b250(a)(2):

leakage occurs at a bolted connection, the bolting shall be removed. VT-3 visually examn.ed for corrosion. and evaluated in accordance with IWA-3100. Pursuant to 10 CFR 50.55a(a)(3)(1). the licensce

] proposed the following alternative to the requirements of IWA 5250(a)(2):

'The source of all leakage at bolted connections detected by VT 2 examination during a system pressure test shall be evaluated to

! determine the susceptibility of the bolting to corrosion and potential failure. This evaluation will consider the following variables at a j minimum:

1. Location of leakage

2. History of leakage

3. Fastener materials

4. Evidence of corrosion, with the connection assembled

5. Corrosiveness of the process fluid

6. History and studies of similar fastener material in a similar environment

7. Other components in the vicinity that may be degraded due to the leakage.

"When the evaluation of thG above variables is concluded and if the evaluation determines that the leaking condition has not degraded the

i r

! 3-I fasteners. then no further action is necessar However, reasonable t attempts to stop the leakage shall be taken. y.

i "If the evaluation of the variables above indicates the need for further ,

evaluation, or no evaluation is performed, then a bolt closest to the source of leakage shall be removed. The bolt will receive a VT 1 examination and be evaluated for corrosion in accordance with IWA- -

3100(a) and dis >ositioned in accordance with IWB 3140. If the leakage '

is identified wien the bolted connection is in service, and the information in the evaluation is supportive. the removal of the bolt for VT 1 examination may be deferred to the next refueling outage. When the removed bolting shows ovidence of rejectable degradation all remaining bolts shall be removed and receive a-VT-1 examination and evaluation in accordance with IWB 3140."

In accordance with lWA 5250(a)(2) if leakage occurs at a bolted connection. '

the bolting must be removed. VT 3 visually examined for corrosion, and ,

evaluated in accordance with IWA 3100. In lieu of this requirement, the (

lic3nsee has to corrosion. proposed Based ontothe evaluate the bolting items included to evaluation in the determine process, its susceptibility the staff concludes that the evaluation proposed by the licensee provides a sound  :

engineering approach. In addition if the initial evaluation indicates the  !

need for a more detailed analysis, the bolt closest to the source of the .

leaka 3100(ge a). will be removed. VT-1 examined, and evaluated in accordance with IWA-i Based on the bolting evaluation criteria contained in the interim relief request, the staff concludes that the licensee's proposed alternative to the '

requirements of IWA 5250(a)(2) is a conservative and technically sound engineering approach. As a result, significant patterns of degradation will -

be detected, providing an acceptable level of quality and safety.

3.0 CQN.G M 108 i The staff evaluated the licensee's Interim Request for Relief IWA-5250(a)(2) and concludes that the proposed alternative contained in the request provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(1) for the current interval or until Code Case N 566 is evaluated and authorized in an NRC Safety Evaluation. ,

d )

8 W

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

, ENCLOSURE 2 TECHNICAL LETTER REPORT ON THE SECOND 10 YEAR INTERVAL INSERVICE INSPECTION 1

INTEftlM REQUEST FOR REL!EF 1

EQR SOUTH CAROLINA ELECTRIQ,& GAS CO.

VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 .

4 QQCKET NUMBER! 50 395

1.0 INTRODUCTION

By letter dated February 25,1997, the licensee, South Corotiria Electric & Gas Co.

(SCE&G), proposed an alter.iative to the requireneents of the ASME Coda, Cection XI, for the V. C. Summer Nuclear Station, Unit 1, s6cond 10 year inservice Inspection (ISI) Interval. The alterna'ive proposed is to use Codo Case N 566 in !!ou 4

4 of the requirements of IWA 5250(a)(2), Code Case N 566 is currently under review by the Nuclear Regulatory Commies!on (NRC) staff and h48 not been approved for i

use by rafercnce in Regulatory Guldr 1.147, /nservice /nspection Code Case Acceptabilitye ASME Section XI, DMsion !. In respo'sso to an NRC request for saditional Information (RAI) regarding the use of thle Code Case, ths 2cesisee submitted an ir.terim request for relief from the requirewnts of IWA 5kb0fa)(2) in s letter dated July 30,1997. The Idaho National Engineeririg and Environmental Laboratosy (IN8 EEL) staff has evaluated the subject requeat for relief in the following i

section.

2.0 EyALUATION The Code of record for tha V, C. Summer, Unit 1, second 10 year ISIinterval, wh!ch began January 1994,!s the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code. The inforrnat'on provided iv,' tha licensee in support of the

l

• d 1 2 1 l request for relief from Code requirements has been evaluated and the basis for disposition is documented b6fow. ,

i  :

Interim Raouest for Reflef IWA 5250(a)(2). Corrective Actions for Bolted Connections l

Code Raoulrement: Section XI, lWA 5250(a)(2), requires that if leakage occurs at a bolted connection, the bolting shall be removed, VT 3 visually examined for

corrosion, and evaluated in accordance with lWA 3100.

2 l

Licensee's Pronosed Alternative: Pursuant to 10 CFR 50.55a(a)(3)(l), the licensee proposed the following alternative to the requirements of IWA 5250(a)(2):

l "The source nf allleakage at bolted connections detected by VT 2 examination during system pressury 6tt shall be evaluated to determine the susceptibility of the

' bolting to corrosion artti paiential failure. This evaluation will consider the following variables at a minimurm

, 1. Location of leakage

2. History of leakage

3. Fastener rraterials I 4. Evidence of corrosion, with the connection assembled

5. Corrosiveness of the process fluid

6. History and studies of similar fastener materialln a similar environment

7. Other components in the vicinity that may be degraded due to the leakage.

"When the evaluation of the above variables is concluded and if the evaluation determines that the leaking condition has not degraded the fasteners, then no further action is necessary. However, reasonable attempts to stop the leakage shall be taken.

• lf the evaluation of the variables above indicates the need for further evaluation, or no evaluation is performed, then a bolt closest to the source of leakage shall be removed. The bolt will receive a VT 1 examination and be evaluated for corrosion in accordance with IWA 3100(a) and dispositioned in accordance with IWB 3140, if the leakage is identified when the bolted connection is in service, and the information in the evaluation is supportive, the removal of the bolt for VT 1 1

5 4

y ,. ,, -. -- . . . , - - - , . . -. ~ .# , ,, _,-_ --..mm . .. _ . - . .- - ._ ,,,, _.,.,, _ - - ,. .

.o' 3-1 examination may be deferred to the next refueling outage. When the removed bolting shows evidence of rejectionable degradation, all remaining bolts shall be removed and receive a VT.1 examination and evaluation in accordance with IWB 3140.*

Licensee's Basis for Reauestino Relief (as stated):

1 "Some of the problems associated with the current requirements of IWA 5250(aH2)  ;

are summerized as follows: '

1.

lWA 3100 does not provide an acceptable standard for a VT 3 bolt inspection.

2. The requirement calls for bolt removal without rogard to the size of the leakage.

3. The requirement increases the radiological dose to workers for leaks that are often not a challenge to operational nor structurallimits.

4. Bolts sometimes cannot be removed without damaging the bolt or cannot be removed due to the component configuration.

5. It is not a requirement of the Code that the Owner must stop the leakage and inspection of the bolting is not necessarily going to stop the leak.

6. Removing one bolt at a time,if allowed by system conditions, may actually increase the leakage.

7. In many cases, implementation of the requirement would cause the plant an unnecessary transient or delay startup.

• In addition to the problems associated with the requirements of IWA 5250(a)(2), a Special Task Group of the ASME Committee has concluded that the Code does allow the acceptance of leakage by the analytical evaluation methods of IWB 3142.4, and that the actions required by IWA 5250 should not preclude this acceptance. Also, the Working Group Pressure Testing concluded that the system integrity of a bolted connection is not necessarily compromised by leakage and recommended the approval of Code Case N 566.

"This interim relief request is more prescriptive and more conservative than the Code Case, it also addresses many of the implementation and radiological hardships associated with IWA 5250(a)(2) and yet maintains the conclusion of the ASME Committee by assuring that a proper evaluation of the connection and/or the bolting is performed. The joint evaluation must consider specific factors which, if indicative of degradation, must be dispositioned in accordance with IWB 3140 of Section XI.

Due to the fact that this engineering evaluation is more comprehensive than the simple bolt inspection currently required by IWA 5250, coupled with the benefit that these alternative requirements ensure structuralintegrity is maintained, and reduces the operational, maintenance, and radiological hardships of the current requirements, this relief request should be considered as an acceptable alternative in accordance

,,,,o 4

With 10 CFR 50.55a(a)(?.)(l). This conclusion is further supported by the fact that the ASME has approved Code Case N 566 and this interim relief request is essentially a conservative subset of the Code Case."

Evaluatiom In accordance with IWA 5250(a)(2), if les'. age occurs at a bolted connection, the bolting must be removed, VT 3 visually examined for corrosion, and i evaluated in accordance with IWA 3100, in lieu of this requirement, the licensee has proposed to evaluate the botting to determine its susceptibility to corrosion. Based on the items included in the evaluation process, the INEEL staff believes that the evaluation proposed by the licensee provides a sound engineering approach. In add; tion, if the initial evaluation indicates the need for a more detailed analysis, the bolt closest to the source of the leakage will be removed, VT.1 examined, and evaluated in accordance with IWA 3100(a). ,.,

Based on the bolting evaluation criteria contained in the interim relief request, the INEEL staff concludes that the licensee's proposed alternative to the requirements of IWA 5250(a)(2) is a conservative and technically sound engineering approach. As a result, significant patterns of degradation will be detected and an acceptable level of quality and safety will be provided.

3.0 CONCLUSION

The INEEL staff evaluated the licensee's Interim Request for Relief IWA 5250(a)(2) and concludes that the proposed alternative contained in the request will provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

.