ML20198T362

Safety Evaluation Authorizing Supporting Second 10-year Interval ISI Program Requests for Relief RR-17 & RR-25 as Submitted by Southern Nuclear Operating Co on 970529

ML20198T362
Person / Time
Site:	Vogtle
Issue date:	10/24/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198T333	List: ML20198T328 ML20198T362
References
NUDOCS 9711140267
Download: ML20198T362 (22)

Text

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,y k_ UNIYED STATES j

s NUCLEAR REGULATORY COMMISSION

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SAFETY EVALUATION BY THE OFFICE OF NUCI FAR REACTOR REGULATION OF THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION. PROGRAM PLAN REQUESTS FOR RELIEF VOGTI E Ft FCTRIC GENERATING PLANT. UNITS 1 AND 2 SOUTHERN NUCLEAR OPERATING COMPANY. INC.

DOCKET NOS. 50-424 AND 50-425

1.0 INTRODUCTION

In order to demonstrate the operability of American Scciety of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code (ASME Code) Class 1,2, and 3 components, the Technical Specif cations (TS) for Vogtle Electric Generati- "2nt, Units 1 and 2 (Vogtle), state that the inservice inspection of the ASME Code Class i, e, and 3 components shall be performed in ,

accordance with Section XI of the ASME Cnde and appl'::able addenda as required by Title 10 of the Code of Federal Reaulations (10 CFF ), Section 50.55a(g), except where specific writtsn relief has been granted by the Commission ,xJrsuant 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 alternatives would provide an acceptable level of quaMy and safety or (ii) compliance with the specified requirements would result in hardship or i unusual difficulty without a compeneming increase in the level of qualny and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 cc..nponents (including supports) nhall meet the requirements, except the design and access provisions and the pre-service 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 regulation 3 require  ?

that inservice examination of components and mystem pressure tests conducted during the first

' 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Sect:on XI of the ASME Code incorporated by reference in 10 OFR 50.55a(b) 12 months prior to the start of the 120-month interval rubject to the limitations and modifications listed therein. The applicable ed4 ion of Section XI of the ASME Code for the

. Vogtle second 10-year inservice inspection (ISI) interval is the 1989 Edition.

Pursuant to 10_CFR 50.55a(g)(5), if the licensee determines that conforman::e 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 made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose attomative requirements that are determined to be authorized by law, will not endanger life, 9711140267 971024 PDR ADOCK 05000424

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property, or the common defense and security, and are othewise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

In a letter dated May 29,1997, the Southem Nuclear Operating Company, Inc. (licensee),

submitted requests for relief to ASME Code Section XI requirements for the Vogtle Electric Generating Plant, Units 1 and 2. The licensee provided additional information in its letters dated September 10 and 17,1997.

2.0 EVALUATION The staff, with technical assistance from its contractor, the Idaho National Engir,eering and En, tironmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its second 10-year inservice inspection interval program plat, requests for relief for the Vogtle Electric Generating Plant, Units 1 and 2. Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the attached Technical Letter Report.

Request for Rollef RR 17: Examination Category C-F-1, item C5.11, Class 2 Circumferential Piping Welds requires 100% volumetric and surface examination, as defined by Figure IWB-2500-7, of circumferential piping welds greater than or equal to 3/8-inch nominal wall thickness for piping greater than 4-inch nominal pipe size. The welds selected for examination shali include 7.5%, but not less ths.128 welds, of all austenitic stainless steri, welds or high alloy welds not exempted by 'WC-1220.

Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed the following attemative (as stated):

Southem Nuclear Operating Company proposes to not only include Class 2 pressure-retaining austenitic stainless steel piping welds greater than 4 inches ,

NPS [ nominal pipe size) end less than 3/8 inches nominal wall thickness in the ,

total weld count to whie the 7.5% sampling rate is applied as required by Examination Cctegory C-F-1, but to also include these welds as part of the weld peculation fron which the 7.5% sample of welds receiving NDE is selected. As a result, NDE will be performed on a 7.5% sample of all non-exempt piping welde greater than 4 S;,nes NPS. The examinations will be cistributed among the Class 2 systems prorated, to the degiee practical, on the number of nonexempt austenitic stainless steel or high alloy welds in each system. Within a system, the examinations will be distributed among terminal ends and structural discontinuities prorated, to the degree practica!, on the number of non-exempt terminal ends and structural discontiquities in that system; and within each system, examinations shall be distributed be' ween line sizes prorated to the degree practicable. Structural discor3uities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe tittings (such as elbows, tees, reducers, flanges, etc., conforming to ANSI B16.9), and pipe branch connections, and fittings. Each weld selected will receive a surface and volumetric examination with the exception of socbt welds

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'3 and pipe branch connections of branch piping which will receive a surface examination only. -

The Code requires examination of a 7.5% sample of austenitic stainless steel or high alloy welds not exempted by IWC-1220. Examinations are limited to welds in piping greater than ,

3/8-inch nominal wall thickness and 4-inch NPS. In lieu of the Code requirements, the licensee .  ;

' proposed to include piping welds less than 3/8-inch nominal wall thickness in the 7.5% >

examination sample. These welde will receive both surface and volumetric examination with the exception of socket welds and branch connection welds, which would only receive the Code-required surface examination.

As currently written, the Code includes piping wekis less than 3/8-inch nominal wall thickness in the total population of both nonexempt and exempt piping welds, but excludes the piping less than 3/8-inch nominal wall thickness froa volumetric examination. Although not required by the Code, the staff considers volumetric examination of these thin-walled piping walds in the residual heat removal, emergency core cooling, and containment heat removal systems to be _ '

technically prudent. Furthermore, the licensee is meetirm the other selection criteria of the Code (i.e., terminal ends and structural discontinuities). Since distributing the examinations among these welds does not change the population size or the number of surface and volumetric examinations being performed on the population, the licensee's proposal provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee's proposed attemative is authorized .

Request for Relief RR-22: The Code requires that repairs and replacements be performed in accordance with IWA-4000 and IWA-7000, respectively. Pursuant to 10 CFR 50.55a(a)(3)(i), I the licensee proposed to use Code Case r4-544, Repair and Replacement of Smallitems,  :

Section XI, Division 1, in lieu of the repair and replacement requirements of the Code.  !

l Code Case N-544 replaces existing repair and replacement subsections of the Code and relaxes certain requirements, particularly for piping and tubing NPS 1-inch and smaller. The staff's evaluation of this Code Case indicates that the present Code requirements to which the licensee is proposing an altemative are generally administrative in nature. However, the Code Case lacks clarity regarding the repair and replacement of heat exchanger tubing; specifically, Class 1 steam generator tubing. If this Code Case is meant to extend to s;eam Generator i tubing, the staff finds this alternative unacceptable. Code Case N-544 has not yet received  !

thorough technical review by the staff. Further evaluation, with consideration given to the effect l on steam generator tubing, is required for the staff to have an adequate basis for approving the proposed attemative. in addition, the licensee has not shown that this Code Case provides an acceptable level of quality and safety. Therefore, the licensee's proposed attemative is denied.

f Request for Relief RR-25: IWA-525C(a)(2) of the 1989 Edition of ASME Code Section XI states that if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100. Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed an attemative to the requirements of IWA-5250(a)(2) regarding corrective actions for leakage at bolted connectic'1s. The licensee stated:

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When leakage is identified'st bolted connections by visual, VT-2 examination duing system pressure testing, an evaluation will be perforr.ed to determino the

• susceptibility uf the bolting to corrosion and assess the potential for failure. The following factors, may be considered, as applicable, bt4 not limited to, when evaluating the acceptability of the botting; ,

1, Botting materials

2. Corrosiveness of process fluid leaking -

3. Leakagelocation

4. Leakage history at connection or other system component

5. Visual evidence of corrosion at connection (connection assembled)

6. Service age of bolting materials t When the pressure test is performed with the system in service or required by-the Technical Specifications to be operable, and the bolting is susceptible to corrosion, the evaluation shall address the connection's structural integrity until the next component / system outage of sufficient duration. If the evaluation concludes that the system can perform its safety related function, remova! of the bolt closest to the leakage and VT-1 visual examination of the bolt will be performed when the system or component is taken out of service for a sufficient duration for accomplishment of other syttem maintenance activities.

For botting that is susceptible to corrosion, and when the initial evaluation indicates that the connection cannot conclusively perform its safety function until the next component / system cutage of sufficient duration, the bolt closest to the source of leakage will be removed, receive a VT 1 visual examination, and be evaluated in accordance with IWA-3100(a),

in accordance ...in IWA-5250(a)(2), if leakage occurs at a bolted connection, the botting must be removed, VT-3 visually examined for corrosion, and evaluated ir accordance with IWA-3100. In lieu of this requiremert, the licensee has proposed to evaluate the bolting to determine its susceptibility to corrosion. The proposed evaluation will consider, as a min l mum, botting materials, the corrosive nature of the process fluid, the leakage location and history, the

- service age of the botting materials, and visual evidence of corrosion at the assembled connection. Based on the items included in the evaluation process, the staff concludes that the avaluation proposed by the licensee presents a sound engineering approach to assure the ,

intogrity of the bolting materials, in addition, if the initial evaluation indicates the need for a mote detailed analysis, the bolt closest to thc source of the leakage will be removed, visually exan.ined, and evaluated in accordance with IWA-3100(a).

Based c,n the bolting evaluation criteria contained in the proposed attemative, the staff concludes that the licensee's proposed attemative to the requirements of IWA-5250(a)(2) is a -

technically sound engineering approach to detect significant pattems of degradation and will  ;

provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.C3a(a)(3)(i), the licensee's proposed attemative is authorized.

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Request for Relief RR 26: For systems borated for the pumose of controlling reactivity, Sectica XI, Subparagraph IWA-5242(a) requires removal of insulation frc m pressure-retaining bolted connections for VT 2 visual examination during system pressure testing. Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee has proposed an attemative to the requirements of IWA-5242(a). The licensee stated:

Class 1 comoonents:

Insulated Class 1 pressure-retaining bolted connections will be uninsulated and VT-2 examined once each refueling outage while the connections are at atmospheric or static pressures. The bolted connections will be examined again with the insulation installed during the regularly scheduled system pressure test at nominal system operating temperature and pressure as required by Table

'WB>2500-1, Examination Category B-P.

Gass 2 arid 3 comoonents:

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Insulated Chss 2 and 3 pressure-retaining boKed connections will be unir:sulated and VT-2 examined once each inspection period. These examinations may be performed when the connections are not at the pressures required by IWA-5000, IWC-5000, and IWD-5000. The bolted connections wili be examined again with the insulation installed during the regularly scheduled (once per inspection period) system passure test as required by Tabl: ENC-2500-1, Examination Category C-H and Table IWD-2500-1, Examination Category D A.

The Code requires the removal of all insulation from pressure-retaining bolted connections in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tests. As an attemative, the licensee has proposed to perform the VT-2 visual examination required by the Code with the insulation in place and to perform a separate, direct, visual examination with the insulation removed, but without pressurizing the system and temperature each refueling outage for Class 1 components and each period for Class 2 and 3 compone,its.

The licensee's proposed attemative for Class 1 systems is essentially the same as that contained in Code Case N 533, Altemative Requirements for VT-2 Visual Examination of Class 1 Insulated Pressure-Retaining Bolted Connections,Section XI, DMsion 1, which is currently under review by the staff and has not yet been approved for use by incorporation into Regulatory Guide 1.147, inservice inspection Code Case Acceptability. However, the Code Case has been found acceptable on a plant specific basis when a licensee proposes to use a 4-hour hold time on Class 1 insulated systems. In addition, tha licensee has proposed to extend the use of the attemative to Code Class 2 and 3 bolted connections.

For Class 1 systems, the licensee's propoa.ed attemative provides a reasonable approach for ensuring the leak-tight integrity of systems borated for the purpose of controlling reactivity.

However, since the licensee proposed an attemative in lieu of the Code,Section XI, Paragraph IWA-5250(a)(2), the licensee is required by the Code, Sectien XI, Paragraph IWA-5213, to apply a 4-hour hold time during the pressure test to allow time for any leakage to penetrate the

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insulation. This provides a means of detecting any significant leakage with the insulation in place. Secondly, by removing the insulation each refueling outage, the licensee will be able to detect minor leakage that could occur by tha presence of boric acid arystals or residue. This two-phased approach provides an acceptable level of quality and safety for bolted connections in borated systems. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee's proposed attemative is authorized for use on Class 1 systems.

For Class 2 and 3 systems, the licenses has not shown that the once per period proposed frequency for insulation removal meets the criteria for authorization of a 10 CFR 50.55(a)(3) attemative. Therefore, the licensee's proposed attemative is denied for Class 2 and 3 systems.

3.0 CONCLUSION

3 The staff has reviewed the licensee's submittals and concludes that for Requests for Relief RR-17 and RR 25, the licensee's proposed alternatives provides an acceptable level of quality and safety. Thanfore, pursuant to 10 CFR 50.55a(a)(3)(i), the licensee's propoced attematives contained in Requests for Relief RR-17 and RR-25 are authorized.

Pursuant to 10 CFR 50.55.(a)(3)(i) the licensee's proposed a'temative contained in Request for Relici RR 26, is authorized for Class 1 systems only. For Class 1 systems, it is concluded that the licensee's proposed altemative, which is equivalent to Code Case N-533, provides an acceptable level of quality and safety. Furthermore, as required by the Code, Paragraph IWA 5213, the licensee is required to apply a 4-hour hold time during the pressure test to allow time for any leakage to penetrate the insulation, which provides a means of detecting any significant leakage with the insulation in place. The use of the above attemative for Class 2 and 3 systems is denied, because the frequency of once per period proposed for insulation removal has not been shown to meet the criteria for authorization of an attemative per 10 CFR 50.55(a(3).

For Request for Relief RR-22, Code Case N-544 is urder staff review and the licensee did not provide an adequate technical basis showing that the code case provides an acceptable level cf quality and safety. Therefore, the proposed alternative is denied.

Attachment:

Technical Evaluation Report PrincipalContributor: T. McLellan Date: October 24, 1997

TECHNICAL LETTER REPORT SECOND 10-YEAR INTERVAL INSERVICE. INSPECTION REQUESTS FvR REL1EF SOUTHERN NUCLEAR OPERATING COMPANY. INC t V0GTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 DOCKET NUMBERS 50-424 AND 50-425

1.0 INTRODUCTION

By letter dated May 29, 1997, the licensee, Southern Nuclear Operating Company, Inc., submitted its second 10-year interval inservice inspection (ISI) program for Vogtle Electric Generating Plant (VEGP),

Units 1 and 2. In addition, the licensee requested expedited review of four, second 10-year interval ISI requests for relief from the requirements of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. In response to a Nuclear Regulatory Commission (NRC) request for additional information, the licensee provided further inform 3t'.on in letters dated September 10, 1997 and September 17, 1997. The Idaho National Engincaring and Environmental Laboratory (INEEL) staff has evaluated the information provided by the licerisee in support of these requests for relief in the following section.

2.0 EVALUATION The Code of record for the VEGF, Units 1 and 2, second 10-year ISI interval, which began May 31,1937, is the 1989 Edition of ASME,Section XI.

A) Reauest for Relief RR-17. Examination Cateaory C-F-1. Item C5.11.

Class 2 Circumferential Pioina Welds Code Reauirement: Examination Category C-F-1, Item C5.11 requires 100% volumetric and surface examination, as defined by Figure IWB-2500-7, of circumferential piping welds greater than or equal to pipe size. The welds selected for examination shall include 7.5%, but not less than 28 welds, of all austenitic stainless steel welds or high alloy welds not exempted by IWC-1220.

Attachment

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Licensee's Proposed Alternative: Pursuant to 10 CFR

- 50.55a(a)(3)(1), the licensee proposed the following alternative (as stated):

" Southern Nuclear Operating Company proposes to not only include Class 2 pressure-retaining austenitic stainless steel piping welds greater than 4 inches NPS and less than 3/8 inches nominal wall thickness in the total weld count to which the 7.5% sampling rate is applied as required by Examination Category C-F-1, but to also include these welds as part of the weld population from which the 7.5% sample of welds receiving NDE is selected. As a result, NDE will be performed on a 7.5% sample of all non-exempt piping welds greater than 4 inches NPS. The examinations elli be distributed among the Class 2 systems prorated, to the degree practical, on the number of nonexempt austenitic stainless steel or high alloy welds in each system. Within a system, the examinations will be distributed among terminal ends and structural dircontinuities prorated, to the degree practical, on the number of non-exempt terminal ends and structural discontinuities in that system; and within each system, examinations shall be distributed between line sizes prorated to the degree practicable. Structural discontinuities include pipe weld joints to vessel n:azles, valve bodies, pump casings, pipe fittings (such as elbows, tees, reducers, flanges, etc. conforming to ANSI B16.9), and pipe branch connections and fittings. Each weld selected will receive a surface and-volumetric examination with the exception of socket welds and pipe branch connections of branch piping which uill receive a surface examination only."

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

"The majority of the Class 2 pressure-retaining austenitic stainless steel piping welds in this scope at VEGP-1 and 2 are greater than 4 inches nominal pipe size (NPS) and.less than 3/8 inches nominal wall thickness. These welds are not exempted by IWC-1220 but because of their wall thickness are not required to have NDE performed per Examination Category C-F-1. Therefore, if only the non-exempted welds greater then 4 inches NPS and greater than or equal to 3/8

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3 inches nominal wall thickness are considered for NDE, the selection process may exclude entire systems from NDE.

"In the First Ten-Year Interval, the " Multiple Strerm Ccncept" was used in the selection of Class 2 piping welds in tt desidual Heat Removai (RHR) Emergency Core Cooling System (FC;5), and the Containment Heat Removal (CHR) systems per the 1974 Edition of ASME Section XI with Addenda through Summer 1975 as required by 10 CFR 50.55a(b)(2)(iv). Georgia Power Company (GPC), the former licensee of VEGP, chose not to apply the pressure and temperature exemption of RHR, ECCS, t.nd CHR systems and required that 7.5% of the welds in these systems not exempted by IWC-1220(a) or IWC-1220(c) be volumetrically examined once each ten year interval. GPC also committed to volumetrically examine, once each ten year, a 7.5%

sample of the welds in these systems which only required surface examination by t M Code. (In addition, GPC committed to volumetrically examine 7.5% of the welds four inches and smaller but greater than or equal to two inches on high pressure safety injection systems once each ten year interval). Therefore, the scope of the Class 2 pressure-retaining austenitic stainless steel welds for VEGP-1 and 2 in the First Ten-Year Interval was similar in the scope defined by the 1989 Edition of the ASME Section XI. SNC is simply requesting a continuation of the philosophy that was used during the first interval for the selection of the subject Class 2 welds, except that in lieu of the "Haltiple Steam Concept", a straight 7.5% sample will be taken.

"These piping welds greater than 4 inches and less than 3/8 inches nominal wall thickness make up the majority of the VEGP-1 and 2 Class 2 pressure-retaining austenitic stainless steel piping welds.

Including 7.5% of these welds in the NDE scope will provide a better representative sample of the total number of Cless 2 austenitic .

stainless steel piping welds found in both units as well as providing an acceptable level of quality and safety. Therefore, it is requested that the proposed alternative be authorized pursuant tc 10 CFR 50.55a(a)(3)(1)."

Evaluation: The Code requires examination of a 7.5% sample of austenitic stainless steel or high alloy welds not exemoted by IWC-1220. Examinations are limited to welds in piping greater than 3/8-inch nominal wall thickness and 4-inch NPS. In lieu of the Code requirements, the licensee proposed to include piping welds less than 3/8-inch nominal wall thickness in the 7.5% examinar.on sample.

These welds will receive both surface and volumetric examination with the exception of socket welds and branch connection welds, which would only receive the Code-required surface examination.

4 As currently written, the Code includes piping welds less than 3/8-inch nominal wall thickness in the population, but exclu' des then from examination. Although not required by the Code, the INEEL staff considers examination of these thin-walled piping welds in the RHR, ECC and CHR systems to be technically prudent. Furthermore, the licensee is meeting the other selection criteria of the Code (i.e., terminal ends and structural discontinuities). Since distributing the examinations among these welds does not change the population size or the number of surface and volumetric examinations being performed on the population, the licensee's proposal provides an acceptable level of quality and safety. Therefoie, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(1).

B) Reauest for Relief RR-22. Use of Code Case N-544. Reoair and Reolacement of Small items.Section XI Division 1 Code Reouirement: The Code requires that repairs and replacements be performed in accordance with IWA-4000 and IWA-7000, respectively.

Licensee's Proposed Altarnativn Pursuant to 10 CFR 50.5Ea(a)(3)(1), the licensee proposed to use Code Case N-544, Repeir and Replacement of Small items,Section XI, Division 1, in lieu of the repair and replacement requirements of the Code.

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

"The 1989 Edition of ASME Section XI Code provides an exemption for replacement items 1-inch NPS and smaller from the requirements of IWA-7000, but repairs to such items are not similarly exempted.

Therefore, a repair to en item is subject to more restrictive ,

requirements than. replacing an item. Code Case N-544 allows '

application of the alternative requirements for replacement to all  ;

repair and replacement activities. The ASME Code Committee i evaluated the proposed alternatives contained in the Code Case and l determined that they are acceptable for repair and replacement activities on piping, valves, and fittings 1-inch NPS and smaller.

In addition, the provisions of this code case were added to IWA-4131 and IHA-4132 in the 1995 Addenda to ASME Section XI. SNC has determined that implementation of the Code Case will not affect the l

5 level of quality and safety, nor decrease the margin of public healti and safety.

"While the cost savings associated with Code Case N-544 have not been quantified as a Cost Beneficial Licensing Action item, its implementation is consistent with the intent to eliminate non-beneficial work activities and their associated costs. Therefore, it is requested that the proposed alternative be authorizea pursuant to 10 CFR 50.55a(a)(3)(1).

Evaluation: The Code requires that repairs and replacements be performed in accordance with IWA-4000 and IWA-7000, respectively.

The licensee has proposed to meet the requirements of Code Case N-544, Repair and Replacement of Small items,Section XI, Division 1, in lieu of the repair and replacement requirements of the Code.

Code Case N-544 replaces existing repair and replacement subsections of the Code and relaxes certain requirements, particularly for piping and tubing NPS 1-inch and smaller.

Preliminary evaluation of this Code Case indicates that the present Code requirements to which the licensee is proposing an alternative are generally administrative in nature. However, the Code Case lacks clarity regarding the repair and replacement of heat exchanger tubing; specifically, Class I steam generator tubing. If this Code Case is to be extended to steam generator tubing, the staff may find this alternative unacceptable. Code Case N-544 has nSt yet received thorough technical review by the NRC staff and further evaluation of this alternative to the Code is warranted. In addition, the licensee has not demonstrated that the proposed alternative provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative not be authorized.

C) Reauest for Relief RR-25. IWA-5250f aH2). Corrective Actions for Bolted Connections Code Reauirement: IWA-5250(a)(2) of the 1989 Edition of ASME Section XI states that if leakage occurs at a bolted connection, the bolting

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- shall be removed, VT-3 visually examined for corrosion, and _

evaluated in accordance with IWA-3100. .

Licensee's Pronosed' Alternative: Pursuant-to 10 CFR 50.55a(a)(3)(1),

l the licensee proposed to an alternative to the requirements of IWa-5250(a)(2) regarding corrective actions for leakage at bolted connections'. The licensee stated:

"When leakage is identified at bolted connections by visual, VT-2 examination during system pressure testing, an evaluation will be performed to determine the susceptibility of the bolting to corrosion and assess the potential for failure. The following factors, may be considered, as applicable, but not limited to, when-evaluating the acceptability of the bolting;

, 1. Bolting materials

2. Corrosiveness of process fluid leaking

3. Leakage location

4. Leakage history at connection or other system components

5. Visual evidence of corrosion at connection (connection assembled)

6. Service age d bolting materials "When the pressure test is performed with the system in service or required by the Technical Specifications to be operable, and the bolting is susceptible to corrosion, the evaluation shall address the connection's structural integrity until the next component / system outage of sufficient duration. If the evaluation concludes that the system can perform its safety related function, removal of the bolt closest to the leakage and VT-1 visual examination of the boit will be performed when the system or component is taken out of service for a sufficient duration for accomplishment of other system maintenance activities.

"For bolting that is susceptible to corrosion, and when the initial

, - evaluation indicates that the connection cannot conclusively-perform its safety function t.ntil the next component / system outage of .

sufficient duration, the bolt closest to the source of leakage will be removed, receive a VT-1 visual- examination, and be evaluated in accordance with IWA-3100(a)."

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

" Removal of pressure retaining bolting at mechanical connections for visual, VT-3 examination and subsequent evaluation in locations-where leakage has been identified is not always the most prudent course of action to determine the acceptability of the bolting. The Code requirement to remove, examine, and evaluate bolting in this

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7 situation does not allow the Owner to consider other factors which may indicate the acceptability of mechanical joint bolting.

Southern Nuclear Operating Company (SNC) considers this requirement to be unnecessarily prescriptive and restrictive.

"Other factors which should be considered when evaluating bolting acceptability when leakage has been identified at a mechanical joint include, but should not be limited to: joint bolting materials, service age of joint bolting materials, location of leakage, history of leakage at the joint, evidence of corrN!on with the joint assen. bled and corrosiveness of process fluid.

"A NE Section XI is written to primarily address examination and testing during periods of plant or system shutdown. No guidance is given to atiress components that are examined or tested while the plant or system is in service. However, many Code Class 2 and 3 systems are pressure tested, including VT-2 visually examined, utilizing the ' Inservice Test' requirements of IWA-5000.

" Performance o' the test while the system is inservice w identify leakage at a bolted connection that, upon evaluation, r -

concluded that the joint's structural integrity and preswe retaining ability is not challenged. It would not be prudent to negatively impact a safety system's availability by removing the system from service to address a leak that does not challenge the system's ability to perform its safety function.

"In addition, a situation frequently encountered at commercial nuclear plants such as VEGP is a complete replacement of bolting materials (studs, bolts, nuts, washers, etc) at mechanical joints during plant outages. When the associated system process piping is pressurized during plant start-up, leakage may be identified at these joints. The root cause of this leakage is most often due to thermal expansion of the piping and bolting materials at the joint and subsequent process fluid seepage at the joint gasket. Proper retorquing of the joint bolting, in most cases, stops the leakage.

Removal of any of the joint bolting to evaluate for corrosion would be unwarranted in this situation due to the new condition of the bolting materials.

"Souihern Nuclear Operating Company proposes the following citernative (see ' Proposed Alternative' section above] methodology to the requirements of IWA-5250(a)(2) which will provide an acceptable level of quality and safety when evaluating leakage and bolting material acceptability at Class 1, 2, and 3 bolted connections."

Evaluation: In accordance with IWA-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.

e 8

In lieu of this requirement, the licensee has proposed to evaluate the bolting to determine its susceptibility to corrosion. The proposed evaluation will consMar, as a miniraum, bolting materials, the corrosive nature of the process fluid, the leakage location and history, the service age of the bolting materials, and visual evidence of corrosion at the assembled connection. Based on the items included in the evaluation process, the INEEL staff believes that the evaluation proposed by the licensee presents 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 leakage will be removed, visually examined, and evaluated in accordance with IWA-3100(a).

Based on the bolting evaluation criteria contained in the proposed alternative, the INEEL staff concludes that the licensee's proposed alternative to the requirements of IWI.-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. Therefore, it is recommended that the licensee's propased alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

D. Reouest for Relief RR-26. IWA-5242(a). Visual Examination of Insulated Comoonents Code Reouirement: For systems borated for the purpose of controlling reactivity, Subparagraph IWA-5242(a) requires removal of insulation from pressure-retaining bolted connections for VT-2 visual examination during system pressure testing.

Licensee's Proposed Alternative: Pursuant to 10 CFR 50.55a(a)(3)(1),

the licensee has proposed an alternative to the requirements of IWA- i 5242(a). The licensee stated:

. e t 9

• Class 1 comnonents:

" insulated Class 1 pressure-retaining bolted connections will be uninsulated and VT-2 examined once each refueling outage while the connections are at atmospheric or static prersures. The bolted -

connections will be examined again with the insulation installed durir.g the regularly schedule system pressure test at nominal system operating temperature and pressure as required by Table IWB-2500-1, Examination Category B-P.

" Class 2 and 3 comoonents:

" Insulated Class 2 and 3 pressure-retaining bolted connections will be uninsulated and VT-2 examined once each inspection period. These examinations may be performed when the connections are not at the pressures required by IWA-5000, IWC-5000 and IWD-505. The bolted connections will be examined again with the insulation installed during the regularly scheduled (once per inspection period) system pressure test as required by Table IWC-2500-1, Examination Category C-H and Table IWD-2500-1, Examination Category D-A."

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

" Subparagraph IWA-5242(a) specifies that insulation must be removed from pressure-retaining bolted connections for VT-2 visual examination during the performance of system pressure testing. This is applicable to the following systems:

Reactor Coolant System (System consists of Class 1 and 2 components),

Chemical and Volume Control System (System consists of Class 1, 2, and 3 components),

Residual Heat Removal System (System consists of Class 1 and 2 components),

Safety Injection System (System consists of Class 1 and 2 components),and Nuclear Sampling System - Liquid (System consists of Class 2 components).

" Class 1 Components:

" Table IWB-2500-1, Examination Category B-P requires a system leakage test (IWB-5221) and corresponding VT-2 visual examination of Class 1 components each refueling outage prior to plant startup. This system leakage test is performed in Mode 3 when the Reactor Coolant System is at Nominal Operating Pressure (=2235psig) and Nominal Operating Temperature (=550*F to 650*F). The majority of the Class I components are in the Reactor Coolant System however some portions of the Class 1 boundary extend to include portions of Safety Injection, Chemical Volume Control, and Residual Heat Removal systems. All Class I components are in containment. The removal and installation

i 10 of insulation during the performance of system pressure testing inside containment presents the following hazards:

Increased potential for debris to be in containment which could migrate to the Containmerd Emergency Sumps and restrict tt.o suction of the Emergency Core Cooling System during accident (LOCA) conditians. All debris is required to be removed from containment prior to entering Mode 4,

-Increased potential for personnel heat stress s'nce the containment ambient temperature may be as high as 100*F, Increased potential for personnel burn injuries due to installation of insulation in proximity of extremely hot components, Increase personnel safety hazard since ladders would have to be used to inspect many of the bolted connections and replace the insulation. Temporary work platforms / scaffolding inside containment are removed prior to entering Mode 4, Increased radiation exposure to personnel since temporary shielding is removed prior to entering Mode 4, and Increased potential for impacting outage duration due to amount of manpower required to support insulation removal and examinations during Mode 3 following refueling outage activities.

" Class 2 Components:-

" Table IWC-2500-1. Examination Category C-H requires a system pressure test (IWC-5221) during system functional and system inservice tests and corresponding VT-2 visual examination on Class 2 components once each inspection period (40 months). The following discusses the applicable systems and the basis for relief for each Class 2 system:

1. Basis for Relief of Reactor Coolant System (RCS):

The Class 2 portions of RCS are located adjacent to the Class 1 boundary and are classified as Class 2 based on line size and isolation valve criteria. The system inservice tests for these Class 2 portions of RCS are VT-2 examined in Modes 1, 2 and 3 and therefore the same basis for relief as provided for Class 1 applies. These Class 2 pressure boundaries are located in containment.

2. Basis for Relief for the Chemical and Volume Contr'.1 System (CVCS):

For those portions of CVCS which are located inside containment (Charging, Letdown, Excess Letdown, Alternate Pressurizer Spray,

I 11 Reactor Coolant Pump Seal Leakoff, etc.) the same basis for relief as provided for Class 1 above applies except that the system operating temperatures are less resulting in less potential for burn-related injuries. The VT-2 examinations are performed in Modes 1, 2, and 3.

For those portions of CVCS which are located outside containment, radiation levels, high componant temperatures, and availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized.

CVCS is inservice during power operation and, as such, many, if not all, components will be uninsulated, VT-2 examined, and reinsulated when the pressure-retaining bolted connections are pressurized. However, as previously addressed, conditicas may be present which may not allow insulation removal except during refueling outages.

3. Basis for Relief for the Residual Heat Removal System (RHR):

RHR is placed inservice during shutdown prior to refueling activities in Mode 4 when the Reactor Coolant System is =350

• F and =350 psig. RHR remains inservice in Modes 5 and 6 during the refueling outage and remains inservice in Mode 4 during startup following the refueling outage. The VT-2 examinations are performed in either Mode 4 or Mode 5 when the Reactor Coolant System is = 350 ps'g.

For those portions of RHR which are located inside containment the same basis for relief as provided for Class 1 above applies except that the system operating temperatures are less resulting in less potential for burn-related injuries. It is impractical to attempt to limit VT-2 examinations to Mode 5 in order to avoid the complications of performing RHR pressure test in Mode 4. The V1-2 examinations are performed in Modes 4 or 5.

For those pon ions of RHR which are located outside containment radiation levels, high component temperatures, availability of personnel, and increased thermal loads on chilled water room cooling systems may preclude removing. insulation while the pressure-retaining bolted connections are pressurized. It is tignificantly more prudent to uninsulate, VT-2 examine, and reinsulate the pressure-retaining bolted connections in RHR when the system is not pressurized during non-outage times or during refueling outages when the system is not at the required pressure.

4. Basis for Relief for the Safety Injection System (SI):

The system pressure tests performed on SI are either system functional tests or system inservice tests as follows:

....l, i 12 Some of the system functional tests are performed during Modes 1, 2, and 3 when Reactor Coolant System pressure is greater than SI pump discharge pressure. VT-2 examinations are performed on 2

portions of SI during various activities and tests which require a SI pump to be in operation. The score of these VT-2 examinations includes components both inside and outside containment. The perfumance of the VT-2 examinatione during these activities is generally performed in less than one hour to minimize run time on the SI pumps.

The remainder of the system functional tests are performed during Mode 6 and defueled conditions with the reactor head removed. VT-2 examinations are performed on portions of SI which are pretsurized during check valvo flow testing activities which involve injection of water into the reactor pressure vessel. The scope of these VT-2 examinations includes components both inside and outside containment. The performance of the VT-2 examinations during these activities is generally performed in less than one hour to minimize run time on the applicable pumps and to minimize the impact on critical path testing during refueling outages.

Some of the system inservice tests are performed during Modes 1, 2, or 3 on portions of SI which are pressurized by the SI accumulator tanks. The SI accumulator tanks are generally depressurized during refueling outages. The scope of these VT-2 examinations includes components which are only inside containment.

The remainder of %e system inservice tests are performed on portions of SI which are pressurized by the static head of the refueling water storage tank. The VT-2 examinations on these portions of SI are generally performed during power operation (Mode 1) but may be performed in other Modes if tank levels are adequate. The scope of these VT-2 examinations includes components which are only outside contairment.

For those portions of SI which are located outside containment radiation levels and availability of personnel during non-outage times or during system functional testing may preclude removing insulation while the pressure-retaining bolted connections are pressurized.'

For those portions of SI which are located inside containm3nt and VT-2 cxamined during Modes 1, 2, and 3 the same basis for relief as provided for Class 1 above applies except that the system operating temperatures are less resulting in less potential for burn-related injuries.

For those portions of SI which are located inside containment and VT-2 examined during Mode 6 and defueled conditions with the reactor vessel head removed containment radiation levels and availability of personnel during system functional testing may

e._.' O

. 3 13 preclude removing insulation while the pressure-retaining bM ted connections are pressurized.

S. Basis for Relief for the Nuclear Sampling System - Liquid:

The liquid portions of the Nuclear Sampling System are used for providing samples for analysis purposes of the Reactor Coolant System, Chemical and Volume Control System, and Residual Heat Removal System. This system is located both inside and outside '

containment and is subject to the same system pressure tests as the systems for which it used to provide samples. Therefore, the same basis for relief as discussed above is applicable to the liquid portions of the Nuclear Sampling Sys: tem Class 3 components:

" Table IWD-2500-1 Examination Category D-A requires a system inservice test (IWD-5221) and correspending VT-2 visual examination on Class 3 components once each inspection period (40 months).

Subperagraph IWA-5242(a) is applicable to the boric acid storage tank and boric acid transi'er portions of the Chemical and Volume Control System. System inservice test are performed as follows:

"Some of the system inservice tests are performed on portions of CVCS which are pressurized by the static head of the boric acid storage tank. The scope of these VT-2 enminations includes components which are only outside containment. The VT-2 examinations on these portions of CVCS are generally performed during power operation (Mode

1) but may be performed in other Modes if tank levels are adequate.

These system inservice tests are generally performed during power operation atd, as such, many, if not all, components will be uninsulated, VT-2 examined, and reinsulated when the pressure-retaining bolted connections are pressurized. However, availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized.

"The remainder of the system inservice tests are performed on portions of CVCS which are pressurized when a boric acid transfer pump is operating. The scope of these VT-2 examinations includes componets which are outside containment. The VT-2 examinations on these portions of CVCS are generally performed during power operation (Mode 1) with a boric acid transfer pump running with system valves aligned in a recirculation flowpath which preclude injecting high concentrations of borated water into CVCS and ultimately into the RCS. The boric acid transfer pumps are operated as necessary to perform system functions and necessary testing and, as such, are not continuously in operation. Since these pumps are not continuously in operation availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining connections are pressurized.

I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _

14 "The 1983 Edition through Summer 1983 Addenda of ASME Section XI was applicable for the First Ten-year interval at VEGP. IWA-5242 of the 1983 Edition through Summer 1983 Addenda of ASME Section XI did not require insulation removal; therefore, this request for relief was not needed at VEGP during the First-Year Interval.

" Justification for Grantina Relief:

"The fcilowing items are justification for granting relief for Class 1 Components:

1. Evidence of leakage through pressure-retaining bolted connections which are in systems which are borated for the purpose of controlling reactivity is readily detectable by visual observation when systems are not at operating temperature and oressure. The boric acid concentrations are sufficiently higi such that boric acid residues will be '

present if leakage has oce.urred at the pressure-retaining bolted connection.

2. In addition, the ASME Section XI Code Committee has issued Code Case N-533 which allows as an alternative for Class 1 pressure-retaining bolted connections that insulation may be removed and VT-2 examined when the connection is not pressurized. The Code Case also requires that any evidence of leakage be evaluated in accordance with IWA-5250. Refer to Request for Relief RR-25 for details concernino relief from IWA-5250(a)(2).

3. Compliance with the Code presents hardships previously discussed [in " Basis for Proposed Alternative" above).

4. For the reasons discussed above, SNC has determined that implementation of the proposed alternatives to the Code requirements provides an acceptable level of quality and safety and therefore requests that the proposed alternative pursuant to 10 CFR 50.55a(a)(3)(1).

Evaluation: The Code requires the removal of all insulation from pressure-retaining bolted connections in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tee s. As an alternative, the

, licensee has proposed to perform the VI-2 visual examination required by the Code with the insulation in place. In addition, a separate, direct visual examination with the insulation removed but without pressurizing the system will be performed during each refueling

f J. .. .. e 15 outage for Class I components and each period for Class 2 and 3 components.

The licensee's proposed alternative for Class I systems is essentially the same as that contained in Code Case N-533, Alternative Requirements for VT-2 Visual Examination of Class 1 Insulated Pressure-Retaining Bolted Connections,Section XI, Division 1, which is currently under review by the NRC_ staff and has not yet been approved for use by incorporation into Regulator Guide 1.147, inservice Inspection Code Case Acceptability. However, it the Code Case has been found acceptable on a plant-specific basis with the commitment of a 4-hour hold time on Class 1 insulated systems.

Further, the licensee has proposed to extend the use of the alternative to Code Class 2 and 3 bolted connections.

For Class I systems, the licensee's proposed alternative provides a reasonable approach for ensuring the leak-tight integrity of systems borated for the purpose of controlling reactivity. As required by the Code, the pressure test with a 4-hour hold time allows time for any leakage to penetrate the insulation, which provides a means of detecting any significant leakage with the insulation in place.

Secondly, by removing tha insulation each refueling outage, the licensee will be able to detect minor leakage that could occur by the presence of boric acid crystals or residue. This two-phased approach provides an acceptable level of quality and safety for bolted connections in borated systems. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(1) for use on Class I systems.

For Class 2 and 3 systems, the-frequencies proposed for insulation removal have not been found acceptable by the NRC staff. Therefore, the licensee's proposed alternative should not be authorized for Class 2 and 3 systems.

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. . .: o-I 16 3.0 : CONCLUSION The INEEL staff has reviewed the licensee's submittal and concludes that for Requests for Relief RR-17 and RR-25, the licensee's proposed alternatives will provide an acceptable level of quality and safety. Therefore, it is recommended the proposed alternatives.be authorized pursuant to 10 CFR 50.55a(a)(3)(i). For Request for Relief PR-26, it is recommended that the licensee's proposed alternative be authorized for Class I systems only. For Class 1 systems, it is concluded that the licensee's proposed alternative, which is equivalent to Code Case N-533, provides an acceptable level of quality and safety. However, the use of this alternative for Class 2 and 3 systems is still under NRC staff review. Therefore, the proposed alternative should not be authorized for Class 2 and 3 systems.

For Request for Relief RR-22, Code Case N-544 is under staff review and the licensee has not shown that the proposed alternative provides an acceptable level of quality and safety. Use of this Code Case should not be authorized.

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