ML20203E037

Forwards RAI Re Inservice Insp Program Plan & Associated Requests for Relief for Plant,Units 1 & 2

ML20203E037
Person / Time
Site:	Farley
Issue date:	02/12/1998
From:	Jacob Zimmerman NRC (Affiliation Not Assigned)
To:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
References
TAC-M98858, TAC-M98859, NUDOCS 9802260238
Download: ML20203E037 (9)

Text

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, , february 12, 1998 Mr. D. N. Mor:y Vice President Farley Project Southern Nuclear Operating Company, Inc.

s Post Office Box 1295 Birmingham, Alabama 35201 1295

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE INSERVICE INSPECTION PROGRAM PLAN AND ASSOCIATED REQUESTS FOR RELIEF JOSEPH M. FARLEY NUCLEAR Pl. ANT, UNITS 1 AND 2 (TAC NOS. M98858 AND M98859)

Dear Mr. Morey:

By letter dated May 28,1997, and supplemented by letter dated November 3,1997, you submitted the Joseph M. Farley Nuclear Plant, Units 1 and 2 Inservice Insptction (ISI)

Programs in accordance with the requirements of 10 CFR 50.55a(g)(4)(ii).

The staff, with assistance from its contractor, Idaho National Engineering and Environmental Laboratory (INEEL), has reviewed your submittals and determined that additional information is required. The enclosure identifies the requested additionalinformation (RAI) needed.

in order to maintain a timely review, it is requested that the information be provided within 60 days of receipt of this letter. In addition, to expedite the review process, please send a copy of your RAI response to the NRC contractor identified in the enclosure. If you require any clarification regarding this reqti st, please call me at (301) 415 2426.

Sincerely, ORIGINAL SIGNED BY:

Jacob 1. Zimmerman, Project Manager Project Directorate 112 Division of Reactor Projects -l/ll

. Office of Nuclear Reactor Regulation  !

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Docket Nos. 50-348 and 50-36*

Enclosure:

As stated cc w/ encl: See next page hI Qigiribution:

PDil 2 Rdg. JZimmerman JZwolinski ACRS doeN9t Me HBerkow PSkinner, Ril OGC PUBLIC LBerry JJohnson, Ril TMcLellan DOCUMENT NAME: G:\FARLEY\M98858 RAI To receive a copy of this document, Indicate in the box: "C" = Copy without attachment / enclosure "E" =

Copy with attachment /englosure "N" = No copy . A OFFICE PM POMh4 lE PM PDil-2 hM t LA',PDilp!#'\ l l NAME JZWfMSMAN'en LBERRY IQ1 HC'ERK(W/ /

DATE 2// 8 798 2/ lM98 7 t/ N 9B .ii6 c * *

• OFFiclAL RECORD COPY 9802260238 980212 -_I ~"' 7f PDR ADOCK 05000348- *" d " '-~ .- A h G PDR

s a ct:

,y e t UMTED STATES s g NUCLEAR HEGULATORY COMMi3SION "o g WASHINGTON, D.C. 30t#4'J01

%* . . , , , 'd' Tebruary 12, 1998 Mr. D. N Morey Vice President- Fa iey Project Southem Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201 1295

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE INSERVICE INSPECTION PROGRAM PLAN AND ASSOCIATED REQUESTS FOR RELIEF

- JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 (TAC NOS. M98858 AND M98859)

Dear Mr. Morey:

By letter datud May 28,1997, and supplemented by letter dated November 3,1997, you submitted the Joseph M. Farley Nuclear Plant, Units 1 and 2 Inservice Inspection (ISI)

Program iin accordance with the requirements of 10 CFR 50.55a(g)(4)(li).

The staff, with assistance from its contractor, Idaho National Engineering and Environmental Laboratory (INEEL), has reviewed your submittals and determined that additional information is required. The enclosure identifies the requested additionalinformation (RAI) needed.

In order to maintain a timely review, it is iequested that the infon?.ation be provided within 60 days of receipt of this letter. In addition, to expedite the review process, please send a copy of your RAI response to the NRC contractor identified in the enclosure. If you require any clarification regarding this request, please call me at (301) 415-2426.

Sincerely, h

Jacob 1. immerman, Project Manager Project Directorate ll 2 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50 364

Enclosure:

As stated cc w/ encl: See next page

l l

Joseph M. Farley Nuclear Plant  ;

cc:

Mr. R. D. Hill, Jr. I General Manager - ,

i Southem Nuclear Operating Company Post Omco Box 470 Ashford, Alabama 36312  !

Mr. Mark Ajiuni, Licensing Manager I Southem Nuclear Operating Company .

Post Office Box 1295 Birmingham, Alabama 35201 1205

Mr. M. Stanford Blanton

! Balch and Bingham Law Firm Post Office Box 306

• 1710 Sixth Avenue North Birmingham, Alabama 35201 -

Mr. J. D. Woodard Executive Vice President Southem Nuclear Operating Company

Post Office Box 1295 Birmingham, Alabama 35201 State Health Officer .

Alabama Department of Public Health 434 Monroe Street Montgomery, Alabama 30130-1701 9

Chairman Houston County Commission Post Office Box 6406 Dothan, Alabama 36302 Regional Administrator, Region ll U.3. Nuclear Regulatory Commissim Atlanta Federal Center 61 Forsyth Street, S.W., Suite 23T85 '

Atlanta, Georgia 30303 ,

~

Resident inspector U.S. Nuclear Regulatory Commission 7388 N. State Highway g5 Columbia, Alabama 3631g

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4 M ie!L M I' j ADanadion Program Plan

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1. Scopa/ Status f Review i

! Throughout the service life of a water-cooled nuclear power facility, Title 10 of the Code of l Fadatal Regulations (10 CFR) 50.55a(g)(4) requires that components (including supports) that  ;

j are classified as American Society of Mechanical Engineers (ASME) Boller and Pressure l

, Vessel Code (ASME Code) Class 1 Class 2, and Class 3 meet the requirements, except i j design and access provisions and preservice examir.ation requirements, set forth in Section XI i of the ASME Code, " 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. This section of the regulations also requires that inservice examinations of  :

! components and system pressure tests conducted during the successive 120-month inspection

intervals comply with the requirements in the latest edition and addenda of the Code  ;

incorporated by reference in 10 CFR 50.65a(b) on the date 12 months prior to the start of an j interval, subject to the limitations nnd modifications listed therein. The components (including ,

. _ supports) may meet requirements set forth in subsequent editions and addenda of the Code

that are incorporate i by reference in 10 CFR 50.55a(b) subject to the limitations and

modifications listed therein and subject to Nuclear Regulatory Commission (NRC) approval. .

The !icensee, Southem Nuclear Operating Company, has prepared the Joseph M. Farfey

! Nuclear Plant, Units 1 & 2, Third 10-Year IntervalInservice Inspection Program Plan, .

j Revision 0, to meet the requirements of the 198g Edition of Section XI of the ASME Code.

I L The staff has reviewed the available information in the Joseph M. Farfey Nuclear Plant, Un/t 1, <

Third 10-Year IntervalInservice Inspection (ISI) Program Plan, the Joseph M. Farfey Nuclear l Plant, Unit 2, Updated Inservice Inspection (ISI) Program Plan, Revision 0, submitted May 28,

j. 1997, and the licensee's requests for relief from ASME Code Section XI requirements.

2. AdditionalInformation Required >

l Based on the proceding review, the staff has concluded that additional information and/or clarification is required to complete its review, of the ISl Program Plan, j A. By a Safety Evaluation Report (SER) dated March 20,1997, the NRC allowed Southem Nuclesr Operating Company (SNC) to update the Unit 2 ISI Program approximately  ;

i 44 months early, to coincide with the required update of the Unit 1 program. As stated by SNC, 'this update will become effective on December 1,1997, and will continue in

.' place through November 30,2007. Unit 2 is presently in tlw second interval with the l third interval scheduled to start on July 30,2001; therefore, this update will cover part of

the second and third intervals?

~ The method of implementation for the Unit 2 program is unclear because the plan covers portions of two intervals. Explain SNC's plan for completing interval requirements, For example, when the Code allows deferral of examinations to the end ,

of the interval, what date will be used for the end of the interval? Also, describe how the 4 - trancition from the second 10 year interval to the third 10 year interval will take place. .

How will SNC close out the second 10-year interval prior to beginning the third 10 year i interval? Will all of the secona 10 ycar interval requirements and commitments be met? -

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, 2-IWB 2420 requires that the sequence of component exsminations established during the first inspection interval be repeated during each su;cessive inspection interval. By using the same examination sequence each interval, essentially no more than 10 years willlapse between examinations. Since SNC is combining portions of two intervals into one program plan, will the sequen;e of examinations be altered significantly from the first 10 Year interval? Will more than 10 years elapse between examinations for any component?

B. The Program submittal does nat include an implementation schedule for performance of examinations as required by IWA 2420(b). This is necessary to determine compliance with Tables IWB 24121, IWC 24121, and IWD 24121. Provide an implementation schedule including the total number of components in each item Number, the total number of componentt, selected for exatnination in cach item Number, and the period (include interval for Unit 2) in which the examinations will occur.

C. In accordance with 10 CFR 50.55a(c)(3),10 CFR 50.55a(d)(2), and 10 CFR 50.55a(o)(2), ASME Code Cases may be used as alternatives to Code requirements. Code Cases that the NRC has approved for use are listed in Regulatory Guide 1.147, inservice Inspection Code Case Acceptability, with any additional conditions that the NRC may have imposed, When used, these Code Cases must be implemented in their entirety. The licensee may adopt an approved Code Case by notifying the NRC in writing. Published Code Cases awaiting approval and subsequent listing in Regulatory Guide 1.147 may be adopted only if the licensee

.aquests, and the NRC authorizes, thelc use on a case by-case basis. Provide a list of all Code Cases, both approved and unapproved, that SNC intends to implement during the third 10 year interval.

D. Augmented examinations have been established by the NRC when added assurance of structural reliabihty is deemed necessary. Examples of documents that address augmented examinations are:

(1) Branch Technical Position MEB 31, High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containmsnt, and (2) Regulatorj Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations.

Address the degree of compliance with these documents and any other augmented examinations, which may have been incorporated in the third 10 year interval inservice inspection program plan.

E. Provide the staff with the status of the augmented reactor pressure vessel examinations required by 10 CFR 50.55a(g)(6)(ii)(A), effective September 8,1992, and provido a technical discussion of how the regulation was/will be implemented at Joseph M. Farley Nuclear Plant, Units 1 and 2. Include in the disec a e description of the approach and any specialized techniques or equipment that was# be used to complete the required augmented examination. Also, provide the percent of the volume examined for each

l 3-wold. Section 50.55a(g)(6)(ii)(A)(2) requires essentially 100% of the volume of each weld to be examined. Confirm that ' essentially 100T of each Examination Category 81.10 weld (RPV shell welds) has been examined, or that an alternative has been submitted for staff review.

F. Provide a list of all ultrasonic calibration blocks to be used during the third 10-year  !

interval, including identifications, material specifications, and sizes. Also, provide a list of all ultrasonic examination proceduren that will be used in the ISI Program; include titles and general descriptions of the components to which each procedure is applicable.

G. SNC must state %s specific paragraph of the Regulations under which each proposed request for rahef or attemative is submitted. In some cases, it appears that SNC has cited the wrong paragraph. For example, SNC has requested that RR 16 be granted pursuant to 10 CFR 50.55a(s)(3)(l). Howner, SNC states that ' Denial of this relief request would cause an excessive burden upon Southern Nuclear Operating Company because it hac been demonstrated that it is impractical to muet the Code requirements with presently developed techniques." Because SNC states that the required examination is a burden and imoractical, and no alternative examination is presented, it la unclear whether this relief request falls within the category of 50.55a(a)(3)(ii) or l 50.55a(g)(5)(lii).

l A licensee may pi ' pose an altemative to CFR or Code requirements in accordance with l 10 CFR 50.55s(a)(3)(1) or 10 CFR 50.55a(a)(3)(ii). When submitting a proposed alternative, the licensee must specify the appropriate regulatory basis. Under 10 CFR 50.55a(s)(3)(i), the proposed alternative must be shown to provide an acceptatste level of quality and safety, i.e., essentially be equivalent to the original requiremont in terms of quality and safety. Under 10 CFR 50.55a(s)(3)(ii), the licensee must show that compliance with the original requirement results in a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Examp!as of hardship and/or unusual difficulty include, but are not limited to excessive radiation exposure, disassembly of components solely to provide access for examinations, and development of sophisticated tooling that would result in only minimalincreases in examination coverage.

In accordance with 10 CFR 50.55a(g)(5)(iii), a licensee may submit a request for relief from ASME Code requirements. If a licensee determines that conformance with certain ASME Code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in Section 50.4, information to support that determination. Wher, a licensee determines that an inservice inspection requirement is impractical, e.g., the system would have to be redesigned or a component would have to be replaced to enable inspection, the licensee should cite this part of CFR to support the criteria for evaluation The NRC may, giving due consideration to the burden placed on the Heensee, impose an alternative examination requirement.

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4 SNC should review the submitted requests for relief and provide the required references to ensure that each request for relief is evaluated in accordance with the appropriate criteda.

H. Request for Rollef RR 6,7,9,17. Several requests for relief were submitted based on estimated examination coverages. For example, in RR 6, SNC noted that the examination coverage listed in the request for relief is based upon several factors, including NDE data sheet evaluations, design drawing configurations, field observations, and measurements of the components. In addition, SNC stated that *Due to these factors, examination coverage is given as a range for these examinations. Actual examination results should fall within the ranges listed; however, variations outside this range are possible and this relief request would be modified only if the values are significantly out of range.'

The Code requires that all examinations be performed to the extent practical. The licensee provioes a ' range' of coverage, making it difficult to determine if the licensee has met or exceeded previous examination coverages. Relief cannot be granted hased upon a range of examination coverage. For the requests currently being evaluateu,if the actual coverages do not meet or exceed the actual coverages previously obtained, ths licensee will be required to resubmit for relief. Therefore, the licensee should provic'e the actual coverages expected or consider submitting the subject relief requests following the examinations when actual coverages cen be calculated. Provide information as to the course of action SNC proposes to take regarding examination coverages.

l. Request for Relief RR 9. It is unclear which welds are affected by this relief request.

Are the four branch connections listed in RR 9 the only compononts for which relief is requested? Identify all branch connections that will be affected by this request for relief.

J. Request for Rollef RR 14. IWC 1222, of the 1989 Addenda of ASME Section XI, exempts vessels, pumps, and valves and their connections in piping NPS 4 and smaller within systems (or portions of systems) other than residual heat removal (RHR),

emergency core cooling (ECC), and containment heat removal (CHR) systems, with the followiq note: 'In piping" is defined as having a cumulative inlet and a cumulative outlet pipe cross sectional area neither of which exceeds the nominal outside diameter (OD) cross sectional area of the designated size.

Provide information concerning the components listed in Request for Relief RR 14, including the si: es of inlet and outlet lines. Confirm that the cumulative inlet or the cumulative outlet pipe cross sectional areas do not exceed the nominal OD cross +ectional area of the designated size, it appears from drawing D175039, sheet 2 of 7, that the Volume Control Tank may have at least three 3-inch inlet lines. This would make the cumulative inlet cross sectional area greater than the nominal OD cross-sectional area of the designated size. Consequently,it would not be exempt from examination.

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Also confirm that the subject components are within systems (or portions of systems) other than the RHR, ECC, and CHR systems. Provide the necessary information to verify that the subject components comply with the 1989 Addenda of ASME Section XI, IWC.1222 K. Request for Rollef RR 17, It appears from Attachment 171 that, for reflectors transverse to the shell-to-flange weld, a greater volume can be examined than the approximate 30% coverage (scanning from shell side of weld) stated in RR 17 for reflectors transverse to the shell-to flange weld. RR 17 identifies the limitations for this examination as the sharp taper on the flange side and the configuration of the examination sled. What specific techniques or equipment modifications have been employed to maximize the percentage of volume examined?

L. Request for Rollef RR 18. RR 18 proposes an attemative schedule for reactor pressure vessel nozzle examinations. Code Case N 521, Altomat/ve Rules for Deferral ofInspection*: of Nozzle to Vessel Weld' Inside Radius Sections, and Nozzle-to-Safe End Wolds of e Pressudzed Water Re Vessel, has been found acceptable for use with the following conditions: (a) no inst ,ce repairs or replacements by welding have ever been performed on any of the subject areas;(b) none of the subject areas cor,tain identified flaws o elevant conditions that currently require successive inspections in accordance with IWB 2420(b); and (c) the unit is not in the first interval. An additional requirement imposed by the NRC is that all subject areas be scheduled for examination such that the new sequence of examinations will not exceed 10 years beiwoen examinations. Confirm that all of tne above conditions will be met.

M. Request for Rollef RR 23. Pursuant to 10 CFR 50.55a(a)(3)(l), SNC proposed to use Code Case N 566, Corrective Action for Leakage at Bolted Connections,Section XI, Olvision 1, in lieu of the corrective measures specified by the Code for leakage at bolted connections. This Code Case is currently being reviewed by the NRC staff and, as of yet, not been found acceptable as written. The NRC has approved the use of alternatives to the requirements of IWA 5250(a)(2) proposed by other licensees that have included a detailed and well defined evaluation of the botting and the bolted connection. Considering that Code Case N 566, as currently written, will probably not be authorized at this time, discuss the intended action regarding use of this Code Case at Farley, N. Verify that there are no requests for relief in addition to those submitted, if additional requests for relief are required, SNC should submit them for staff review.

The schedule for timely completion of this review requires that SNC provide, by the specified date, the above requested information and/or clarifications regarding the Joseph M. Farley Nuclear Plant, Unit 1, Third 10-Year IntervalInservice inspection (ISl} Program Plan and the Joseph M. Farley Nuclear Plant, Unit 2, Updated Inservice Inspection (ISI) Program Plan, Revision 0.

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