ML20059A839

Forwards Response to 931018 RAI Re Monticello Third 10-Year ISI Program & Requests for Relief

ML20059A839
Person / Time
Site:	Monticello
Issue date:	12/20/1993
From:	Richard Anderson NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M82545, NUDOCS 9401030161
Download: ML20059A839 (14)

Text

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. Northem States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401-1927 ,

Telephone (612) 330 5500 December 20, 1993 10 CFR Part 50 Section 50.55a US Nuclear Regulatory Commission Attnt Document Control Desk Washington, D.C. 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Response to Request for Additional Information Concerning the Monticello Third 10-Year Interval Inservice Inspection Procram and Associated Recuests for Relief (Tac No. M82545)

By letter dated October 18, 1993 from Marsha Gamberoni (NRC) to Roger O.

Anderson (NSP), the NRC staff requested additional information concerning Revision 1 of the Monticello third 10-Year interval Inservice Inspection Program and associated Requests for Relief. .Our responses to the questions contained in your October 18 Request for Additional Information are provided in the attached document.

This letter contains the following new NRC commitments:

1. Upon receipt of NRC approval of the program, we will issue ~ revised relief request No. 7 (as well as any other program changes that result from NRC review) to all ISI program manual holders.

, 2. If the Third Ten-Year ISI Program is revised in the future for other reasons, we will also consider making changes to clarify program intent relative to Code Cases N-416 & N-498, as well as the reference to the 74S75 version of the Code.

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9401030161 931220 if PDR ADDCK 05000263 F p .. PDR t

r USNRC NORTHERN STATES POWER COMPANY December EJ, 1993 Page 2 Please contact Terry Coss, Sr Licensing Engineer, at (612) 295-1449 if you require additional information.

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R er O Anderson Director Licensing and Management Issues cc: Regional Administrator-III, NRC NRR Project Manager, NRC Resident Inspector, NRC State of Minnesota, Attn: Kris Sanda J Silberg Mr Boyd W. Brown,.EG&G Idaho, Inc.

Attachment 1: Response to October 18, 1993 RAI Concerning the Monticello Third 10-Year ISI Program Enclosures (1) Revision 1 to ISI Relief Request No. 7 (3 pages)

(2) Excerpt from "CRD Operation and Maintenance Guidelines", GE [

Co. and Drawings / Sketches of CRD Housing Joints (5 pages) t I

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' December 20, 1993 Attachment 1 I Page 1 Attachment 1 Response to October 18, 1993 RAI Concerning the Monticello Third 10-Year ISI Program Our responses to the questions contained in your October 18, 1993 Request for Li Additional Information (RAI) are provided below. The questions. contained in the RAI have been paraphrased and are presented herein to facilitate review of our responses  ;

Ouestion 2.As Auamented Insoection Samolet l It appears that the entire 7.5% sample of welds selected for (augmented volumetric) examination in the core spray system falls l

consecutively from the class break. A more random distribution ,

would be_a better indicator of the system's overall inservice condition. Please provide justification for having the total sample of welds fall consecutively from the class break. Include the isometric drawings that are not included in the available ISI program showing the sample distribution of the entire core spray system.

The ASME Code Class 2 piping downstream'of the Core Spray

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NSP Response system pump contains 44 welds for Loop A'and 48 welds for Loop B. The majority of the welds (38 of 44 welds in loop A and 42 of 48 welds in +

loop B) are exempt from inspection per IWC-2500-1 because the pipe wall .;

thickness is less.than 3/8 inch, however, these welds are still factored into the total population when calculating the 7.5% inspection sample '

size. .Therefore, the sample size equals 4 welds per loop. The welds that are included in the population, but are exempt from inspection, have been identified in-the Monticello Plant Third 10-Year Interval .i Inservice Inspection Program Plan. These welde-have been given the~ >

Section XI classification of C X.XX in the Third Ten-Year Interval'In- ,

Service Inspection Program Plan and Schedule, and are identified on the following program drawings: }

Loop h Looo B >

ISI-13142-26-C ISI-13142-31-C l ISI-13142-26-B ISI-13142-31-B As noted above, there are only 6 welds in each loop that are suitable-for inclusion in the inspection sample, and 4 of the 6 welds in each 1 loop must be inspected to satisfy the 7.5% criteria. It is for this reason that the sample distribution cannot be more random. These welds were selected for augmented volumetric examination during the second 10- ,

Year Interval.

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December 20, 1993 i Attachment 1 Page 2 i

Ouestion 2.Bt Relief Reauest No. 7. Leakaae at Bolted Connectionst Relief Request No. 7 (Rev. 0) seeks relief from the requirement to remove the bolting at a leaking bolted connection. The control  ;

tod drive (CRD) housings were cited in the Basis for Relief as having a bolted connection that may leak slightly during a pressure test, but does not leak during power operation. To support this position, an evaluation of CRD housing leakage conditions (with input from General Electric) and a subsequent VT-2 acceptance criterion established for this system were cited.

This information may support a determination that minor leakage at .;

CRD housing bolted connections is acceptable during VT-2 '

examination. However, more information is necessary. Describe the impracticality associated with the code requirement for the CRD housings and submit the CRD housing leakage evaluation ' and VT-2 acceptance criteria for staff review. '

The remainder of this relief request appears to be generic in c nature--no burden associated with other connections is stated. '

Generic relief cannot be granted. Revise this relief requast to only L3clude CRD housing bolted connections or to address other situations where specific circumstances make the Code requirement impractlcal. '

NSP Response: We have prepared a revision to the subject relief request i (enclosed) such that it pertains specifically to CRD housing bolting and provides additional information to clarify the specific basis for -

requesting relief. We are also providing an excerpt of the applicable portion of a General Electric (CE) Company. document on this subject, as well as some drawings of the CRD housings, to facilitate your review.

The revised relief request will not be implemented or distributed until ,

NRC review of the Third Ten-Year ISI program is comp 1ete. Upon receipt of NRC approval of the program, we will issue the revised relief request to all ISI program manual holders. '

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We continue to believe that generic relief from this examination requirement is warranted based on ser determi..eilon th&L the-VT-3 bolting examination is generally of little or no practical'value. "

However, we were unable to develop a revised Relief Request.that would adequately address all systems and components covered by the ASME code in the time available. We view this issue as a potential cost Beneficial Licensing Action (CBLA) and may revisit the concept of a generic relief request at some future time.

r Ouestion 2.C Usaae of Code Cases N-416 and N-4982 Page 1,3-1 of the ISI Program contains a list of Source Documents with little or no explanation defining the extent of their use or their function. Regulatory Guide 1.147, Rev. 9, with Code Casos  ;

December 20, 1993 Attachment 1 Page 3 ,

N-416 and N-498, is referenced in Source Documents. This implies that these two code cases are being incorporated into the.

Monticello Third 10-Year Interval program. 'It should be noted that the staff considers the concurrent use of these two code cases _ unacceptable. Confirm that Code Cases N-416 and N-498 are not being used concurrently.

NSP Response: We understand that the two code cases cannot be used. ,

concurrently and will not do so. We will consider adding a clarifying l statement to this effect in the ISI Program if it is revised in the.

future for other reasons. Both Code cases were listed as Source Documents because the ANI had requested that our ISI Program identify. '

any Code cases that might be used. Listing both Code cases allowed us to comply with that request, yet defer the decision'as to which option ,

we would ultimately choose until such time as a decision was necessary.

As of this time, we have not yet made this decision, although it is considered unlikely that we would ever elect to use Code. Case N-416 over N-498.

Ouestion 2.D: Clarification of Avolication of 1974 Edition. Summer 1975 Addenda to Class 1 Comvonents:

10 CFR 50.SSa(b)(2)(ii) states that the extent of examination for Code Class 1 pipe velds may be determined by the requirements of .

Table IWB-2500 and Table IWB-2600 Category B-J of Stctioni XI in the 74S75. Please define what is meant by the statement " Portions of the 1974 Edition, Summer 1975 Addenda to the' Code are  ;

applicable to class 1 components" found in the Introduction, Page 1.2-1, of the Monticello Third 10-Year Interval Program.

NSP Response The statement was included in the Third 10-Yoar ISI' ,

Program because the extent of examination for Code Class I pipe welds is l determined by the 1974 Edition, Summer 1975 Addenda to the Code. No l other portion of the 74S75 Code will be used. We will consider clarifying this statement in the ISI Program if it is revised in the future for other reasons. j i

l Ouestion 2 E: Safety Classification of Main Steam Line Ploina a

Regulatory Guide (RG) 1.26 and Standard Review Plan (SRP) 3.2.2 l give guidance for quality Group Classification of componentn. In . ~

these documents, the main steam line from the outermost ^

containment isolation valve to the turbine stop and bypass valves, and connected piping up to and including the first. valve that is either normally closed or capable of automatic closure,:la to be 1 designated as Quality Group B or ASME Class 2. Based on RG 1.26 and SRP 3.2.2, it appears that this piping has been incorrectly classified as non-safety related. Provide technical ]'ustification I

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December 20, 1993 .!

Attachment 1 Page 4 for classifying this piping as non-safety related rather than N class 2 in accordance with RG 1.26 and SRP 3.2.2.

NSF Response The subject piping is classified as non-safety related because it performs no safety, function. The applicable 10 CFR Part 20 ,

and 10 CFR Part 100 licensing basis. analyses for Monticello take no  ;

credit for the integrity of the main steam system piping outboard'of the  ;

outboard Main Steam Isolation. Valves. The radiological consequences of l complete severance of a main steam line outboard of the isolation valves have been analyzed and found to be within acceptable limits. A, discussion of this analysis can be found in Chapter 14 of the Monticello.

Updated Safety Analysis Report (USAR).

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As noted in your Request for Information, Reg Guide 1.26 and SRP'3.2.2 )

are considered guidance documents only. The' licensing basis for  :

Monticello does not require compliance with RG 1.26 or SRP 3.2.2, nor -

has Moaticello ever committed to comply with these documents. I I

Nonetheless, the guidance of RG 1.26 was carefully considered when tne _

classification of the subject piping was determined. Prior to 1991, the l subject piping was conservatively treated as if it were safety related 'I as a quality program enhancement. The piping was included in the.First i I

lO-Year Interval ISI Program (and initially in the Second 10-Year' Interval ISI Program) and inspected as ASME Class 2.

In 1991, the issue of the proper classification of the main steam line piping outboard of the isolation valves was-revisited and it was determined that the disadvantages of continuing to treat this piping as safety related (increased cost.and complexity'of maintenance, parts procurement and modifications; increased personnel radiation exposure i and inspection costo due to inclusion in ISI program) outweighed any perceived benefits of continuing this practice. Thus,:the decision was '!

made to downgrade the piping classification to no:n-Safety Related. The I piping was subsequently dropped from the Secor4d 10-Year ISI Program and -  ;

was not included in our Third 10-Year'ISI program. I I I 1

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NORTHERN STATES POWER INSERVICE INSPECTION MONTICELLO 3RD INTERVAL EXAMINATION PLAN ISI Relief Rsquest No. 7 (Rev.1)

(Page 1 of 3) 1.eakage at Bolted CRD Housing Connections Systems Control Rod Drive Housing Bolted Joint Class: 1 Category: NA Item: NA Examination Recuirement from which Relief in soucht:

IWA-5250(a)(2): If leakage occurs at a bolted connection, the bolting shall be removed and VT-3 examined.

Basis for Relief:

lu CFR Part 50, Section 50.55a(a)(3), which states (in part):

" Proposed alternatives to the requirements of paragraphs (c), (d), (e),

(f), (g), and (h) of this section or portions thereof may be used when....

(ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in~the level of quality and safety."

The CRD (Control Tod Drive) housings are flanged connections beneath the reactor vessel that are used to secure the 121 CRD mechanisms in position below the vessel. Each of the 121 CRD to CRD housing bolted joints utilizes eight bolts, washers, and nuts to hold the CRD mechanism in position. The joint also utilizes three hollow metal o-rings to provide a water tight seal capable of withstanding full reactor pressure at normal operating temperatures.

The CRD housing joints are VT-2 examined as part of the periodic reactor pressure vessel Leakage and Hydrostatic pressure tests. These tests are conducted with the vessel temperature much less than the design operating temperature. For a typical test, the vessel temperature would be <212.*F, as compared to a normal operating temperature of about-540 *F. It is not unusual for these bolted joints to leak slightly during periodic reactor vessel pressure tests conducted at test temperatures below normal operating temperature. This is a condition identified in the original design of the connection by the Architect / Engineer, General Electric (GE). GE developed guidance to permit evaluation of a leaking CRD housing bolted connection over a period of time, while at test pressure, to determine whether the leak will stop once the vessel heats up to normal operating pressure. This leakage evaluation criteria is incorporated into the VT-2 tests for these joints.

NORTHERN STATES POWER INSERVICE INSPECTION MONTICELLO 3RD INTERVAL EXAMINATION PLAN q

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ISI Relief Request No. 7 (Rev.1) *

(Page 2 of 3)

Leakage at Bolted CRD Housing Connections  ;

Compliance with Code requirement IWA-5250(a)(2) represents a hardship (burden)

• in the case of the CRD housing bolted joints because .

1) Examining the bolting would involve the accumulation of ,

considerable personnel radiation exposure, since the work must be l performed in a relatively high dose rate area inside the drywell, l immediately below the reactor vessel. Typical shutdown-dose rates in the vicinity of the bolting flanges would be on the order of.50 to 100 mR/hr.

2) Since the reactor pressure vessel test is a critical path item, the additional time needed to depressurize the vessel, remove the bolting, perform the exam, and then repressurize the vessel to retest the joint would delay plant startup from an outage by an equivalent amount of time. The cost of such delays is significant, since it is estimated that the cost of extending the -

duration of an outage is $379,000 per day (including replacement power costs).

Compliance with Code requirement IWA-5250(a)(2) would not result in a compensating increase in quality or safety because:

1) CRD Housing joint leakage during (relatively) low temperature testing is not unexpected due to the design of the bolted joint.

I This joint is unusual in that it has hollow metal 0-rings that require the CRD housing bolts to be tightened within a specific torque range in order to function properly at normal operating temperature. Thus, the bolts cannot simply be tightened to stop leakage as might be done for a conventional gasketed joint.

2) As noted previously, CE developed guidance to evaluate any CRD housing leakage to determine if the leakage will persist at normal l operating temperature / pressure and should therefore be corrected.  !

Leakage that is found to be acceptable per the guidance is not considered adverse to quality or safety and need not be corrected l before startup. This type of analysis is consistent with Section XI Code paragraph IWB-3142, which allows analysis of the leakage ,

for acceptability.

3) Performance of the VT-3 bolting examination does not represent a corrective action for the joint leakage and will not reduce the likelihood of joint leakage upon retest. Therefore, the VT-3 l

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7 i- NORTHERN STATES POWER INSERVICE INSPECTION-L MONTICELLO-3RD INTERVAL EXAMINATION PLAN 1-r ISI Relief Request No. 7 (Rev.1)

(Page 3 of 3) h- Leakage at Bolted CRD Housing' Connections l

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bolting examination does not contribute to increased quality or safety. .j

4) The bolts in the CRD housing connection are periodically examined l:

when the joint is disassembled, per Table IWB-2500-1, Item B7.80.' ,

Four of the eight bolts on each housing joint were replaced with .

new bolts in 1991. It was also reported in General Electric SIL f 483 that only three uniformly distributed housing bolts are required to support the CRD mechanism. These factors provide a  :

high degree of confidence in the long term safety and integrity of the CRD housing joints.

Earlier Section XI Code editions invoked by Monticello's first and'second. ten- ,

year inspection interval programs did not include the subject examination requirement. A subsequent code revision which_has been approved for use by

! the NRC, the 1990 Addenda to the 1989 Edition, limits the exam requirement.to:

one bolt nearest the leak. Current Code committee activities' include an )

effort to write a code case that limits this examination even further to specific joints and. materials. All of these changes support the conclusion that this code requirement, as written, is overly restrictive and represents a hardship without a compensating increase in the level of quality or safety.

Alternative Reauirement:

Any leakage found at a CRD housing bolted joint during a periodic pressure

• test performed at a temperature much less than opercting temperature will be evaluated to determine whether it will stop leaking at operating temperature.

If this evaluation shows the leak will stop as temperature increases to normal operating temperature, no further action will be taken. The acceptance criteria will be based on guidance provided by General Electric and will be-included in the VT-2 tests for the joint (Note this criteria has been >

submitted for NRC review). If the leak is determined to be unacceptable and the joint is disassembled to correct the leak, one of the bolts will be VT-3 examined in accordance with the 1990 Addenda requirement.

Aporoval Status Not yet approved. Currently under review by the NRC.

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l "CR0 HOUSING FLANGE LEAKAGE II}

The fact that a flange joint leak exists on a newly installed CRD (either at initial CR0 installation or subsequent reinstallation following maintenance) does not mean that an adequate flange joint seal has not been achieved. Experience to date indicates that flange joints may leak when they are initially tightened.

However, if a decrease in l'eak rate is detectable in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of being subjected to pressures of 1,000 psig or greater, the joint will eventually stop leaking. A decreasing leak rate will also eventually seal without being internally pressurized, providing the flange bolts remain properly torqued. Recognizing these conditions, the following guide lines are established:

A) All drip type leakages which show a decreas-ing leak rate af ter being pressurized for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period at pressures greater than 1,000 psig do not require any corrective maintenance action. They should, however, be observed at later periods whenever the '

opportunity exists.

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8) Maintenance should be considered if a drip type leak at pressures of 1,000 psig or greater is constant or the leak rate shows an increasing rate over an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

C) Maintenance should be considered for any spray type leak from the flange joint."

(1) Excerpted from "CR0 Operation and Maintenance Guidelines",

by N. J. Biglieri, General Electric Co. , Nuclear Energy Division, San Jose, Calif. 8/30/70.

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