ML20202D960

Forwards Request for Relief 97-04 from Requirements of ASME B&PV Code.Request Is Submitted in Order to Seek Relief from Performing Immediate Acceptable ASME Code Repair of Valves 1/2RN-291 & 1/2RN-351

ML20202D960
Person / Time
Site:	Catawba
Issue date:	11/26/1997
From:	Gordon Peterson DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9712050199
Download: ML20202D960 (8)

Text

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g' (~% Duke Power Company j, ) A ta t=v We 6L] r Causse Nudar Suron d '* We 4800 Cornord Road  ;

York SC 29745

, c ,y n,p,,,,,,, (S03) 8314231 wnct y,., y , y , . (803) 831 M26ut November 26, 1997 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 t

Subject:

Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Request for Relief Number 97-04 Relief Request from Immediate ASME Code Repair of Valves 1/2RN-291 and 1/2RN-351 Pursuant to 10 CFR 50.55a(g) (5) (iii), please find attached Request for Felief 97-04 from the requirements of the ASME  !

Boi3er and Pressure' Vessel Code. This request is being .

submitted in order to seek relief from performing an immediate acceptable ASME Code repair of the subject valves.

The attachment to this relief request includes all information necussary to ensure timely processing of this request.

Catawba's intention to submit this relief request has already been discussed with Mr. P.S. Tam of your staff. If you have uny questions concerning this material, please call L.J. Rudy ,

at (803) 831-3084.

Very truly yo rs, hg (W JM s D Gary R.-Peterson Ix7R/s Attachment. g 9712050199 971126

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Document Control Desk Page 2 November 26, 1997 xc (with attachment):

L.A. Reyes, Regional Administrator Region II D.J. Roberts, Senior Resident Inspector P.S. Tam, Senior Project Manager ONRR 1

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Attachment i Request for Relief 97-04 i Relief Request from Inunediate ASME Codo Repair of Valves 1/2RN-291 and 1/2RN-351 I. Identify the component for which relief is requested:

a) Name and number as given in the UFSAR:

Nuclear Servicu Water (RN) Syst.em valves 1RN-291, 2RN-291, 1RN-351, and 2RN-351 are the component cooling system heat exchanger cooling water outlet  :

I flow control valves.

b) Description of function of components i

These valves control component cooling water temperature by throttling the flow of RN water through the component cooling water system heat i exchangers. One valve per unit normally throttles RN flow in the temperature control mode on the operating component cooling water train, while the valve on the opposite non-operating train is ucually full open to serve as a RN pump miniflow protection lir e. The train functions are switched approximately every two to three weeks. Valves 1/2RN-291 are "A" Train valves, whila valves 1/2RN- ,

351 are "B" Train valves.

The valves are 12-inch Fisher Controls pneumatically-operated V-ball valves. The "V" is a notch designed into the ball to allow for fine throttling control at low flow conditions.

c) ASME Section III Code Class or ASME Section XI Code Class:

These valves are 300 lb ANSI Class ASME Section III Class 3 valves with carbon steel bodies.

d) ASME Section III:

ASME Section III, Subsection ND e) valve testing:

These valves are tested quarterly per the Catawba Inservice Testing (IST) Program by stroking in the

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s closed-to-open direction. No relief from any

. testing requirements is being proposed in this request for relief.

II. specitically identify the AsNE Code requirement for which relief is desired: l Problem

Description:

Valves 1RN-291 and 1RN-351 have thinned areas in the body adjacent to the "V" notch that has pitted areas where through-wall leakage has occurred.

While operating in the temperature control mode at low cooling flow demand, the valve in alnost closed and the '

"V" notch passes most of the flow required. This ball position allows the flow to impinge on the valve body in  ;

a narrow area which appears to have set up a corrosion-assisted flow erosion degradation area. Both of the Unit 1 valves have experienced through-wall body leakage as a  ;

result. Removable rubber housekeeping patches have been.

installed on the valves to prevent further leakage. The wall thinning area and thickness have been determined by detailed mapping using ultrasonic testing. The valves have been installed since initial unit startup in 1985, without any repairs to this body area. Valves 2RN-291 and 2RN-351 hr.t) also been examined using ultrasonic testing and also have thinned areas, but no through-wall body leakage.

Requested Area of Relief: Article IWA-4000 of Section y of the ASME Code describes the code repair procedures, code repair requires the removal of the flaw and a subsequent weld repair. Catawba proposes to perform a code repair on both of the Unit i valves during the end-of-cycle 10 refueling outage scheduled to begin in late November 1997. The code repair will consist of either a valve body replacement or a welded repair per ASME Code requirements. The Unit 2 valves will undergo a code ,

repair or valve body replacement during the end-of-cycle 9 refueling outage scheduled to begin in September 1998.

No other non-code repairs are proposed for either unit's valves prior to the respective refueling outages.

III. Provide information to support the determination that relief from the requirement in II above is necessary (i.e., burden):

Performing a code repair would require removal from.

service of each component cooling water system train, as well as rendering inoperable every system that relies on component cooling. These include residual heat

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removal / low-head safety iniection, intermediate-head .

- =afety injection, centrifugal charging /high-head safety  :

,. '.' tion, containment spray, auxiliary feedwater, and l t r.e a xiltery shutdown panels. This action would result  ;

in a lo." of safety system availability that is not

• c asensul to with the nature of the subject RN valve ilswo. L $ work necessary to effect a code repair of the  ;

erbject vos es could not be completed within the l technica2 aecification allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />  ;

v e (Le at _uted component cooling water system tra in. f The subb t valves are Idrge, air-operated valves that

-d ? has to be removed from the piping using rigging and tut di iasembled to make the body available for weld repair. The tocal repair area is large and will necessitate multiple wold passes. The ASME Code did not require volumetric non-destructive examination during

manufacturing; therefore, it is not uncommon to find porosity which must be repaired. The actual scope of the required weld work can therefore be uncertain. After-repair, the valve will require actuator replacement and control setup. with associated retest and functional  :

test. The total scope of this work would most likely exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

IV. Specif/ the alternate inservice testing / examination that will be performed in lieu of the ASME Code Section XI requirements:  ;

No relief is being proposed from any ASME Code Section XI inservica testing or inspection requirements. Due to the '

short time interval nr tf1 the beginning of the Unit 1 end-of-cycle 10 refueling outage, no additional inspections of the Unit i var /es will be performed prior to performing a code repair of these valves. The Unit 2 valves will bn inspected using ultrasonic techniques.

Work orders for the Unit 2 valves are already written end ultrasonic testing is scheduled for February 1.998 The information from this inspec; ion will be evaluated in order to determine if further inspection is required prirr to the nett outage. No further evaluation is expected to be required concerning structural integrity.

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V. Provide an explanation as to why the alternate proposed

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• inservice testing / examination will provide an acceptable 1crel of quality and safety and not reduce the levol of public health and safety:

As indicated previously, no relief is being proposed from ,

any ASME Code Section XI inservice testing or inspection requirements. Catawba will continue to test these valves per IST Program requirements. Catawba will continue to monitor the status of the Unit 2 valves until the end-of- ,

cycle 9 refueling outage next year. Degradation of these valves is not expected to increase substantially over this time period, as the flaws developed gradually since initial unit startup. Should the Unit 2 valves develop through-wall leakage prior to the end-of-cycle 9 refueling outage, Catawba will evaluate utilizing housekeeping patches similar to those utilized on the Unit i valves. A supplement to this request for relief will be made only if a non-code repair is utilized which is different from that employed on Unit 1.

VI. Provide a schedule for implementation of the inservice inspection described in IV above As previously noted, all norme.1 inservice inspections will continue to be performed. The Unit 2 valves do not have a through-wall leak and have been satisfactorily analyzed in their current condition. Further ultrnsonic testing will be done in February 1998 to ensure that the  ;

degradation rate has not changed beyond the scope of the analyzed condition. No further testing will be performed prior to the Unit 2 outage unless conditions different than expected are encountered.

SUNNARY OF STRUCTURAL INTEGRITY AND FLAW EVALUATIONS (GENERIC LETTER 90-05 EVALUATION) ,

Pour separate calculations were performed which documented the Generic Letter 90-05 Evaluations for valves 1RN-291, 1RN-351, 2RN-291, and 2RN-351. The calculations were performed using the "through-wall flaw" approach of Generic Letter 90-05. The pinhole leaks on the Unit i valves and the pit type flaws on the Unit 2 valves were evel uted using this methodology. Since the valve bodies are me.de of carbon steel, the critical stress  ;

intensity factor K will be 35.0 ksiVin per Generic Letter 90-05.

The actual flaw size produced a stress intensity factor K for each valve as follows:

Valve K (ksiVin) 1RN-291 29.47 1 1RN-351 32.21 2RN-291 6.52 2RN-351 17.51 The stresses in the valve bodies were also determined to be acceptable for tbn flawed conditions. Based on these evaluations, the structural integrity of these valves in their current conditions wa- determined acceptable, f

These calculations are available for NRC inspection.

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LEAKAGE EVALUATION AND OVERALL IMPACT ON RN SYSTEM I The combined leakage from both 1RN-291 and 1RN-351 is less than 0.5 gallons per minute. This leakage rate does not have a significant effect on the RN system flow balance, Standby Nuclear Cervice Water Pond (SNSWP) inventory (the SNSWP is the assured long-term source of RN water during accident conditions, i.e.,

the ultimate heat sink), or auxiliary building flooding calculations. Specifically:

1. Regarding the impact on the RN system flow balance, the affected valves are on the outlet side of the component cooling heat exchangers, so the flow to these heat exchangers is not being degraded. Other safety related loads served by the RN system are in parallel with the component cooling heat exchanger flowpaths and therefore are not being degraded either.

2. Regarding the impact on the SNSWP inventory, the inventory and surface area of the pond is not significantly affected. The inventory loss and corresponding change in surface area and level over a thirty-day period is negligible relative to the approximately 145 million gallons of inventory and approximately 39 acres of surf ace e.rea.

3. Regarding the impact on the auxiliary building flooding calculations, the leakage rate is negligible relative to the unidentified leakage rate assumed in the auxiliary building flooding analysis.

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