ML20101A434

Safety Evaluation Supporting Amends 95 & 89 to Licenses NPF-35 & NPF-52,respectively

ML20101A434
Person / Time
Site:	Catawba
Issue date:	04/14/1992
From:	Office of Nuclear Reactor Regulation
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ML20101A430	List: ML20101A428 ML20101A434
References
NUDOCS 9204220144
Download: ML20101A434 (5)

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UNITED STATES l

NUCLEAR REGULATORY COMMISSION

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o W ASHINGloN D C. 20556 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR RfACTOR REGULATION RELATED T0fENDMENT NO. 96 TO FACILITY OPERATING LICENSE NpF-35 AND AMENOMENT NO. 89 TO FACILITY OPERATING LICENSE NPF-52 DUKE POWER COMPA,NL ET AL, CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 1.0 INTRODUClIOJ On June 25, 1990, the Nuclear Regulatory Commission (NRC) issued Generic Letter 90-06, " Resolution Of Generic Issue 70, ' Power-Operated Relief Valve and Block Valve Reliability,' and Generic issue 94, ' Additional Low-Tempcratw e Overpressure Protection for Light-Water Reactors ' Fursuant to 10 CFR 50.54(f)."

.The generic letter represented +he technical resolution of the aLove mentioned generic issues.

Generic Issue 70, " Power-Operated Relief Valve and Block Valve Reliability,"

involves the evaluation of the reliability of power-operated relief valves (PORVs) and block valves, and their safety significance in PWR plants. The generic letter discussed how PORVs are increasingly being relied on to perform safety-related functions and the corresponding need to improve the reliability of both PORVs and their associated clock valves. Proposed staff positions and improvements to the plant, technical specifications were recommended to be implemeated at all affected facilities. This issue is applicable to all Westinghouse, Babenck & Wilcox, and Combustion Engineering designed facilities with PORVs.

Generic Issue 94, " Additional low-Temperature Overpressure Protection for Light-Water Reactors," addresses concern'. with the implementation of the requirements set forth in the resolution of Unresolved Safety Issue (USI)

A-26. " Reactor Yessel Pressure Transient Protection (0verpressure Protection) "

The generic letter discussed the continuing occurrence of overpressure events and the need to further restrict the allowed outage time for a low-temperature overpressure protection channel in operating modes 4, 5, and 6.

This issue is only applicable to Westinghouse and Combustion Engineering facilities.

By letter dated May 9,1991, Duke Power Company, et al. (the licensee), proposed changes to the Catawba Nuclear Station, Units 1 and 2, Technical Specifications in response to Generic letter 90-06.

By letter dated December 18, 1991, the HRC provided clarifications and requested revisions to the licensee's submittal.

By letter dated February 6, 1992, the licensee responded with revisions that are consistent with the guidance of GL 90-06.

The February 6, 1992, letter provided c.arifying information that did not change the in. M 1 proposed no significant bezards consideration determination.

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.4' 2.1 Evaluation For Generic Issue 70 The actions ~' proposed by the NRC staff to improve the reliability of PORVs and block valves represent a substantial increase in overall protection of the public health and safety-aLd a. determination has been mau that the attendant costs are justified in view of this increased protection. The technical-find'ngs and the regulatory analysis related to Generic issue 70 are discussed in NUPEG-1316. " Technical Findings and Regulatory Analysis Related to Generic Issue 70, ' Evaluation of Power-Operated Relief Valve Reliability in PWR Nuclear Power Plants.'"

The Technical Specification (TS) changes in response to Generic issue 70,

" Power-Operated Relief Valve and Block Valve Reliability," consist of changes to TS 3/4.4.4, Relief Valves. An assessment of the proposed-TS against the model TS of Generic letter (GL) 90-06 for a Westinghouse plant with three PORV's follows.

Action statement a. is changed.to require that power be maintainea to the block valves when they are closed due to excessive PORY leakage.

Action statements a., b.,

c., and d. have been modified such that they terminate in HOT SHUTDOWN within six hours of the preceding action instead of terminating in COLD SHUTDOWN within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of the preceding action.

Action statement b. is changed to include the case where one or two PORVs L

(versus one before) are inoperable.

Action statement c. is changed to require L

.that at least one PORV must be restored, etc., with three PORVs inoperable L

instead of requiring each PORV to be restored, etc., when more than one is inoperable.

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.The licensee states that the change submitted for action statement d deviates slightly from the guidance in the GL in'that the. action statement only applies L

when the block valves are inoperable and not closed (per the addition of the phrase L

"and not closed"). The licensee considers that if the block valver, are inoperable while closed, then the PORV flow path itself would be considered to be inoperable, and.accordingly, action statement b, or c. would govern the required action.

. Action statement d. a so deviates from the GL in the directions for positioning l

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of the PORV switches in the event-of inoperable block valves (s). The G1. guidance L

,was to place the PORY (with an operable block valve) in manual control to preclude E

its automatic opening and subsequent potential for a stuck-open PORV. The Catawba PORV switches are labeled "open,"'"close," and " auto" so the license submits that-

-its proposal to place the PORV switches in the "close" position in such circumstances will likewise preclude automatic PORV opening and the subsequent potential for a stuck-open PORV when the' block valve is inoperable.and not closed.

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,The licensee's initially proposed Surveillance Requirement (SR),4.4.4.1 for

. operating-the PORY through one complete cycle of full travel did not include

. the stipulation. that this be done in MODE" 3 or 4.

The licensee stated that it does stroke the valves during MODE 4, t concludes that it would not be a'ppropriate to include an SR for_a MODE 4 action in this TS since the TS's applicability is only to MODES 1, 2, and 3.

This was addressed by a letter l

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. from the NRC staff date;l December 18, 1991. The licensea's response dated February 6, 1992, indicates that this testing will be done at temperatures greater than 200*F which is consistent with entry into MODE 4 condicions.

This is an acceptable response to this concern.

The licensee's proposed SR 4.4.4.1 does not require operating the solenoid air control valves and check valves on accumulators in PORV control systems through a complete cycle of full travel. This is because the action required by SR 4.4.4.3., fully stroking the PORVs while aligned to the einergency nitrogen supply, cycles the necessary valves. Therefore, the liceraee did not expand SR 4.4.4.1 to include this requirement, The guidance contained in the GL for SR 4.4.4.3 indicates that motive and control power for the PORVs and block valves should be manually transferred from the normal to the emergency power bus. This would be directly applicable to a design wherein non-safety related electrical power supplies for both motive and control power are provided for these valves. However, the Catawba PORVs are air operated; the block valves are electrically powered from an essential (emergency or safety related) bus, and control power is from essential l

sources for the PORV and the block valves. The Catawba SR 4.4.4.3, as currently written, appropriately addresses the PORY motive power transfer from normal (air) to the emergency (nitrogen) supply to demonstrate operability of the emergency nitrogen supply. Since the block vahes* motive and control power is normally from essential electrical power, their inclusion in SR 4.4.4.3 is extraneous anu the licensee has proposed its removal from SR 4.4.4.L The NRC staff has reviewed the licensee's proposed modifications to the Catawba Nuclear Station Technical Specifications. Since the proposed modifications are consistent with the staff's position previously stated in the GL and found to be justified in the above mentioned regulatory analysis, the staff finds the proposed modifications to be acceptable.

The licensee has also expanded the BASES Section 3/4.4.4 to identify the major function of the PORVs and block valves as follows:

1)

Hanual control of Reactor Coolant System pressure following accidents, 2)

Maintaining teactor coolant pressure boundary integrity by contro11irJ

1eakage, 3)

Hanual control of block valves to isolate and unblock PORVs (for manual pressure control and t'or controlling PORY seat leakage),

4)

Automatic control of Reactor Coolant System pressure, except for limited periods when the PORV has been isolated due to excessive seat leakage and except for limited periods where the PORV and/or block valve -18 closed because of testing and is fully capable of being returned to its normal alignment at any time, provided that this cyclution is covered by an approved procedure. This is a function '3at reduces challenges to the code safety valves for overpressurization events.

5)

Manual control of block valves to isolate a stuck-open PORV.

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1 These expanded BASES are consistent with the guidance of GL 90-06, 2.2 Evaluation For Generic issue 94 The actions.' proposed by the NRC staff improve the availability of the

. low-temperature overpressure protection (LTOP) system represents 4 substantial

increase in the overall protection of the public health and safety and a l

determination has been made that the attendhnt costs are justified in view of this increased protecti_on. ' The technical findings and the regulatory analysis related to Generic Issue 94 are discussed in UUREG-1326, " Regulatory Analysis for the Resolution of Generic _ lssue 94, ' Additior.a1 Low-Temperature Overpressure Protection for Light-Water Reactors.'"

4 The TS changes in response to Generic Issue 94, " Additional Low-Temperature Overpressure Protection for light Water Reactors," include changes to

- TS 3/4.4.9.3, " Overpressure Protection Systems." An assessment of the proposed TS against the model TS of GL 91-06 for a Westinghouse plant follows.

The licensee notes that the GL TS proposes that the APPLICABILITY af the limiting Condition for_ Operating?(LCO) for TS 3.4.9.3 be changed to exclude MODE 6 when the Reactor Coolant System (RCS) is adequately vented and that the depressurizing

'and venting of the RCS not be classified as'An overpressure protection System.

L The GL'also proposes that an additional action statement be added to specify verifying the vent pathway when the RCS_is depressurized and vented. The licensee concludes that this proposed structure appears inappropriate, because once'the RCS is vented, LCO 3.4.9.3 woufd no lenger apply and the action statement requiring vecification of the vert pathway would, therefore, not have to be entered.

For.this reasoni the.licensN proposed that.the present7 structure

-of the Catawba TS be maintained-in that'the depressurizing and venting-of the RCS will continue to be classified as an overpressure protection system and the H

requirement to verify-the vent pathway when the system is depressurized and

~_ vented will continue to be. governed by SR 4.4.9.3.2.

The NRC staff has considered the licensee's proposal and agrees with it since it would not be consistent with the intent of the GL to fail' to verify the vent-pathway when;the vent in being used for overpressure protection. Therefore,

- the TSs proposed by. the licensee in its item 3.4.9.3.b and 4.4.9.3;2 are acceptable.-

The licensee preposes to change the language of the APPLICABILITY statement from

...with the-reactor vessel-head on." to "...when the head is on the reactor

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vessel.," consistent with the language of the GL. This is-acceptable.

Action statement a. is proposed to be modified to clarify that it is only applicable in MODE 4. - This is consistent with the guidance in the GL and is acceptable.

~ Action Lstatement b. is added to reduce the allowable outage time f or an inoparable-PORY in MODES 5 or 6 from 7 days to 24-hours. This is consistent with-a key position of GL 90-06 for the resolution of Generic Iss>e 94 and is acceptable.

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. Action statement a., new statement b.,

and renumbered statement c. are clarified by inclusion of the words "... complete depressurization and venting of..." in lieu of "...depressurize and vent..."

This clarifies that these actions must be completed within the specific period. This clarification proposed by the licensee is acceptable.

The licensee proposes to simplify SR 4.4M.3 by removing requirements that exist beca'se of general requirements applicable to all surveillance requirements as specified in Section 4.0 of the TS.

This is consistent with GL 90-06 guidance and is acceptable, lhe NRC staff has reviewed the licensee's proposed modifications to the Catawba Nuclear Station Technical Specifications. Since the proposed modifications are consistent with the staff's position previously stated in the generic letter and justified in the above mentioned regulatory analysis, the staff finds the propose <1 modifications to be acceptable.

3.0 STATE C00. !TATION In accordance with the Comission's regulations, the Scuth Carolina Strie official was notified of the proposed issuance of the amendmcars. The State official had no commento 4.0 EllVIRONMENTAL,C,0NSIDEP.AT10N The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in incivio9al or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that tue amendaents involve no significant hazards consideration, and there has been no public coment on such finding (56 "R 31433). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issaance of the amendments.

5.0 CONCLUSION

The Comission has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activitio will be conducted in comt,11ance w;th the Comission's regulations, end (3) the issuance of the amendments will not be inimical to the comon refense and security or to the health and safety of the public.

Principal Contributor:

C. Liang, SRX8 G. Hamer, EMEB R. Martin, PDii-3 Date: April 'i4.1992