ML20077S011

Safety Evaluation Supporting Amend 25 to License NPF-8

ML20077S011
Person / Time
Site:	Farley
Issue date:	09/08/1983
From:	Office of Nuclear Reactor Regulation
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ML20077S010	List: ML20077S011
References
NUDOCS 8309210143
Download: ML20077S011 (5)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 25 TO FACILITY OPERATING LICENSE NO. NPF-8 ALABAMA POWER COMPANY 1

1 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO. 2 i

DOCKET NO. 50-364 i

j Introduction l

1 By letter dated August 10, 1982, the licensee requested a Technical Specifi-cation change that would permit using a modified Unit 1 Pressure Isolation Valve (PIV) Technical Specification allowable leakage for Unit 2.

We advised the licensee that the long term NRC staff review on a generic basis would not be completed in time for the first refueling outage on Unit 2.

For this

. reason, the licensee then proposed a one-time change by letter dated October 11, e

1982. License Amendment No. 20, dated November 24, 1982, approved the one-time proposal for the first refueling outage testing.

Subsequently, by letter dated December 23, 1982, the licensee proposed changes to Technical Specifications 4.4.7.2 and 4.4.7.2.2 for Units 1 and 2, respectively.

The licensee stated "that the Unit 2 leak test acceptance criteria of 1 gpm versus i

the Unit 1 criteria of 1 to 5 gpm (with certain limitations) has proven to be excessively restrictive without providing increased assurance of valve operability."

j We have continued our generic review and have recently developed revised generic leak rate criteria, less stringent than 1 gpm, which vary according to valve size; and is currently in the process of obtaining acceptance to use these criteria from the Committee to Review Generic Requirements (CRGR). After CRGR review it is the NRC staff's intent to permit licensees with recently licensed plants, such as Farley 2, to use the revised criteria in lieu of the original 1 gpm. Because it is anticipated that these criteria will not be accepted by CRGR before the September 1983 Farley 2 refueling outage; these criteria cannot be applied to Farley 2 for the testing to be performed at this outage.

In addition, the licensee proposed deletion of Surveillance Requirement 4.4.7.3.2 p

stating that it is superseded by Action Statement 3.4.7.3.2 and'is, therefore, not necessary. This proposal is identified in licensee letter dated July 8, 1983. Also, the licensee proposed another one-time change for the Unit 2 i

for the second refueling ' outage testing.

Our evaluation follows.

Discussion and Evaluation i

l-i The Unit 1 Technical Specifications allow for leakage rates of 1 to 5 gpm; I

however, the measured leak rate for any given test cannot reduce the difference between the results of the previous test and 5 gpm by more than 50%. The pro-posed change restricts the maximum leakage on 2" valves to 3 gpm, but retains this same indexing criteria. The original Unit 2 Technical Specification restricts leakage to 1 gpm for each valve, regardless of size.

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1 Conservative leak test criteria were established by the staff as a result of a concern which was brought to light by the Reactor Safety Study, WASH-1400.

The study indicated that the failure of two in-series valves which fonn the interface between high pressure and low pressure systems would almost surely

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.j result in an intersystem loss of coolant accident; and that the probability of such an event was unacceptably high.

Frequent independent tests of each valve was considered to be a relatively convenient method of reducing the j

probability of this type of failure.

The staff originally developed two sets of allowable leakage criteria; one i!

for new plants (1'gpm) and one for older plants (1-5 gpm with certain restric-tion); as it was felt that the newer valves would more easily meet the more stringent 1 gpm criterion.

The 1 to 5 gpm criterion is included in the Farley Unit 1 Technical Specifi-cations together with the 50% indexing provision noted above. This criterion

'i was ordered by the staff about two years ago to be effective for operating reactors.

For these older plants the staff has concluded that these valves had experienced numerous operating cycles and could not be expected to be in i

the "like new" condition, although the valves would be expected to fulfill their pressure isolation function.

The staff currently is in the process of drafting a proposal for revising the allowable leakage criteria for approval. A consultant, EG&G, Idaho, has 1

completed an initial reevaluation of the existing criteria, both theoretically and through a series of operating plant surveys.

EG&G has recommended that the staff consider allowing leak rates of 1/2 gpm per inch of nominal valve size, with a maximum of 5 gpm regardless of valve size.

EG&G has also recom-I mended that the same 50 percent indexing provision, as discussed above for Farley Unit 1, be. adopted for all plants. Also, the EG&G plant survey results do not appear to. support the staff's initial judgement that new plant valves can more easily meet the stringent 1 gpm criteria that has been required for recently licensed plant, such as Farley 2.

The staff has plans to propose the EG&G reconnended criteria for approval.

Use of these criteria will result in a somewhat reduced frequency of valve maintenance, i.e., lapping of valve seats, with no decrease in assurance that the valves would be expected to fulfill thair pressure isolation function.

At the time of the last refueling outage for Farley 2, i.e., the first refueling outage, the licensee requested a one time Technical Specification change to use the modified Unit 1 allowable leakages, discussed above, for the Farley 2 PIVs. Staff approval to use the modified Unit 1 allowable leakages for that outage was given.

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At that time the staff approval was based on:

1.

An extensive evaluation of actual leakage data accumulated over approximately two years of leak testing PIVs on Units 1 and 2 to j;

the two different criteria. The staff concurred with the licensee h

conclusion that the Farley plant specific accumulated data indicated that considerably more valve maintenance and related personnel

]o radiation exposure was required to meet the Unit 2 1 gpm leakage rate.

In addition the staff accepted licensee statements to the effect that no discernable differences in valve seating surfaces could be found, and no evidence of impending valve failures were found in any of the valves that failed either leakage rate.

a 2.

The Technical Specifications for both Units 1 and 2 require that h

leakage testing be performed during plant startup so that all valves will be tested after their last disturbance.

This licensee routinely leak tests the PIVs during each cooldown to refueling in an effort to determine if any pressure isolation valves may require J

maintenance. This is a precautionary measure voluntarily utilized d

by this licensee to increase the probability of successful leak test results during the return to power when the testing is performed on 1

the schedular " critical path." Valves that pass this test, or'are maintained if they do not pass this test, would be expected to pass the Technical Specification test during startup because significant valve degradation would not be expected to occur during the relatively short period of the refueling outage.

3.

The EG&G reconnended criteria changes discussed above, which at the time of the Farley 2 first refueling outage, were available to the staff in preliminary form.

4.

At staff request the licensee provided leak test data measured

i during the voluntary testing performed during cooldown for the Unit 2 first refueling outage.

During this testing only one valve failed its leak tests; 27 of the 35 valves had no leakage and the remaining 2

valves had ieak rates of less than 0.5 gpm. Therefore, during this U

" unofficial" testing only one out of the 35 PIVs failed to meet the original Unit 2 Technical Specification allowable leakage of 1 gpm and, it also failed the modified Unit i leakage allowance of 2.5 gpm.

After approval of the staff's proposal, it is the staff's intention to permit licensees with recently licensed plants, such as Farley 2, to use

t the EG&G recommended allowable leakage in lieu of the original 1 gpm.

Since these generic criteria changes are in the process of being made, approval of the licensee request for a permanent Technical Specification change to use the modified Unit 1 allowable leakages is not granted.

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' 1 Approval of the staff proposal is not expected until sometime after the Farley 2 second refueling outage, scheduled to begin in September 1983.

Although approval of the staff's proposal will not be obtained in time for the PIV testing to be performed during this outage, the staff recognizes

j the strong technical justification provided by the licensee, summarized above, for the Technical Specification change that was previously approved for the Farley 2 first refueling outage.

In addition, since that outage the licensee has provided by letter dated June 3,1983 from F. L. Clayton, Jr. to the Director of Nuclear Reactor Regulation; Attention: Mr. S. A. Varga, the results of the PIV leak testing performed during startup from the first refueling outage.

In response to the licensee's request for the permanent Technical Specifi-cation change, taking into account the additional valve test data provided and the status of the staff's generic criteria revision proposal, we are herein granting approval to extend the modified Unit i leakage allowance criteria, previously approved for the Unit 2 first refueling outage, to the second refueling outage testing.

It is expected that revised generic criteria will be available for all plants prior to the start of the Farley 2 third refueling (;utage.

l Revised Table 3.4-1 of the Farley 2 Technical Specifications, attached to this Safety Evaluation, specifies the allowable leakages that are applicable for i

each of the Farley 2 PIVs for testing to be performed during startup after the second refueling outage. The allowable leakages in the revised table were obtained by applying the modified Unit I leakage allowance criteria to the actual valve leakages as determined by the licensee during startup from

.i the first refueling outage.

With the expection of the two-inch check valves, the allowable lukages are all within the maximum allowable leakages that are determined by application of the staff's proposed revised generic criteria.

For the two inch check valves, application of the staff proposed revised criteria would limit allowable leakage to 1 gpm. As shown in the attached table, application of the modified Unit 1 criteria results in allowable leakages of 1.575 - 2.0 gpm for these valves depending upon what their tested leakage rate was during the startup from the first refueling outage.

Although these leakages are slightly less conservative than would be permitted if the staff proposal is adopted; we have concluded, based on our extensive review of the entire leak testing history of these valves to date, that appli-cation of these criteria for the Unit 2 second refueling outage will provide sufficient assurance the valves are capable of perfonning their pressure isolation function.

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i Summary i

j Based on our extensive review of the information provided to date by tt j

licensee for the Unit 1 and Unit 2 PIVs, review of recommendations fror.

the staff consultant for generic leak test criteria revisions and related li supporting multi-plant leak test data, the licensee's standard practice j

- of perfonning the leak tests during both plant shutdown and startup, and our expectation that no significant valve degradation would occur during the short period of the second refueling outage, the staff has concluded that the allowable leak rates specified in attached revised Table 3.4-1 are acceptable for the second refueling outage.

We have also reviewed the licensee's proposal relating to deleting Surveil-lance Requirement 4.4.7.3.2 for Unit 1.

We find that Action Statement 3.4.7.3 and Surveillance Requirement 4.4.7.3.2 are separate and distinct requirements and, therefore, one does not supersede the other. We consider the elimination of Surveillance Requirement 4.4.7.3.2 to have generic signif-france.

It was incorporated in plant Technical Specifications at the time of issuance of the Event V Orders.

It does not appear to result in any significant burden for APCo for Farley 1.

Therefore, until we can assess the generic implications of the requested deletion, this Technical Specifi-cation revision is denied.

Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmentai impact. Having made this determination, we have further concluded that the amendment involves an action which is insig-nificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

l-Conclusion

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We have 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, and (2) such activities 4

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will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

i Date: September 8,1983 4

l Principal Contributors:

l G. Hammer ll E. A. Reeves d

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