ML20237G706

Safety Evaluation Supporting Amends 60 & 49 to Licenses NPF-10 & NPF-15,respectively

ML20237G706
Person / Time
Site:	San Onofre
Issue date:	08/14/1987
From:	Office of Nuclear Reactor Regulation
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ML20237G687	List: ML20237G684 ML20237G706
References
NUDOCS 8709020375
Download: ML20237G706 (9)

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1 SAFETY EVALUATION BY THE OFFICE'0F NUCLEAR REACTOR REGULATION SUPPORTING AMENDMINT NO. 60 TOMCIL TY OPERATING L1 CENSE NPF-10 l

AND AMEffMENT NO. 49 TO FACIL?Tf 10PEMTING LICLhSE NPF-15 I

l SOUTHERN MIFORPIA EDISON t.CNPANY, ET AL'.

SAN ON0FRE NIELEAR GENERAT7NG STATION, UTITS 2 & 3 l

DUCKET NOS. 50-361 AND 50-362 l

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1.0 INTRODUCTION

Southern Califom'ia Edison Compeay (SCf), on behalf of itself and the other licerseer, San Diego Gas and Electric Company, The City of Rivo side, I

California, and The City of Anaheim, California, has submitted a nurdier

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I of applicaticas Mr license amendments fc6 San Onofre Nuclear Generating l

Station (SONGS), Units 2 and 3.

The NRC staf f's evaluatior, at three of l

these applications is described below.

2.0 DISCUSSION A.

PCN-84 1

This proposed change deletas Licente Condition 2.C(5) of the San Onofre l

l Nuclear Generating Statlen, Units 2 and 5 Operating Licenses, NPF-10 and i

i NPF-15, respectively. hete license conditions defined the requirements l

for the enviro.nmental goalification program for safety related electrical l

equipment. On February 23, 1903, tr.e Cumission issued a new regulation, l

l 10 CFR 50.49, which defines the requirements for environmental.qualifica-tion of safety related electrical equipment and supersedes th6 previously 1

1 issued SONGS 2 and 3 license conditions.

4 Specifically, License Condition 2.C(5)a requires thst fCE be in compliance with the provisions of NUREG-0588, " Interim Staff Pe ition on Environment d Qualification of Safety Related Equipment," Revision I dated July 1981.

10 CFR 50.49 encompasses NUREG-0588 and states that equipment qualified in accordance with NUREG-0588 need not be requalified in accordance wi W i

the provisions of 10 CFR 50.49. However, replacement 6.quipment must be qualified under the provisions of 10 CFR 50.49. License Condition a

2.C(5)b requires that complete auditable records, which describe the environmental qualification status of all safety related electrical equipment, be available and maintained and that such records be updated and maintained current as equipment' is ' replace) or further tested.

10 CFR 50.49 encompasses the record keeping requirements 6f this Mcense condition. License Condition 2.C(5)c requires < implementation of an environmental qualification maintenance procedures program. Additionally, License Cont'ition 2.C(5)d requires the imphmention erf an. impro.ved '

8709020375 870314 PDR ADOCK 05000361 I

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.ntal qualification surveillance program to detect age-related 1

q;w ation and any improved maintenance program procedures required by t

sutn e surveillance program.

10 CFR 50.49 encompasses these requirernents by requiring that qualified equipment will meet its specified performance requirements when it is subjected to the conditions predicted to be present when it must perform its safety function, up to the end of its aualified life, and requires that replacement equipment must also be quaH fied. By letters dated August 23, 1983, and April 10, 1985, SCE affinmd that maintenance program procedures had been implemented and that an' improved surveillance program had been implemented.

B.

FCN-183 l

Thi', proposed change revises Technical Specification 3/4.3.3.6 " Accident l

1 Monitoring Instrumentation." Technical Specification 3/4.3.3.6 defines numb @ of required charnels to be opera}ation, operability requirements, types of accident monitoring instrumen ble, actions to be taken in the event that the operability requirements are not met, and periodic surveillance testinJ to verify operability. The operability of post accident monitoring instrinnentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables ~following

'j an accident. Ice proposed change adds an additional type of accident nonitoring ins.trumentation subject to these requirements. Specifically, the proposed chdnge adds the reactor vessel level monitoring system 1

(RVLMS) to the technical specifications. The proposed change reflects the addition of the heated junction thermocouple (HJTC) system-reactor vessel level. monitoring system.

Two channels (f the HJTC system are required, one of which must be i

operable as a nJnimum.

Each channel ircludes eight sensors in an HJTC probe. A charr.el is considered to be operable if four or more sensors (one in the reict.or vessel upper head region and three sensors in the lower l

head region) are operable. Should these minimum operability requirements

.not be met, the proposed change defines actions to be taken.

If one channel is inoperable, the proposed change requires that channel to be restored to opcrable status within seven days if repairs are feasible tithout shutting down the reactor, or a special report be submitted to the Commission within the following 30 days which outlines the action taken, the cause of the inoperability and the plans and schedule for restoring the system to operable status.

If both channels are inoperable, the propcsed change will require that one or both of the channels be restored to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if repairs are feasible without shutting down the reactor, or that an alternate means of monitoring reactor vessel im entory te initiated, a special report be submitted outlining actions taken, the cause of the inoperability and plans and schedule for restoring the system to operable status, and that both channels be restored to' operable status at the next sche'dule'd refueling outage.

In addition, to verify operability of the system, the proposed change requires monthly channel checks and channel calibrations to be performed at refueling outage intervals.

, l C.

PCN-207 This proposed change revises Technical Specification (TS) 3/4.3.2, j

" Engineered Safety Feature Actuation System Instrumentation (ESFAS), and i

3/4.7.] ~, " Main Steam Isolation Valves (MSIV's)." TS 3/4.3.2 specifies the number of channels and type of ESFAS instrumentation required to be operable, response times and periodic surveillance tests to verify i

operability, and actions to be taken when the minimum operability require-l ments are not met. TS 3/4.7.'.5 defines operability requirements for i

MSIV's and actions to be taken when one or both MSIV's are inoperable.

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The operability requirements for the main steam isolation valves ensure that no more than one steam generator will blow down in the event of a main steam line rupture assuming a single failure. Ensuring that only one steam generator blows down prevents the containment design pressure from being exceeded and limi+s positive reactivity addition due to cool down of the reactor coolant system.

During normal plant operation, the MSIVs are maintained open by hydraulic pressure working.against compressed nitrogen gas. The energy stored in the compressed gas provides the motive force for valve l

closure. Technical Specification 3/4.3.2 currently requires an MSIV closure time of 6.0 seconds *, The pressure required to maintain the valve open and provide 6.0 second response time is high. Dynamic effects on components in the MSIV hydraulic circuits, due in part to the high pressures, have resulted in component failures and spurious MSIV closures during plant operation. A spurious MSIV closure during power operation will result in a reactor trip. Reducing the MSIV opereting pressure will result in increased component reliability but will also result in a slower MSIV response time.

The proposed change would increase MSIV closure time from 6.0 to 8.0 seconds. Specifically, the response time listed for the MSIV's in Table 1

3.3-5, "ESFAS Response Times", under Main Steam Isolation Signal (MSIS) is increased from 6.9 to 8.9 seccnds (0.9 seconds is allowed for instrumentation response time, the remainder for the valve). Also, the I

i response time for the MSIV's listed in TS 3/4.7.1.5 is increased from 6.0 to 8.0 seconds.

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• At the time PCN-207 was proposed, the required MSIV closure time was 5.0 seconds. Since then, Amendments 46 and 35 to NPF 10 and NPF-15, respec-i l

tively, dated May 16, 1986, changed the time to 6.0 seconds.

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3.0 EVALUATION l

A.

PCN-84 l

l The NRC staff has evaluated the proposed change and finds it acceptable on the basis that 10 CFR 50.49, which was issued after the SONGS 2 and 3 l

operating licenses were issued, defines the current NRC requirements for an acceptable environmental qualification program. Specifically, this rule identifies record keeping requirements, implementation schedule, and environmental qualification criteria. The rule suaersedes previous staff criteria for environmental qualification such as tiose contained in NUREG-0588. License Conditions 2.C(5)a and 2.C(5)b identify schedule, environmental qualification criteria and record keeping requirements.

Because these requirements have been sup(erseded by 10 CFR 50.49, the pro-conform to changes in the regulations a,nd is therefore acceptable.

In addition, the Standard Review Plan (SRP) Section 3.11. " Environmental Qualification of Mechanical and Electrical Equipment," describes the general requirements for the design and environmental qualification of all equipment. The specific acceptance criteria for assessing the acceptability of en environmental qualification program for safety related electrical equipment is provided by reference to NUREG-0588.

10 CFR 50.49 has superseded NUREG-0588 and thus specific SRP acceptance criteria are now found in 10 CFR 50.49. The proposed change deletes License Conditions 2.C(5)c and 2.C(5)d which define the requirements for an environmental qualification maintenance and surveillance program and requires the affirmation of implementation of the improved surveillance and maintenance procedures.

10 CFR 50.49 requires that safety related electrical equipment remain qualified for its qualified life and that replacement equipment also be qualified. These two aspects of 10 CFR 50.49, while not specifically identifying the means of accomplishing these requirements, will achieve the same goal as License Conditions 2.C(5)c and 2.C(5)d; thus, while deletion of these License Conditions removes the specificity of how equipment will be maintained in a 1

qualified condition for its life, the same ultimate requirement is mandated by 10 CFR 50.49. Since the requirements of 10 CFR 50.49 must be met and since they define an acceptable environmental qualification program, the environmental qualification program will continue to meet the acceptance criteria even with the deletion of License Conditions 2.C(5)c and 2.C(5)d. Based on the above, the NRC staff finds the proposed deletion of License Condition 2.C(5) to be acceptable.

B.

PCN-183 The NRC staff guidance for this proposed change are given in Item II.F.2, " Instrumentation for Detect' ion'of Inadequate Core Cooling,"

of Generic Letter 83-37, "NUREG-0737 Technical Specifications."

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The generic lettnr contains the following statements:

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Subcooling margin monitors, core exit thermocouple, and a reactor I

coolant inventory tracking system (e.g., the differential pressure measurement system designed by Westinghouse, the Heated Junction 1

Thermocouple system designed by Combustion Engineering, etc.) may be I

used to provide irdication of the approach to, existence of, and

. recovery from inadequate core cooling (ICC). This instrumentation should be operable during the Power Operation, Startup, and Hot Shutdown modes of operation for each reactor.

Subcooling margin monitors should have already been included in the l

present Technical Specifications. Technical Specifications for core exit thermocouple and the reactor coolant inventory tracking system l

should be included with other accident monitoring instrumentation in the Four core-exit thermocouple in each present Technical Specifications.,the reactor coolant tracking system core quadrant and two channels in 1

l are recuired to be operable when the reactor is operating in any of the

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l above mentioned modes. A minimum of two core-exit thermocouple in each 3

i quadrant and one channel in the reactor coolant tracking system should l

l be operable at all times when the reactor is operating in any of the above mentioned modes. Typical acceptable LC0 and surveillance requirements for accident monitoring instrumentation are provided in.

l The licensee responded to the generic letter by proposing Technical l

Specification 3/4.3.3.6, which added requirements for a HJTC system and a RVLMS to the technical specifications.

I The NRC staff has evaluated the proposed technical specification and finds that only the Actions part of the proposed technical specification

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l proposed technical specification permits continued operation (y the devirtes from the guidance in the generic letter. Specifically l

after 7 l

l days operation with one less than the required number of channels or after 68 hours7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br /> operation with one less than the minimum number of channels) provided a special report is submitted to the Commission pursuant to Technical Specification 6.9.2 within 30 days following the i

event and outlining the cause of the inoperability and plans and schedule l

for restoring the system to operable status. Generic Letter 83-37 recommends going into the hot shutdown mode if repairs cannot be made after the 7-day or 48-hour period. Although deviating from the generic letter guidance, SCE's Actions for the inoperability of the HJTC and clarification of operability (four or more sensors operable, one sensor in the upper head and three sensor in the lower head) are consistent with the NRC Staff's approval of the Combustion Engineering Owners Group proposed technical specificat, ions for the HJTC system.

For the subcooled margin meter, the proposed technical specifications comply with the guidance in the generic letter in all aspects (Limiting Conditions for Operations, Applicability, Actions and Surveillance Requirements).

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, The pronosed technical specifications for the core-exit thermocouple (TC) exceed the criteria in the guidance: for " required number of channels" (seven TCs in the core quadrant versus the guidance of four l

TCs in the core quadrant), and for " minimum channels operable" (four TCs in the core quadrant versus the guidance of two TCs in the core quadrant).

In all other aspects the proposed technical specifications for core-exit thermocouple ccmply with the generic letter's guidance.

As a result of the review cf the cited material, the licensee's response for technical specifications for Item II.F.2, inadequate core cooling (as instrumentation, is judged to meet the guidance in the generic letter approved by the NRC staff for the Combustion Engineering owners group generic technical specification), and is, therefore, acceptable.

i C.

PCN-207 l

The NRC staff has evaluated the two principal aspects of the

)roposed l

change. These are (1) the effect of the proposed change on tie plant safety analyses described in Section 15 of the FSAR, and (2) the effect of the proposed change on peak containment pressure.

l With regard to the effect of the proposed change on the plant safety analyses, the worst cases (steam line breaks inside and outside containment) were reanalyzed and found acceptable by the licensee as part of their Cycle 3 reload analysis. This analysis, which conservatively assumed a 10 second MSIV closure time (rather than the proposed 8 seconds) was reviewed and approved by the NRC staff as part of Amendment 47 to NPF-10 and Amendment 36 to NPF-15. These amendments were issued on May i

16, 1985, and included a safety evaluation describing the basis for the l

staff's approval.

With regard to the effect of the proposed change in peak containment pressure, the licensee submitted a reanalysis of containment pressure following a main steam line break (MSLB) based on an 8 second MSIV closure time. The staff compared this analysis with the analysis in the FSAR, which is based on 5 second MSIV closure time. The reanalysis shows that while the peak pressure and temperature within containment are reached earlier in the accident, the peak containment pressure is still bounded by the value of 55.7 psig given in the FSAR analysis.

This is due to the fact that the licensee has made several modifications in the modeling of the MSLB event and valve response. The changes in methodology mainly involved more realistic and detailed (but still l

conservative)modelingthandidtheFSARanalysis. The licensee states that " Changes were made to the original FSAR methodology to accommodate longer.MSIV response times and achi, eve. acceptable results in the safety Analysis."

1 In the amendment request and in the licensee's September 6, 1986 l

response to a staff request for additior.al information (RAI), the i

licensee has enumerated the modifications made to the analytical model:

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(1) For the revised analysis, the MSIV flow area is assumed not to change during the first second of closure duration. The flow area is then linearly decreased to zero over the remainder of the closure i

duration.

(2) The rapd d depressurization of the ruptured steam generator causes more than 50'., of the total feedwater flow to be diverted to this unit.

The origina'. analysis for the FSAR assumed that 100% of the total feed-1 water flow is diverted to this unit. However, subsequent analysis i

indicated that at full power only 65% of the total flow would be diverted i

to this unit. The revised analysis assumes that 87.5% of the total flow j

l is diverted to the ruptured unit.

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1 (3) The steam lines were modeled as two separate nodes associated with i

each steam generator instead of being combined with the steam generator I

node.

l (4) Three separate flow resistances (from each steam generator, to the l

cross tie, and for the cross tie itself) were used instead of a single, i

combined flow resistance.

j (5) The Darcy equation, used to calculate steam-line flow was modified to account for compressibility.

I (6) Choking at various points in the steam lines was considered when appropriate conditions existed.

i The staff has reviewed these modifications and found them to be both l

realistic and conservative.

In particular, the changes outlined in (1) and (2) above would have the indicated physical effect of lowering the mass / energy release to the containment. According to the submittal, the peak pressure reported to the FSAR is calculated using the code COPATTA, and is based on mass and energy releases determined by the cede SGNIII/CONTRANS. On this basis the NRC staff accepts the calculated results as realistic and conservative.

l An additional factor outlined in the responses to the RAI adds to the overall conservatism of the revised analytical model. The need to increase the response time became apparent after licensing San Onofre Units 2 and 3 when there appeared to be little margin between the five second technical specification response times and actual measured response times in the field.

Initi, ally., the licensee proposed an l

increase in MSIV response time from five to six seconds in PCN-96. This response time increase was approved by the NRC staff on May 16, 1986, in Amendments 47 and 36 to the San Onofre 2 and 3 licenses.

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This increase in response time was justified using the original FSAR 1

methodology with actual valve flow characteristics instead of those assumed in the original analysis. Because of difficulties with the l

original valves delineated in the responses to the RAI, valves manufactured by Paul Monroe Hydraulics were used to replace the original Marotta dump valves.

It was anticipated that the MSIV response time would increase with these modifications. The NSSS vendor, Combustion i

Eng.ineering, was requested to reanalyze the main steam line break events to support longer valve closure times. This action was carried out in parallel with the detailed design of the rrodifications. As it turned out, when modified, the MSIVs close in less than six seconds, the PCN-96 approved response time. This adds to the conservatism of the model j

assuming the eight second closure time.

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l Additional conservative assumptions made in both the criginal analysis and in the revised submittal remain, thus adding further to the overall conservatism of the analytical model.

• In summary, the licensee has submitted a proposed change to the technical specifications which increases the required closure time for the MSIVs. A modified form of the model used in the original FSAR analysis has been utilized to demonstrate that no NRC safety requirements are compromised.

The NRC staff supports both the realism and the conservatism of the revised model based on the codes used to calculate mass / energy release (SGNIII/CONTRANS) to containment and peak containment pressure (COPPATTA).

The staff finds that the information submitted fully supports the amendment request, and the request is, therefore, acceptable.

4.0 CONTACT WITH STATE OFFICIAL l

The NRC staff has advised the Chief of the Radiological Health Branch, State Department of Health Services, State of California, of the proposed deterrrinations of no significant hazards consideration.

No comments were received.

5.0 ENVIRONMENTAL CONSIDERATION

These amendments involve changes in the installation or use of facility components located within the restricted area. The staff has determined I

that the amendments involve no significant increase in the amounts, and l

no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumu-occupation radiation exposure. The Commission has previously issued proposed findings that the amendments involve no significant hazards consideration, and there have been no public comment on such findings.

Accordingly, the amendments meet th.e el.igibility criteria for categori-cal exclusion set forth in 10 CFR Sec. 51.22(c)(g).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need to be prepared in connection with the issuance of these amendments.

6.0 CONCLUSION

Based upon our evaluation of the proposed changes to the San Onofre Units 2 and 3 Technical Specifications, we have concluded that:

there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. We, therefore, conclude that the proposed changes are acceptable.

Principal Contributors:

F. Allenspach, R. Lagrange, R. Lipinski, H. Rood, and S. Sun Dated:

August 14, 1987 i

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