ML20127A277

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Delineates Instrumentation & Control Sys Branch Position on ESF Actuation Sys Subgroup Relay Surveillance Requirements.Applicant Should Test ESF Actuation Sys Subgroup Relays Every 6 Months
ML20127A277
Person / Time
Site: 05000000, San Onofre
Issue date: 08/30/1982
From: Speis T
Office of Nuclear Reactor Regulation
To: Novak T
Office of Nuclear Reactor Regulation
Shared Package
ML082410334 List:
References
FOIA-85-243 NUDOCS 8209230297
Download: ML20127A277 (4)
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AUG 3 0 Wet MEMORANDUM FOR:

Thomas M. Novak', Assistant Director for Licensing Division of Licensing FROM:

Themis p. Speis, Assistant Director for Reactor Safety Division of Systems Integration SU5 JECT:

SANONOFREkUCLEARGENERATINGSTdTION, UNITS 2&3-ESFAS SU3 GROUP RELAY SURVEILLANCE REQUIRO4ENTS

Reference:

(1) ~ R. Di' etch, " Docket No. 50-3'51, knendment Application No. 8, San Onofre Nuclear Generating Station, Unit 2", SCE, July 23, 1982.

(2)

K. Baskin, " Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generating Statien, Units 2 '

& 3", SCE, August 15, 1982.

plant Name:

San Onofre, Units 2 & 3 Docket Nos:

50-351/352 Licensing Status:

OL Responsible Branch:

LB #3 project Man.ager:

H. Road Review Branch:

ICSB Review Status:

Compl ete The purpose of this memorandum is to delineate the ICSB position with respect to Engineered Safety Feature Actuation System (ESFAS) subgroup relay surveillance requirements for San Onofre Nuclear Generating Station, Uriits 2 & 3 (SONGS 2 & 3).

It has been fundamental regulatory practice to require that plants'be designed such that protection systems can be tested while the plant is at power.

Industry,_ developed standards also recognize the capability for periodic testing with the plant at power to be a fundamental feature to be included in protection system designs. The design bases for plant design include the reconnendations of Regulatory Guide 1.22 and IEEE Std. 338 which state that protection systems, including the actuation devices, should be designed to be testable during plant operation as weil as during intervals when the plant is shut down.

It appears from references 1 and 2 and discussions between the applicant and staff in a meeting held July 2g,1952 tna the assignment of S335 ESFAS actuated equipment to the ESFAS subgroup relays has compromised to a large exten: -he ca; ability of tes-ing the complete actuation circuitry at power anc inus may not fully Nf p.

Contact:

R. Stevens s

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comply with the above regulatory guidance.

Of course, the regulatory guidance does allow exceptions in cases where testing actuated equipment at power could cause unsafe plant conditions / operation as long as sufficient justification (s) is provided.

Regulatory Guide 1.22 and IEEE 338 state that the protecti n system should be designed to permit periodic test-ing Q

through the final actuation device while the plant is operating.

In this particular case the ESFAS subgroup, relays are upstream of the final actuation devices and should therefore be capable of being tested while the plant is at power.

Via reference (1), the SONGS 2 & 3 applicant has. requested that the technical specifications be modified to extend the surveillance interva'l for the ESFAS subgroup relays from the present.6-ronth interval to an 1 B-month (refueling) interval.

Reference (2) provides further technical justification on this matter.

'The applicant has provided s'pecific relay design considerations in conjunction with reliability data and operating history from other Combustion Engineering (CE) NSSS plants to justify a proposed testing interval of 18 months.for the ESFAS subgroup relays. The ICSB has reviewed the applicant's technical justification (references 1 and 2), and has concluded that the infomation 3

is insufficient to justify the proposed 18-month test interval. At least l

several areas of concern remain which are:

A)

The applicability of the data collected on Arkansas Nuclear One, Unit 2 (ANO-2),

B)

The assumptions and modeling used for common-node fai ure, and C)

The failure to provide the additional information requested by the NRC staff at a meeting held on July 29, 1982.

The data collected for ANO-2 does not show the conditions under which the systems were required to operate (such as load groups, environments, length of time the subgroup relays were continuously energized between testing, etc.) and the degree to which these conditions are identical to those anticipated for SONGS 2 and 3 equipment. The applicant states that the data available to calcul.te the conman-rode failure (Beta-factor) vias insufficient._ Therefore, the applicant has assumed a cocren node failure contribution of 0.1.

The IliS is concerned that this contribution may have been underestimated and that plant specific operational data is required before this failure contribution can be adequately assessed.

On July 29,1g52, the NRC staff mat with the SONGS 2 & 3 applicant and

neir contractors to discuss the above subject.

Following a discussion cf the issue on the ESFAE subgroup relay surveillance recuirements, the l

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2 NRC staff requested information to ' support the proposed Te:hnical Specifications and to justify the adequacy of the protection system des.ign with respect to capability for testing at power. The applicant was requested to provide a list of ESFAS subgroup relays and the actuated equipment associated with each tha't:

1.

cannot be tested at power, 2.

can be tested at power but on1y by manually defeating (bypassing) and subsequently restoring the ESFAS tr'ain, or 3.

can be tested at power without defeating.the ESFAS train, but are operationally burdens:me to test.

This information has nct been supplied to the NRC staff to date a'nd is still required from the applicant before any further action can be taken by the NRC staff.

The applicant should aisc provide information to justify why the subgroup relays identified under item 1 above cannot be tested at power and to justify that the actuated equipment assignments j

to subgroup relays were made in a manner to minimize the number of components which cannot be tested with the plant at powe.

In the interim, it is the staff's position that the applicant test all the ESFAS subgroup relays which can be tested at power without adversely affecting the safety or operability of the plant every I

six months.

All ESFAS subgroup relays should be tested every 18-months and, if not tested within the previous six months, at any time the plant is b'rought to a condition where the subgroup relays can be tested without adversely affecting the safety or operability of the plant.

It should be reiterated that the SONGS 2 & 3 applicant must identify, with sufficient justification, ESFAS subgroup relays and associated actuated equipment that cannot be tested at power and justify that the actuated equipment assignments to subgroup relays were made in a manner to minimize the number of components which cannot be tested with the plant at power.

If there are any questions, please contact the ICSB.

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Themis P. Speis, Assistant Director for Reactor Safety Division of Systems Integration cc:

See attached list 9

T. Novak 4..

R. Mattson H. Rood F. Miraglia D. Brinkman D. Hoffman.

C. Rossi R. Stevens T. Dunni.ng F. Rosa D. Skovholt S. Hanauer e

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tS ORANDUM FOR:

Roger J; Mattson, Director Division of Systems Integration rD THRU:

T. Speis, Assistant Director d ' 'N,

  • for Reactor Safety, DSI p

FROM:

F. Rosa, Chief Ins.trumentation and Control Systems Branch, DSI'

SUBJECT:

COMBUSTION ENGINEERING STANDARD TECHNICAL SPECIFICATIONS (NUREG-0212) - PROPOSED REVISION 3 - RELAY TESTING This memorandum responds to your note (September 20, 1982) whereby you commented on the ICSB position pertaining to Engineered Safety Feature Actuation System (ESFAS) subgroup relay. surveillance requirements in proposed revision 3 to the CE Standard Technical Specifications (NUREG-0212).

You requested information pertaining to the effects on expected system reliability which might result from extending the surveillance of ESFAS subgroup relays from the proposed 5-month interval to an 18-month interval.

CE has referred to information submitted by Southern California Edison (SCE) on the SONGS 2&3 docket which includes a reliability study to determine the effects of the relay test interval. A memorandum (T. Speis to T. Novak) dated August 30, 1982 provided the ICSB evaluation of the SONGS 2&3 information to the Division of Licensing for subsequent transmittal to the applicant.

The first step in the study was to calculate relay unavailability. The relay unavailability was obtained using a failure rate calculated from

. ANO-2 operational data. The individual relay unavailability (RU) that results from surveillance intervals of 6 months and 18 months is:

RU (6-months) = 7.2 x 10-#

RU (18-months)= 2.2 x 10-3 The individual relay unavailability increases by a factor of 3 for the 18 month test interval.

The SONGS applicant then considered overall safety system unavailability.

A fault tree analysis was used to quantify the impact of auxiliary relay test intervals on auxiliary feedwater system (AFWS) unavailability. The analysis performed shows the worst case system unavailability to be:

AFMS Unavailability (5 months) = 1.9 x 10-5, AFWS Unavailability (18 months) = d.2 x 10-#

The relay failure rate used to :b sin the above AFWS unavailability data is based on the National Reliability Evaluation Program (GREP) Generic Ca:a Eise.

In this case, tne safety system unavailability increases by a factor of 2.

Using the ANO-2 operating history relay failure rate, analysis sho.is the increase in system unavailability to be less than a factor of 2.

Contact:

R. Stevens, X29456

Roger J. Mattson -

One could question whether the increase in safety system unavailability is significant since it falls within the unreliability range (10" to 10-5) which is defined in Standard Review Plan Section 10.4.9 and is the range used by the staff to determine the acceptability of AFWS reliability analyses.

However, ICSS feels,after reviewing the applicant's information and after discussions with the Reliability and Risk Assessment Branch, that there are ambiguities in the methodology used and a lack of consideration of uncertainties in the data used. Some of the areas of concern are:

1) The sole use of the AFWS in the fault tree analysis to quantify the impact of subgroup relay test intervals on safety system unavailability.

Should a fault tree analysis be performed on other safety sys.tems? Also., there appears to be inconsistency between the SONGS challenge rate (2.17/yr) and the ANO-2 challenge rate (appmximately 6/yr) which was calculated from AND-2 operational data.

2) The applicant's fault tree logic was not provided for staff review.

Before any final conclusions can be reached, a cetailed review of the fault tres will be required, especially the modeling aspects of comon cause failure.

3) The applicability of the data collected on Arkansas Nuclear One, Unit 2.

The data collected for ANO-2 does not show the conditions under which the systems were required to operate (such as load groups, environments, length of time the subgroup relays were continuously energized between testing, etc.) and the degree to which these conditions are identical to those anticipated for SONGS 2&3 equipment and the equipment for other CE plants.

4) The assumptions and modeling used for the coman mode failure contribution. The applicant states that the data available to calculate the common-mode failure contribution was insufficient.

Therefore, the applicant assumed a comon mode failure contribu-tion of 0.1.

ICSB is concerned that this contribution may have been underestimated and that plant specific operational data is required before this failure contribution can be adequately assessed.

A sensitivity analysis should also be used to evaluate the effect of varying the common mode failure rate using a reasonable upper bound.

Aside from the concerns related to the SONGS 2&3 analysis, it appears from the ICSB review that the assignmen; of the SONGS ESFAS actuated e:uipment to the ESFAS subgroup relays has cen:r:mised to a large extent the ci 5bility of testing the c:mplete ac us: ion circui:ry v.hile :he plant is 9

Roger J. Mattson at power and thus, clearly violates regulatory requirements. As stated in the ICSS proposed memorandum, IEEE Std. 338 requires that protection l

systems, including the actuation devices, be designed to be testable during plant operation as well as during intervals when the plant is

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in the shutdown mode. The applicant has been r'equested to provide

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additional information to identify those subgroup relays and associated actuated equipment that cannot be tested at power.

I would like to emphasize that the staff is willing to give exceptions (as allowed by regulatory guidance)to the proposed 6-month surveillance requirement in cases where testing actuated equipment at power could cause unsafe plant conditions / operation as long as sufficient justification'is provided.

Discussions are presently underway with the SONGS applicant

.so that this issue can be resolved.

It should be noted that the current Westinghouse Standard Technical Specifications require applicants to test slave relays (counterpart of J

CE ESFAS subgroup relays) quarterly.

Recent discussions have revealed that a significant amount of work has been expended by various applicants (Seabrook, St. !.ucie 2, SNUpPS, Waterford) in designing plants so that the complete actuation circuitry can be tested to the maximum extent possible while the plant is at power.

Furthermore, recent review of Licensing Event Reports has shown that there have been a considerable

. amount of safety system relay failures in operating plants.

The attach-ment to this memorandum gives several examples of typical relay failures.

No attempt has been made at this time to quantify this information. How-ever, the ICSB believes, based on a qualitative judgment, that operating plant experience indicates that the six-month test interval should not be relaxed.

t Based on the above discussion, ICSB recommends that the proposed test inter-val for the CE ESFAS subgroup relays remain at six months. We are particu-larly concerned that acceptance of an 18-nionth test interval in the Standard Technical Specifications would encourage designs which could not be fully tested with the plant at power.

This would effectively eliminate the possibility of increasing the test frequency if the relay reliability dur-ing actual plant operation were found to be lower than assumed by the SONGS applicant and CE.

In any event, ICSB intends to carefully consider all technical information provided by CE or applicants in determining test requirements which insure high protection system reliability without being excessively burdensome to the operating plant staffs.

In particular, we are continuing to discuss this issue with the SONGS 2 and 3 applicant.

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Roger J. Mattson 4-I hope this discussion has answered your concerns.

If you have any further questions, please contact ICSB.

s y vW yYYA Faust Rosa, Chief Instrumentation and Control Systems Branch Division of Systems Integration

Attachment:

As stated cc: 'CRossi TDunning RStevens DHoffman DBrinkman O

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I Ct)t ti'O f !!!!! 1 CODE HLI.AYX-HELAYS I

l S Y S II:ll C tDil E E-EMER G 6 tit 4 tiR AT OR SYS T, cot 41RULS F AI LUHE 3 Ht_ AC 1014. PW H 3 DEA VLIH VALLiiY I ( Plf R ) ;DRE AKER IGliNER ATOR.

KliY r:Ulti)S D i tiSEL 3 EMEltG t34 CY PO Wi_H. tiLECTit I C I tiST. SY STEtt UPEH At3 I LI T Y ;

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C A ll!Gilit Y I10000;100000 I 87 I I I Oil U 802 i

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A b b i lt A L I t AIE UF EVI.tl l - 090 57 tl.

POW Lit Li' V t:L - 100%.

CAUSE - DIItTY lit t. AY C ON I AC IS.

WillLC Pf iti:UleM if 8G A MOtilllLY S U R Vil l L t_ A N C E

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ALL Ul f ellH AC ELEC I H ICAL PtlWtill SUUltCCS WEftli V Lis t F I ED UI 'E ll Allt.C.

Illd APl*Altt t41 CAUSC WAS DIH1Y Ct)t4 T ACT S ON IIS U18EftA11t4 b 14i L A Y.

i llt:. HELAY tf A b E X Eft C l iAi D AFID Ol* Lit AT 10t,4 ALLY l

Clt.CKLu SAII LF AC10R ALY.

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LURE.

EVENT DATE: 012582-RZPCRT DATE: 051082.

Nggg Gg SYSTEM: RESIDUAL EEAT RENCY sys g CONT CCXPCNENT: RELAYg ^

CAU3E: RELAY CCIL CV:REEATED.

(N3:C 173651) WE:L: PERFCRx:NG 3: 4.5.3.1.D. RH2 PUEPS

'B' Alr3 I U" EYPAS: VALV: 1-FCV-74-30 FA"-ED.C CLCSE As DC3ZGNED. LCCP :

WAS DEaLARgD NCP:RASLE AND UNIT SHUTDOWN WA3 BEGUN (*I.CH SPE**

3*

ENERG ED CCIL FOR RELAY 1cA-E1023 FA".=D.

.EE XCwE. 366A77239 CC L FOR THE GE12HFA 115,vw... 60-EERT RC,.AY WAz g=p A :D.

A49F CCIL cv:REF.ATED,

.w.K ARXA..'RE IN tr:RG:IED PCE TICK, TgUs yngygNTING VALVE CLOSURE.

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: C:::77 1
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  • ?. 01-:*5 F.:V 1 U : ATE CN CONTA:NKENT IECLATICK VALVE FAILURES.

EVENT DATE: 1C2581 R: PORT DATE: 051982 N355: WE TYPE: PWR CY TEM: CUTN13:T ISCLATICW SYS & CONT CCKPCNENT: RELAYS CAUSE: STICKING RELAY IN CLOSING CIRCUIT.-

(US:C 173584) WE:LZ CCET"C"NG COLD EEUTDOW:t VALVE TESTING, EALL VALVE CV-534 DID NCT CLCE: AUTCKAT:CALLY CN INITIATICM OF A C"NTAINKENT ISCLATICN EIGNAL (C:5).

ALSC, CN NOVEX2ER 1, 1981, LURING A CII TEST, WITH THE UNIT IN ROT SHUT CWN, CONTAINE:NT ISCLATICN VALVE CV-107 DID ECT CLCSE AUTCKATICALLY CN RECEIPT OF A C:5. CV-534 Ah"! CV-107 ARE AIR OPERATED VALVE 5 INSTALLED CN TEE PRE 55URIEER RELIEF TANK MAKZUP WATER LINE AND THE RCS DRAIN TANK VENT, RESPECTIVELY.

INVESTICATION INTO THE OPERATING CIRCUITRY OF EACH VALVE REVEALED A STICKING R: LAY IN THE CLOSING CIRCUIT OF EACN VALVE.

EACH CF THE STICKING RELAYS WAS REPLACED Al*D THE VALVES RETURNED TO SERVICE. AN INV:STIGATICN HA5 CETERM:NED THE IXXZDIATE CATSE CF FAILURE TC EE XIOPOSITICNED CONTACTS CN THE RELAYS BRCUGHT

/.ECUT BY AN ARMING CONTA*T W*. RED IN SERIES WITH THE R::35 WINOINGS OF TEE RELAYS.

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1983.

DOCKET

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C A II~.G OR Y a /0000 312 000 0 L O I I Itate 012 J (t 1 0 8 Ct fla t; A l *Y Lt it is t I lt Y A

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POWElf I.EVEL - 100%.

CAUSE - SilOftTED Hi? LAY C u l t..

A It tiLA Y IN till: *>T Alt i C l llCli t l it Y FOlt tile TUlt01t4E' (1841 Viita EMe.itGINCY FE EDW A T Lit AUX 11.1 Ally Ull PUMP SilOR TED Will Cil I lli f *l'E D IllE D. t..

IJill. A Kell SUI'8t Y t t4(s illI S C I l4CU I Tit Y.

TifE"C AUSE dAS Tile AH-li,14ELAY CHIL SilOI(T i t4G.

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Ill E h elO lt iti D 14E L A Y WAS ItEPLACED Willt Aid IMPROVED l

t.U I LE R-i lA MME 84 14IL AY.

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ts A I I; I 180 0 S Y I'l.

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f-AILthtE WAS UUSERVED UY LUSS OF t

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'A' LPCI SYSTEM flAD 08.11t4 DE CL AHtiis ll40PEftAULL A 1 B ile; i1ME (LliH J-uo-oJ/JL)-

E339 esH C'A K L H WAS I M M I D E A lt. L Y I4L CL isS t. D FHUN Ti tti CUNTHOL HUOM.

'A' 8

CUtit SPHAY S Y:, IE M VALVES Wl;HL w i lllOU T POWER FOR LESS TilAN 2 i

MINU1LS.

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l ilti latiLAY WAS REPLACliD AND r

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1; t11; HL,C hlY UUSSES 14EMA lf 4E D DiiLNCI(t,I 11:0 UIil ll NuitMAL POWligt W AS MN4UALLY ItL S 1 0111 0. 'CAUSL WAS IHACLU 10 Ut4DERVOL I AGE filiL AYS

( Mtst4 8 TUR ING lilli h f A tiott Y a d:::S i t h I' h t D E H ) 3: A ILU1411 I ts Oli AC I IV A T E 488" tite LilhS (Jt P U4 ti f t.

Iti t. AY b Willf C C a stiCKCD AND fili T Ultt4ED 10 til titAtsL L S I A 105.

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l itC ' UNDliltVOL I AGC fit: L A Y IN I IS PHEht:14 I M88 L I CA I IUtd ab UI4t)l;It I t4VE SI I G A l I ON.

I

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UNITE 3 STATES h

NUCLEAR REGULATORY COMMISSION r

2

.s REZION lli

'5 1

7senoossvatinoao

,/

otsN ELtvu. stumois som O'ctober 15, 1982 7

MEMORASDUM FOR:

R. D. Walker, Chief, Reactor Projects Section 2C FROM:

N. J. Chrissotimos, Senior Resident Inspector, Quad-Cities Nuclear Power Station

SUBJECT:

TECHNICAL SPECIFICATION INTERPRETATION Recently I have been involved with a response to an Iowa Electric Light and Power Company's denial of a noncompliance involving operability of an emer-gency system.

In responding to the denial, it appears that Technical Specification require-ments allow unit operation.to continue for seven days with two emergency systems concurrently inoperable.

Specifically, (a) when a diesel generator is inoperable, continued reactor operation is permissible for seven days provided that all of the low pressure core and containment cooling subsystems and the remaining diesel generator are operable.

If this requirement cannot be met, an orderly shutdown shall

. be initiated and the reactor be placed in gold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

N (Duane Arnold Technical Specification 3.5.T.1)

(b) When the high pressure coolant injection (HPCI) system is inoperable, reactor operation is permissible for seven days provided that all active components of the ADS subsystem - the RCIC system, the LPCI subsystem and both core spray subsystems - are operable.

If this requirement is not met, the same 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shutdown requirement is applied. (Technical Specification 3 75. D. 2)-

The situation was that the diesel generator was _ unknowingly inoperable for

17 days and within this time frame, HPCI was al~so inoperable for approxi-mately 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br />. The licensee was cited for violating the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO because it was felt that the equipment powered by the inoperable diesel generator (core spray, LPCI subsystem) was also considered to be inoperable and thus the HPCI LCO was violated.

The licensee believed that the inoperability of diesel generator 1G-21 did not render the B core spray subsystem inoperable for purposes of the seven day LCO in effect based on the following:

UnderAm'erkdent77,thedefinitionofOPERABLEisclarifiedtoread:

~

A system, aubsystem, train, component or device shall be OPERABLE or have c_

  • - e.

.) j) f OCT 19 c

.,r

r R. D. Walker 2

10/15/82 T'

OPERABILLIY when it is capable of performing its specified function (s).

[

Implici~t*in this definition shall be the assumption that all necessary attendant instrumentation, controls, nor=al and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

Amendment 77 also clarified the definition of Limiting Condition for Oper-ation as follows: When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:

(1) its corre-sponding normal or emergency power source is OPERABLE: and (2) all of its redundant system (s), subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this specification.

(emphasis added)

Thus, under the foregoing interpretation of. Technical Specification 3.5.D.2 on March 5-6, 1982, the B core spray subsystem was OPERABLE for the purpose of satisfying the then applicable seven day Limiting Condition for Operation because its normal power source was operable and its redundant subsystem (Core Spray Subsystem A) was OPERABLE.

Since the B core spray subsystem was not inoperable for the purposes of Technical Specification 3.5.D.2 this Technical Specification was not violated.

By interpreting the specifications in this manner, we would be allowing a licensee to operate for seven days with both a HPCI system and diesel generator inoperable.

It should be realized that in this situation, under an accident condition with loss of offsite power, there would only be the

, minimal ECCS systems available to cope with the accident.

(One core spray,

pump and two LPCI pumps would not have power.)

When considering both the HPCI and diesel generator LCO's together, it is difficult for me to interpret that the core spray and LPCI pu=ps associated with the inoperable diesel can be considered operable to satisfy the HPCI LCO.

_- f-

~

c s 7 r

O

F R. D. L'alker

,, 3 10/15/82 I am requesting a position from the Office of Nuclear Reactor Regulation on the:piplicability of the definition of operable with respect to this matter.'

  • Although this is a specific problem, it may also apply to other BWR's which do not have standard Technical Specifications and thus should be looked

~

at generically.

N. J. Chrissotimos Senior Resident Inspector Quad-Cities Nuclear Power Station a

UE e*

4 9

p. ~.

. e 7

. -.

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