ML16152A918

Incomplete Memo Responding to 880408 Request for Review of BNL Draft Rept, Value Impact Analysis for Extension of NRC Bulletin 85-003 to Cover All Safety-Related Motor-Operated Valves

ML16152A918
Person / Time
Site:	Oconee
Issue date:	06/02/1988
From:	Houston R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Bosnak R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML16152A917	List: ML16152A916 ML16152A918
References
RTR-NUREG-CR-5140 IEB-85-003, IEB-85-3, NUDOCS 9002260259
Download: ML16152A918 (11)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION VASHINCTON. 0. C. 20555 JUN 0 2 as MEMORANDUM FOR:.

Robert Bosnak, Deputy Director Division of Engineering Office of Nuclear Regulatory Research FROM:

R. Wayne Houston, Acting Director Division of Reactor Accident Analysis Office of Nuclear Regulatory Research

SUBJECT:

REVIEW OF BNL DRAFT NUREG/CR-5140

Reference:

Memorandum from R. Bosnak to R.W. Houston, dated April 8 1988, "Request for Support to Review BNL Draft Nureg."

We are responding to your request in the referenced memorandum to review a BNL draft report entitled "Value-Impact Analysis for Extension of NRC Bulletin 85-03 to Cover All Safety-Related MOVs."

We initially read the January 1988 first craft, and then the April 1988 verson. Our review concentrated mainly on their use of the SARA code and the appropriate data volumes from NUREG/CR-4550, especially Surry (Vol 3) and Grand Gulf (6).

We performed some calculations using SARA in an attempt to verify some of the values given in Tables 5-2 and 5-3. But some of the MOVs considered in the PRA are not explicitly identified as such, and are classified as (part of) pipe segments. It would require considerable effort to locate these (hidden) MOVs and such time was not available; but from what we did simulate, the results were close to those in the tables. In lieu of undertaking such a task, we had some extensive phone conversdtions with Mr. J. C. Higgins of BNL, inquiring as to their scope and procedure for dnalyzing the problem. It was evident from these conversations that the BNL team has done a very thorough analysis and has correctly used the PRA data and SARA code to perform their calculations. We Glso asked the SARA contractor, EGLG/INEL, to contact Mr. Higgins regarding details of this application of SARA, and they reported that the BNL team had formulated a solid approach and seemed to have carried it out properly.

A basic dssumption in the BNL analysis is that MOVs would only see a high pressure differential in high pressure sequences (that is, transients and small LOCAs).

Moreover, the assumption was made that all the MOVs in all the systems would sce a high pressure differential in the high pressure sequenes. The first assumption may be nonconzervative, since it is possible that in some low pressure sequences, vdlves would see a high differential pressure, where high differential pressure is defined as a pressure differential comparable to the design pressure differential for the valve. The second assumption is conservative; valves in certain systems, such as the service water system or the component cooling water system, may not see a high differential pressure, even in high pressure sequences. In addition, the BNL study did not give credit for manual local recovery of the failed MOVs. The amount of 9 22 00 PDrFJ A

16 51 0 - 0 k%

G 16510OL A F E T Y I S~1 S U E S I

NR T: I

/S a

~C C II SAFETY IS UA E E

E N T S Y STE AGE

//87El'

,o15 87 SAFETY ISSUE LEVEL INFORMATION JE NUMBERs II.E.6.1 TITLEs IN-SITU TESTING OF VALVES FACT, 0. ROTHBERG TYPEs GSI IDENTIFYING ORGANIZATION, NRR

STATUS, SPONSORING OFFICE: RES ORITYe M TYPE OF REACTORS AFFECTED, LNR

OTHER, DENT ISSUES: GIt05 INTERFACING SYSTEM LOCA C4110oti Pe b

fauUU OF-V1116.1 vweb1-th akerota DESCRIPTION SOLUTION (NEAR AND LONG TERM)

E OBJECTIVE OF THIS TASK, AS STATED IN NUREG 0660 (THI A.

REVIEW THE RESULTS OF THE INVESTIGATIONS OF HE E CHECK TION PLAN),

WAS TO "EVALUATE WHETHER CURRENT REQUIREMENTS VAL VE-TASK-FORCEAND-PNOPOS ADDtT4,O] AL.CHECK VALVE IN R VALVE TESTING PROVIDE ADEQUATE ASSURANCE OF PERFORMANCE SIfEXAMINATIOH AND/AR CONDDION I BE DER DESIGN CONDITIONS."

SEDED BY INDUSTRY RESPONSE AND INT ATIV PER RRENT IN SERVICE TESTING OF SAFETY RELATED VALVES CON-REVIEW THE RESULTS OF THE RESPONSES TO IE BULLETIN 85-03 S.

ESSENTIALLY, OF EXERCISING THE VALVES PERIODICALLY.

(MOTOR OPERATED VALVE COMMON MODE FAILURES DURING PLANT

ADDITION, THOSE SAFETY RELATED VALVES NHOSE LEAKAGE IS TRANSIENTS DUE TO IMPROPER SWITCH SEINGS) AND PROPOSE NSIDERED IMPORTANT (CONTAINMENT ISOLATION VALVES, PRESSURE ADOIONAL OR HER VALVE MOTOR OPERATOR EXAMINATIONS AND/

OLATIOl VALVES. ARE PERIODICALLY LEAK TESTED AND VALVE RJESTS IN ORDER TO ASSURE NOV OPERABILITY UNDER DESIGN HOTE POSITION INDICATORS ARE OBSERVED PERIODICALLY 10 9AIS CoNDITIq11S.

4 VERIFY THAT THEY ARE OPERATIONAL.

NO TEMPERATURE RE-C.

REVIrWHTTERESULT CbNTRAC-1 bNL TUD OF THER AL RICTIONS OR REQUIREMENTS ARE INVOKED.

LEAK TESTING IS OVERLOAD PROTECTION DEVICES AND THEIR US T ORDfXLT UJALY PERFORMED AT RELATIVELY LOW PRESSURE AND IN SOME DETERMINE WHAT MODIFICATIONS TO CURRENT USE OF THESE IES CALCULATIONS ARE PERFORMED TO EXTRAPOLATE LEAKAGE TO DEVICES ON MOVS MAY BE REQUIRED.

ACIOR OPERATING PRESSURE. UNTIL THE RECENT ISSUE OF IE D. REVIEW RESULTS OF ETEC TESTS ON VALVE LEAKAGE IN ORDER LLETIN 85-03 THERE WAS NO IN SERVICE TESTING REQUIREMENT TO IDENTIFY POSSIBLE CHANGES TO IN-SITU VALVE LEAKAGE STROKE TEST ANY MOTOR OPERATED VALVES Wili PRESSURE IN TEST ACCEPTANCE CRITERIA.

E LINES.

AT THIS TIME THERE IS NO IN SERVICE REQUIREMENT E. PROVIDE RECOMMENDATIONS FOR MODIFICATIONS TO IOCFR 50.55 STROKE TEST ANY MOTOR OPERATED VALVE WITH FLOW IN THE A. REGULATORY GUIDES, CODES AND STANDARDS AND TECHNICAL HES.IETSIGSFTREAEVAVSIGOENDB SPECIFICATIONS. AS APPROPRIATE.

ALTERNATIVELY PROVIDE I SERVICE TESTINIG SAFETY RELATED VALVES IS GOVERNED BY RCMEDDGNRCIVESADO RES UPR L

'CFR55A(G HICH INVOKED SECTION XI OF THE ASHE CODE. RE-RCMEDDGNRCtlESADO RES UPR L

CFSAG HC NOKDSCINX FTH SECDR-CHANSGES WITH THE REQUIRED REGULATORY AND COST/BENEFIT Fc, i.jk, RRED TO HEREAFTER AS "THE CODE",

FOR TESTING OF ALL VALVES ANALYSES.

THOSE FUNCTION IS REQUIRED FOR SAFETY."

CURRENTLY, THE F. IDENTIFY THOSE AREAS OF VALVE PERFORMANCES WHICH ARE NOT AFF IS INTERPRETING THAT THIS RESTRICTS CONSIDERATION TO APPROPRIATE TO VERIFY BY IN-SITU TESTING SUCH AS APPL-r TO

~

VLoTflot 10SE VALVES WHOSE FUNTIONI IS NECESSARY TO ASSURE:

CATION OF THE VALVE OR VALVE COMPONENT MATERIAL PROBLEMS THE INTEGRITY OF THE REACTOR COOLANT PRESSURE

BOUNDARY, AND ISSUES RELATING TO ASSURING THAT PROPER PROCEDURES THE CAPABILITY TO SHUTDOWN THE REACTOR AND MAINTAIN IT ARE. IN FACT, FOLLOWED IN TESTING. MAINTENANCE AND OPER-

)

AN:

IN A SAFE SHUTDOWN CONDITION, AND ATION.

THESE AREAS MIGHT BE ADDRESSED AS SEPARATE TIHE CAPABILITY TO PREVENT OR MITIGATE THE CONSEQUENCES POTENTIAL GENERIC ISSUES.

ACCIDEIIS WHICH COULD RESULT IN POTENTIAL OFF SITE EX-

6.

BASED ON iHE RESULTS OF GENERIC ISSUE 105 PROPOSE LEAK SURED COMPARABLE TO THE GUIDELINE EXPOSURED OF IOCFRA.

RATE OR THER TESTING AT APPROPRIATE INTERVALS FOR PRE SSURE ISOLATION VALVES.

SD YNU Y SNEA IN AT

.)

216310-01 SAFETY ISSUES

~MN AGEMENT sysrE.

ATE' 10/13/87 5 AF ET Y I SS f

M

) A0 EM N

S S

fPAOtC 2

SAFETY ISSUE LEVEL INFORMATION SUE NUMBER, II.E.6.3 TITLE, IN-SITU TESTING OF VALVES NET CHANCE IN DOLLAR COST 511 H

NRC DEVELOPMENT..

)I NRC (IMPLEMENTATION/IMPOSITION).

NRC (ASSURE CONTINUED COMPLIANCE)..*

L PUBLIC/INDUSTRY/OTHER (IMPLEMENTATION)

O PUBLIC/INDUSTRY/OTHER (CONTINUED COMPLIANC) 0

- 4.55 a

.1Ao 5 Io '

IIET C"ANGES IN BENEFITS4

~

PUBLIC EXPOSURE.

. k.o4 10 1

to OCCUPATIONAL EXPOSURE.

Of AL3to 74b Ic10 CORE MELT FREQUENCY..

F'! ~

~

~

~

16K I~

.c3 b 4 (o~

2.

4.

4 o

o-4 HoweeesReusee To Ro (T2o c

bwe

.14Z 3.31 us I.h4AZ

216310-O S A F ET I S S U E S M NN A G E M E N T S Y S T E AGE AE 30/13/87 SAFETY ISSUE LEVEL INFORMATION

SUE NUMBER' II.E.6.1 TITLE' IN-SITU TESTING OF VALVES ISSUE APPROVAL AND PLANNING ISSUE APPROVAL DATE' 11/83C TECHNICAL RESOLUTION

LEAD OFFICE' RES SUPPORTING OFFICE(S)' RES AEOD INIIATION DATE' 11/83 INTER OFFICE REVIEH/COORDINATION COMPLETION DATE$ 09/87 PROPOSED SOLUTIONS/REQUIREMENTS APPROVAL bY OFFICE DIRECTOR DATE, 08/87 REQUIREMENTS REVIEW AND APPROVAL:

INITIAL CRGR REVIEW DATE' 4 I-a?-

RESOLVED WITH REQUIREMENTS*

RESOLUTION DATE' 04/89 R U 1.

E MI A K I N 6 OTHER (5PLCIFY)

FORM.....

ANPR PROPOSED FINAL GEN LETTER OFFICE RESPONSIBLE..

N NRR EDO APPROVAL TO PROCEED (YES/HO)

N N

YES CRGR REVIEW DATE......

x N 1.,8/7 ACRS REVIEW DATE......

N N

N

.12s87 EDO REVIEW DATE.

0 0

0 08/88 APPROVAL (YES/NOl......N N

N YES COMMISSION REVIEW DATE.

E E

E mmNnNN APPROVAL (YES/n0).

.N PUBLIC COMMENT DATE.

N MMMMMM VINAL APPROVAL AND ISSUANCE DATE 4

N gle89 REQUIREMENTS MPSITION NEDCD FORL VEIFCAIO

^. -

Pro5111 t -

LICENSaE PLANT.

MULIIPLANT ACTION CODE' OTHER' APPROVAL OF LICENSEE PROPOSAL NEEDED (YES/NO).

Jp

1216310-0 R.

ATEs t0'i3,87 SAFETY IS S U ES M SNAGEMENT 5 Y S TEM PAGE 4

SAFETY ISSUE LEVEL INFORMATION SSUE NUMBEs II.E.6.1 TITLE, IN-SITU TESTINO OF VALVES B. IMPOSITION -

PLANTS NOT LICENSED:

SRP REVIEW PROCESS' OTHER (SPECIFY).

6LeI-bG(2k6 L Tiy C. IPOSITION -

ALL PLANTS:

OFFICE RESPONSIBLEs zeZ D. fftE FOR VERIFICATION NEEDED OFFICE LYES/NO)

RESPONILE VERIFICATION.

MPA VERIFICATION PRIORITY' REQUIREMENTS IMPtEMENTATION BY LICENSEF AND VRIFICATION BY NRC COMPLETION DATE OF 51ATUS (PANTS)

LAST PLANT IMPOSITION......

0 OF 0

IMPLEMENTATION..

0 OF 0

emann VERIFICATION.......

0 OF 0

nnum REQUIREMENTS IMPEMENTATION BY STAFF:

A. 11Aff REQREMENTS --

N 0 N E COMPLETION DATE: 05/91 OFFICE(S) RESPONSIBLE B. ROUTIE-INSPECTION PROGRAM MODIFICATIONSt MODIFICATIONS NEEDED (YES/NO)'

MODIFICATION COMPLETION DATE' $nua M 0 DIFICATI 0 N TEX T

WNNO MODIFICATION TEXT FOR ISSUE uw

-I 2 is310O-0k S A FE I y I S S U E S N A G E M E N I SY S T E R,

ATEs10/IS/87 PAGE 5

PLANT LEVEL INFORMATION

)

ISSUE NUMBER, II.E.6.1 TITLE: IN-SITU TESTING OF VALVES IMPOS DOCKET REACTOR OL 01 MANAGE IMPOS IMPOS IMPI Em VERIFY PLANT DOCKET S PRIORITY TYPE 155U~

EXPIR METHOD

-ANtI COMPLET COMPLETF MPE KKK N 0 P L A N T D A T.A F 0 R IS S U us sc/ scac.i 4

ENCLOSURE 9 0MB No.: 3150-0011 NRCB P5-03, Supplement 1 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR PEACTOR REGULATION WASHINGTON, D.C. 20555 April 27, 1988 NRC B uLETIN NO. 85-03, SUPPLEMENT 1:

MOTOR-OPERATED VALVE COMMON MODE FAILURES DURING PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS Addressees:

All holders of operating licenses or construction permits for bciling water reactors (BWRs).

Puroose:

The purpose of this supplement to NRC Bulletin 85-03 (8 85-03), "Motor-Operated Valve Common Mode-Failures During Plant Transients Due to Improper Switch Settings," is to clarify (1) which valves are to be included and (2) the meaning of the phrase "...

inadvertent equipment operations (such as in advertent valve closures or openings).

as used in the bulletin.

Backaround:

B 85-03, which was issued on November 15, 1985, was prompted by the June 9, 1985 event at the Davis-Besse plant in which the inability to reopen two valves that had inadvertently been closed led to a loss of both trains of the auxiliary feed water system.

Discussion:

Review of the responses to 8 85-03 from BWR facilities, including those from the owners group, has indicated that there is a misunderstanding in regard to the related issues of (1) which valves are to be included and (2) the meaning of the phrase "... inadvertent equipment operations (such as inadvertent valve closures or openings).

." as used in the bulletin.

The first misunderstanding pertains to which valves are addressed by the bulle tin. As written, the action portion of B 85-03 applies to motor-operated valves in selected systems that "... are required to be tested for operational readi ness in accordance with 10 CFR 50.55a(g)..." At the time the bulletin was issued, the staff believed that the inservice testinq programs required by 10 CFR 50.55a(g) were applicable to most, if not all, of the safety-related valves in the selected systems. However, recent conversations with the owners group and several licersees have indicated that a number of valves in these sysrtems are normally kept in their safety positions and are not covered by the

NPCB 85-03, Supplement I April 27, 1988 Page 2 of 4 inservice testing program. However, if the proper precautions are not taken, these valves which are normally properly positioned could be mispositioned, either before or during the initial phases of an event. This would render the safety system inoperable unless the valves could be repositioned to the proper position. Therefore, the heading of the action section of R 85-03 has been revised to include all safety-related valves in the selected systems.

The meaning of the phrase

. inadvertent equipment operation (such as in advertent valve closures or openings)..." used in action item a of the bulletin can also be misunderstood. This phrase stems from the desire to address the salient feature of the Davis-Bess.e event -- namely, the inability to reposition either of two redundant valves that had been mispositioned earlier in the event. Although it was not the intent of the bulletin to expand the design-basis events for plants, it was intended to ensure the high reliability of individual safety systems. To this end, and given the chain of events associated with the Davis-Resse event, the staff felt that the only way to ensure this high reliability was to verify the ability of all valves to recover from mispositioning. Therefore, action item a of P 85-03 has been revised to clearly indicate that each motor-operated valve must be able to recover from an inadvertent mispositioning.

This revision to B 85-03 may expand the number of valves addressed by some licensees. In addition, some of these licensees may have already completed their scheduled activities to comply with the bulletin. Therefore, the time limits for completing all the activities (action item e) have been modified to allow additional time for those licensees who have already completed their planned activities.

Actions for All BWR Holders of Operatinq Licenses or Construction Permits:

For safety-related motor-operated valves in the high pressure coolant in, 4ection/

core spray and reactor core isolation cooling systems not included in the actions planned or completed in response to the original bulletin, develop and implement a program to ensure that valve operator switches are selected, set, and main tained properly. This should include the following:

a. Review and document the design basis for the operation of each valve.

This documentation should include the maximum differential pressure expected during both opening and closing of the valve for both normal and abnormal events to the extent that the events are included in the existing, approved design basis (i.e., the design basis documented in pertinent licensee submittals such as FSAR analyses and fully approved operating and emergency procedures, etc.). In addition, when determining the maximum dilferential pressure for valves that can be inadvertently mispositioned, the fact that the valve must be able to recover from such mispositioning should be included.

'Any motor-operated valve that is not blocked from inadvertent operation from either the control room, the motor control center, or the valve itself should be-considered capable of being mispositioned.

NRCB 85-03, Supplement 1 April 27, 1988 Page 3 of 4

b. Perform action item b of the original bulletin for any additional valves identified above.

The intent is to provide assurance that a program exists for selecting and setting valve operator switches to ensure a high reliability of safety system valves. If changing the switch settings is not sufficient to ensure the capability for repositioning a particular mispositioned valve, a justification for continued operation should be provided in the bulletin response if the licensee does not elect to implement additional actions, such as administrative or procedural controls or equipment modifications, to minimize the likelihood of valve malfunction.

c. Perform action item c of the original bulletin for any additional valves identified above.

d. Perform action item d of the original bulletin for any additional valves identified above.

e. Within 30 days of receipt of this supplement, submit a written report to

'o NRC that,,for any additional valves: (1) provides the revised results of item a, above and (2) contains a schedule for completion of items b through d, above.

1. No changes from the schedule for complying with the original bulletin are anticipated for plants with an OL that, as of the date of this supplement, had not yet begun the refueling outage during which the activities in the original bulletin were scheduled to be accomplished.

2. Plants with an OL that, as of the date of this supplement, have com pleted their planned activities in response to the original bulletin have until the completion of their next refueling outage to complete any additional activities resulting from this supplement. The final report covering the activities already completed in response to the original bulletin shall be submitted in accordance with the original schedule.

3. No changes from the schedule for complying with the orioinal bulletin are anticipated for plants with a CP.

f. Revise the report requested by the original bulletin to include any addi tional valves. This revised report shall be submitted to the NRC within 60 days of completion of the program for the additional valves.

Additional Related Generic Communications:

8 85-03 identified a number of related generic communications.

Since its issuance oin November 15, 1985, the following additional related information not-ices have been issued:

a. Information Notice No. 86-29, "Effects of Changing Valve Motor-Operator Switch Settings," was issued on April 25, 1986.

NRCR 85-03, Supplement I April 27, 1988 Page 4 of 4

b. Information Notice No. 86-93, "IEB 85-03 Evaluation of Motor-Operators Iden tifies Improper Torque Switch Settings," was issued on November 3, 1986.

c. Information Notice No. 87-01, "RHR Valve Misalignment Causes Degradation of ECCS in PWRS," was issued on January 6, 1987.

The written reports requested above shall be addressed to the U. S. Nuclear Regu'atory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, under oath or affirmation under the provisions of Section 182a, Atomic Energy Act of 1954, as amended. In addition, a copy shall be submitted to the appro priate Regional Administrator.

This request for information was approved by the Office of Management and Budget under clearance number 3150-0011. Comments on burden and duplication should be directed to the Office of Management and Budget, Reports Management, Room 3?08, New Executive Office Building, Washington, D.C. 20503.

Although no specfic request or requirement is intended, the time required to complete each action item above would be helpful to the NRC in evaluatina the cost of this bulletin.

If you have any questions about this matter, please contact the technical contact listed below or the appropriate NRR project manager.

Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical

Contact:

Richard 0. Kiessel, NRR (302) 492-1154

Attachment:

List of Recently Issued NRC Bulletins

Attachment NRC8 85-03. Supplement I April 27, 19F8 LIST OF RECENTLY ISSUED NRC BULLETINS Bulletin Date of No.

Subject Issuance Issued to 87-02, Fastener Testing to 4/22/88 All holders of OLs Supplement 1 Determine Conformance or CPs for nuclear with Applicable Material power reactors.

Specifications 88-03 Inadequate Latch Engagement 3/10/88 All holders of OLs in HFA Type Latching Relays or CPs for nuclear Manufactured by General Power reactors.

Electric (GE) Company 88-02 Rapidly Propagating Fatigue 2/5/88 All holders of OLs Cracks in Steam Generator or CPs for W-desioned Tubes nuclear power reactors with steam generators having carbon steel support plates.

88-01 Defects in Westinghouse 2/5/88 All holders of OLs Circuit Breakers or CPs for nuclear power reactors.

87-02 Fastener Testing to 11/6/87 All holders of OLs Determine Conformance or CPs for nuclear with Applicable Material power reactors.

Specifications 87-01 Thinning of Pipe Walls in 7/9/87 All licensees for Nuclear Power Plants nuclear power plants holding an OL or CP.

86-04 Defective Teletherapy Timer 10/29/86 All NRC licensees That May Not Terminate Dose authorized to use cobalt-60 teletherapy units.

86-03 Potential Failure of Multiple 10/8/86 All facilities ECCS Pumps Due to Single holding an CL or Failure of Air-Operated Valve Cp.

in Minimum Flow Recirculation Line CAL = Operating License CP = Constructicn Permit