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Forwards Response to NRC Re Violations Noted in Insp Repts 50-315/96-12 & 50-316/96-12 on 961021-25 & 961205.Corrective Actions:Crs 96-1699 & 96-1701 Were Written on Power Operated Relief Block Valves
	ML17333A755
Person / Time
Site:	Cook
Issue date:	01/27/1997
From:	Fitzpatrick E; INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	AEP:NRC:1238H, NUDOCS 9701300164
	Download: ML17333A755 (17)
	v • d • e

Indiana Michigan Power Company SCO Circle Drive 8uchanan, Ml 49107 1395 pR OR~ gL)L ~

econd Ol INDIANA NICHIGAN POWER January 27, 1997 AEP: NRC: 1238H 10 CFR 2.201 Docket Nos 50-315 50-316 U. S. Nuclear Regulatoory Commission Washington D

Gentlemen:

Donald C.

Cook N SPECTION REPORT NOS.

oo Nuclear Plant Un'ND 50-316/96012 (DRPi This letter is REPLY TO NOTTICE OF VIOLATION o a letter from G of t forwarded a noti f

y.

The violation f

r ompan

'ce o violati otor operated l

uring the m

o NRC re A.

Dunlop A.

G rough 25, an

uzzman, and and December 5

199 y

performance of a

untimel 6..

The pon receipt of an updated l

e va ve facto i

letter.

e violations is provided in t vi e in the attachment to this Sincerely, ee E.

E. Fitz trick Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS ~~DAY OF 1997 Notary Public My Commission Exp'pires:

Attachment JAN WATSON NOTARYPUBUC, BERRIEN COUNTY, MI MYCOMMISSION EXPIRES FEB. 10, 199'~

A. A. Blind A. B. Beach MDEQ -

DW

& RPD NRC Resident Inspector R. Padgett

ATTACHMENT TO AEP:NRC: 1238H REPLY TO NOTICE OF VIOLATION:

NRC INSPECTION REPORT NOS.

50-315/96012 (DRP)

AND 50-316/96012 (DRP)

yi

Attachment to AEP:NRC:1238H Page 1

'During an NRC inspection conducted

. rom October 21-2

1996, one violation of NRC requirements was identified.

In accordance with the 'General Statement of Policy and Procedure for NRC Enforcement Actions,'UREG-1600, the violation is listed below."

Our response follows.

NRC Violation "D.

C.

Cook Nuclear Plant technical specification (T/S) 6.8.1

states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in appendix A of regulatory guide (RG) 1.33, Revision 2, February 1978.

RG 1.33, appendix A, paragraph 1.b, requires that a procedure be written delineating the authorities and responsibilities for safe operation and shutdown.

Plant Manager Instruction (PMI)

7030,

'Corrective Action,'as written in accordance with RG 1.33.

Step 6.9.a requires an or~ginator to initiate a condition report for known or suspected adverse conditions/events.

Step 5.31 requires, in part, that a

prompt operability determination

'...must be made expeditiously foilowing identification of a potentially degraded condition that has the potential to impact SSC operability.'ontrary to the above, as of October 21, 1996, the licensee failed to initiate a condition report and properly perform and document a

prompt operability determination for the Unit 1 power operated relief valve block valve when valve factor information was obtained that had a potential adverse effect on the operability of the valves to perform the required design-basis function.

This is a Severity Level IV violation (Supplement I).

(50-315/960 12-01 (DRS); 50-316/960 12-01 (DRS))."

Res onse to NRC Violation Admission or Denial of the Alle ed Violation Indiana Michigan Power Company admits to the violation as cited in the NRC notice of viola ion.

Reason for the Violation An operability determination for motor operated valve (MOV) 1-NMO-152, power operated relief block valve, using a 0.4 valve factor was documented in condition report (CR) 96-0594 in April 1996.

During a previous MOV GL 89-10 NRC inspection, the use of the vendor supplied valve factor of 0.309 was questioned.

The NRC requested that the use of the 0.309 valve factor be re-reviewed.

The review found differential pressure testing of a similar valve by EPRI yielded valve factors of 0.267 to 0.296.

Contact with EPRI found the test data was valid but may not be as conservative as the EPRI performance predictive model (PPM) algorithm.

The EPRI PPM coordinator indicated a valve factor of 0.4 was more appropriate.

The valve factor of 0.4 was applied as the design basis to the MOV thrust calculation and the results compared to the current thrust settings.

V

Attachment to AEP:NRC:1238H Page 2

During the continuing review o best available design basis information for the closure of GL 89-10, it was determined the application of the EPRI PPM would be appropriate because only the EPRI differential pressure testing of a

"similar" valve was used as "best available data".

These valves cannot be tested under differential pressure conditions at Cook Nuclear Plant and the only available differential pressure test data that could be found in the industry was the EPRZ testing.

During the week of October 6,

1996, while working on the GL 89-10 closure

document, we were verbally notified by the contractor performing the EPRZ PPM algorithm that a valve factor of 0.51 for the power operated relief block valve was predicted.

At this

time, the design validation review of the calculation was still in progress.

The review was completed on October 11, 1996.

During subsequent discussions within engineering, it was determined that because the EPRI PPM algorithm was performed with information applicable to the new design to be installed under design change package

007, the new EPRZ predicted valve factor 0.51 would be only applied after the design change was implemented.

The 0.4 valve factor from EPRI differential pressure testing of a "similar" valve would remain as the design basis valve factor until design change package 007 was implemented.

The 0.4 valve factor as an operability basis was documented under CR 96-0594.

Based on this

approach, no operability concern

existed, and therefore, no CR was generated.

During the GL 89-10 MOV close-out inspection, it was found the EPRI and Cook Nuclear Plant valves were not the same size or model.

This invalidated the previous methodology, and the use of the 0.4 valve factor, which had been the best available data until this time.

The EPRI PPM algorithm value of 0.51 then became t'ae best available design basis

data, and was applied to the power operated relief valve thrust calculation.

The misapplication of the valve factor information resulted in our failure to initiate a CR and document an operability determination review.

Corrective Actions Taken and Results Achieved On October 24,

1996, CRs 96-1699 and 96-1701 were written on the power operated relief block valves concerning the change in valve factor from 0.4 to 0.51.

An operability review was performed, using the 0.51 valve factor, under PMI-7030 and PMSO.173.

This review found 1-NMO-152 to be inoperable under full design temperature and pressure.

A licensee event report was submitted, valve 1-NMO-152 was closed, and power was removed from the actuator as required by the T/S.

An analysis has been performed to allow the use of this valve under 1500 psid for LTOP service.

Attachment to AEP:NRC:1238H Page 3

To ensure additional occurrences of incorrect assumptions had not occurred, a review of our GL 89-10 best available data for valves factors was performed.

The review noted one additional valve model (Conval) where the valve factor was based solely on EPRZ testing of "similar" valves.

Additional basis for the Conval valve factor was developed by review of differential pressure testing by other nuclear plants.

The valve factor was found to be 1.3 versus 1.1 currently being used.

An additional condition report, CR 96-2087, was generated and an operability review completed.

The review showed the current actuator setting for the Conval valves is correct, and the valves will perform their design function.

Corrective Actions to Avoid Further Violations Corrective actions previously taken in response to the notice of violation contained in inspection report 96006 were judged to be adequate to ensure condition reports are initiated as required, with the exception of the highly technical issues associated with MOV thrust requirements.

The managers and engineers responsible for MOV's have been instructed to initiate a condition report when adverse information is received concerning MOV operability.

The CR will ensure timely operability determination will be performed and documented in accordance with PMZ-7030 and PMS0.173.

Date When Full Com liance will be Achieved Full compliance was achieved on December 11, 1996, with the issuance of CR 96-2087.

0

~

CATEGORY REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCFSSION NBR:9701300164 DOC.DATE: 97/Ol/27 NOTARIZED: YES

- 'OCKET FACIL:50-315 Donald C.

Cook Nuclear Power Plant, Unit 1, Indiana M

05000315 50-316 Donald C.

Cook Nuclear Power Plant, Unit 2, Indiana M

05000316 AUTH.NAME AUTHOR AFFILIATION FITZPATRICK,E.

Indiana Michigan Power Co. (formerly Indiana a Michigan Ele RECZP.NANE RECZPZENT AFFZLZATZON Document Control Branch (Document Control Desk)

SUBJECT:

Forwards response to NRC 961227 ltr re violations noted in insp repts 50-315/96-12 !'0-316/96-12 on 961021-25 a

961205.Corrective actions:CRs 96-1699

& 96-1701 were written on power operated relief block valves.

I DISTRIBUTION CODE:

IE01D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response NOTES:

E INTERNAL:

EXTERNAL:

RECIPIENT ID CODE/NAME PD3-3 PD AEOD/SPD/RAB DEDRO NRR/DISP/PIPB NRR/DRPM/PECB NUDOCS-ABSTRACT OGC/HDS2 LITCO BRYCE,J H

NRC PDR COPIES LTTR ENCL 1

1 1

1 RECIPIENT ID CODE/NAME HICKMAN,J AEOD -

LE CENTER CH HHFB NRR/DRPM/PERB OE DIR RGN3 FILE 01 NOAC COPIES LTTR ENCL 1

1 1

1 1'

1 1

D 0

U E

N NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE!

CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT. 415-2083)

TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED:

LTTR 17 ENCL 17

Indiana Michi Power Compan 500 Circle Drive Buchanan, Ml 491071395 INDIANA NICHIGAN POWER January 27, 1997 Docket Nos.:

50-315 50-316 AEP: NRC: 123 8H 10 CFR 2.201 U.

S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.

C.

20555 Gentlemen:

Donald C.

Cook Nuclear Plant Units 1 and 2

NRC INSPECTION REPORT NOS.

50-315/96012 (DRP)

AND 50-316/96012 (DRP)

REPLY TO NOTICE OF VIOLATION This letter is in response to a letter from Geoffrey E. Grant dated December 27, 1996, that forwarded a notice of violation to Indiana Michigan Power Company.

The violation of NRC requirements was identified during the motor operated valve closeout inspection conducted by Messrs.

A.

Dunlop, A.

Guzzman, and R.

Cain from October 21 through 25, and December 5,

1996.

The violation is associated with untimely performance of a

prompt operability evaluation upon receipt of an updated valve factor number.

Our reply to the violations is provided in the attachment to this letter.

Sincerely, Vice President, SWORN TO AND SUBSCRIBED BEFORE ME THIs ~~DAY oF X+~la4, 1997 Notary Public My Commission Expires:

Attachment JAN WATSON NOTARYPUBUQ,8ERRIEN COUNTY, Ml MYCOMMISSION EXPIRES FEB. 10, 1999 cc:

A. A. Blind A. B. Beach MDEQ -

DW

& RPD NRC Resident Znspector J.

R. Padgett 970%300ib4 970127 PDR ADOCK 050003i5 8

PDR

ATTACHMENT TO AEP:NRC: 1238H REPLY TO NOTICE OP VIOLATION:

NRC INSPECTION REPORT NOS.

50-315/96012 (DRP)

AND 50-316/96012 (DRP)

Attachment to AEP:NRC:1238H Page 1

"During an NRC inspection conducted from October 21-25,

1996, one violation of NRC requirements was identified.

In accordance with the 'General Statement of Policy and Procedure for NRC Enforcement Actions,'UREG-1600, the violation is listed below."

Our response

.follows.

NRC Violation "D.

C.

Cook Nuclear Plant technical specification (T/S) 6.8.1

states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in appendix A of regulatory guide (RG) 1.33, Revision 2, February 1978.

RG 1.33, appendix A, paragraph 1.b, requires that a procedure be written delineating the authorities and responsibilities for safe operation and shutdown.

Plant Manager Instruction (PMI)

7030,

'Corrective Action,'as written in accordance with RG 1.33.

Step 6.9.a requires an originator to initiate a condition report for known or suspected adverse conditions/events.

Step 5.31 requires, in part, that a

prompt operability determination

'...must be made expeditiously following identification of a potentially degraded condition that has the potential to impact SSC operability.'ontrary to the above, as of October 21, 1996, the licensee failed to initiate a condition report and properly perform and document a

prompt operability determination for the Unit 1 power operated relief valve block valve when valve factor information was obtained that had a potential adverse effect on the operability of the valves to perform the required design-basis function.

This is a Severity Level ZV violation (Supplement I).

(50-315/960 12-01 (DRS); 50-316/960 12-01 (DRS) ). "

Res onse to NRC Violation Admission or Denial of the Alle ed Violation Indiana Michigan Power Company admits to the violation as cited in the NRC notice of violation.

Reason for the Violation An operability determination for motor operated valve (MOV) 1-NMO-152, power operated relief block valve, using a 0.4 valve factor was documented in condition report (CR) 96-0594 in April 1996.

During a previous MOV GL 89-10 NRC inspection, the use of the vendor supplied valve factor of 0.309 was questioned.

The NRC requested that the use of the 0.309 valve factor be re-reviewed.

The review found differential pressure testing of a similar valve by EPRZ yielded valve factors of 0.267 to 0.296.

Contact with EPRZ found the test data was valid but may not be as conservative as the EPRI performance predictive model (PPM) algorithm.

The EPRI PPM coordinator indicated a valve factor of 0.4 was more appropriate.

The valve factor of 0.4 was applied as the design basis to the MOV thrust calculation and the results compared to the current thrust settings.

Attachment to AEP:NRC:1238H Page 2

During the continuing review of best available design basis information for the closure of GL 89-10, it was determined the application of the EPRI PPM would be appropriate because only the EPRI differential pressure testing of a

. "similar" valve was used as "best available data".

These valves cannot be tested under differential pressure conditions at Cook Nuclear Plant and the only available differential pressure test data that could be found in the industry was the EPRI testing.

During the week of October 6,

1996, while working on the GL 89-10 closure

document, we were verbally notified by the contractor performing the EPRI PPM algorithm that a valve factor of 0.51 for the power operated relief block valve was predicted.

At this

time, the design validation review of the calculation was still in progress.

The review was completed on October 11, 1996.

During subsequent discussions within engineering, it was determined that because the EPRI PPM algorithm was performed with information applicable to the new design to be installed under design change package

007, the new EPRI predicted valve factor 0.51 would be only applied after the design change was implemented.

The 0.4 valve factor from EPRI differential pressure testing of a "similar" valve would remain as the design basis valve factor until design change package 007 was implemented.

The 0.4 valve factor as an operability basis was documented under CR 96-0594.

Based on this

approach, no operability concern

existed, and therefore, no CR was generated.

During the GL 89-10 MOV close-out inspection, it was found the EPRI and Cook Nuclear Plant valves were not the same size or model.

This invalidated the previous methodology, and the use of the 0.4 valve factor, which had been the best available data until this time.

The EPRI PPM algorithm value of 0.51 then became the best available design basis

data, and was applied to the power operated relief valve thrust calculation.

The misapplication of the valve factor information resulted in our failure to initiate a CR and document an operability determination review.

Corrective Actions Taken and Results Achieved On October 24,

1996, CRs 96-1699 and 96-1701,were written on the power operated relief block valves concerning the change in valve factor from 0.4 to 0.51.

An operability review was performed, using the 0.51 valve factor, under PM1-7030 and PMS0.173.

This review found 1-NMO-152 to be inoperable under full design temperature and pressure.

A licensee event report was submitted, valve 1-NMO-152 was closed, and power was removed from the actuator as required by the T/S.

An analysis has been performed to allow the use of this valve under 1500 psid for LTOP service.

J