ML17334B684

Forwards Addl Info Re CAL Item 3,involving Ability of Plant to Be Cooled Down in 36 H,Per NRC 980109-23 Insp

ML17334B684
Person / Time
Site:	Cook
Issue date:	01/29/1998
From:	Fitzpatrick E INDIANA MICHIGAN POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
AEP:NRC:1260G8, CAL, NUDOCS 9802120308
Download: ML17334B684 (9)

Text

CATEGORY 1 ~

REGULATOR INFORMATION DISTRIBUTIO?PKYSTEM (RIDS)

ACCESSION NBR:9802120308 DOC.DATE: 98/01/29 NOTARIZED: NO DOCKET FACIL'50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 650-316'onald C. Cook Nuclear Power Plant, Unit 2, Indiana,M 05000316 AUTH.NAME AUTHOR AFFILIATION FITZPATRICK,E.E Indiana Michigan Power Co.

RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Forwards confirmatory action resporise validation insp for RAI re item 3, "36 Hour Cooldown.

DISTRIBUTION CODE: IE36D COPIES RECEIVED:LTR TITLE: Immediate/Confirmatory Action Ltr (50 i ENCL J STEE-.

Dkt-Other Than Emergency Prepar NOTES:

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ID CODE/NAME LTTR ENCL ID CODE/NAME' LTTR ENCL PD3-3 PD 1 T HICKMAN<J 1 1 0

INTERNA : FILE CE 01 1 1 NRR/DRPM/PECB 1 1 R

T EIB 1 1 EXTERNAL: NOAC 1 1 NRC PDR D

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NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 7 ENCL 7

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Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INSlANA N1CNESSl N PQWM January 29, 1998 AEP:NRC:1260G8 Docket Nos.: 50-315 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop 0-Pl-17 Washington, D.C. 20555-0001 Gentlemen:

Donald C. Cook Nuclear Plant Units 1 and 2 CONFIRMATORY ACTION RESPONSE VALIDATION INSPECTION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ITEM 3, "36 HOUR COOLDOWN" During the time period January 9, 1998, through January 23, 1998, the NRC conducted a confirmatory action letter (CAL) followup inspection. During the exit meeting for the inspection, on January 26, 1998, we were requested to docket additional information regarding CAL item 3, involving the ability of the plant to be cooled down in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Attachment 1 to this letter contains the information requested.

Attachment 2 contains a flow chart depicting the information in attachment 1.

Sincerely, EF~pm '.

E. Fitzpatrick Vice President

/vlb Attachments J. A. Abramson A. B. Beach MDEQ - DW fc RPD NRC Resident Inspector J. R. Sampson 9802i20308 980i29 PDR ADQCK 050003i5 )

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ATTACHMENT 1 TO AEP:NRC:1260G8 ADDITIONAL INFORMATION RELATED TO 36 HOUR COOLDOHN CONFIRMATORY ACTION LETTER ITEM 3 to AEP:NRC:1260G8 Page 1 Introduction Item 3 of the confirmatory action letter (CAL) dealt specifically with the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown. During the CAL validation inspection, several concerns were raised by the NRC relating to this issue.

This letter is provided to assist with the characterization of concerns. 'hose From August 4, 1997, through September 12, 1997, the NRC conducted an architect engineering (AE) inspection at Cook Nuclear Plant.

During that inspection, several deficiencies were identified related to the component cooling water (CCW) system. Two such deficiencies were:

the 36 hour cooldown analysis completed by Westinghouse for Cook Nuclear Plant contained errors, and no maximum cooling flow limits existed to protect the components cooled by CCW from high, potentially damaging flow rates. Subsequently, concurrent efforts were undertaken to correct the cooldown analysis and establish maximum flow limits.

On August 29, 1997, condition report (CR) 97-2378 was written to capture the AE inspector's concern that maximum flow limits for the CCW system did not exist. The condition report investigation was completed on December 20, 1997. As part of that investigation, technical reviews were completed that established an upper flow limit for each of the components cooled by the CCW system. A commitment was made in that CR response to incorporate the upper flow limits into the updated final safety analysis report (UFSAR).

The due date for that commitment is April 1999. The April 1999 date corresponds to the next planned revision of the UFSAR. As part of the process of updating the UFSAR, a 10 CFR 50.59 evaluation is performed.

On September 19, 1997, the NRC issued a CAL to Cook Nuclear Plant.

The CAL detailed eight specific items that require resolution prior to restart, one of which is the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown issue. In our letter AEP:NRC:1260G3, dated December 2, 1997, we provided our CAL response to the NRC. From January 9, 1998, to January 23, 1998, the NRC conducted a CAL validation inspection to review and verify the CAL response.

Discussion During the validation inspection, the NRC reviewed CAL Item 3, 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown. The review was two-fold. First, the NRC reviewed the actual plant cooldown calculation completed by Westinghouse, as well as the supporting documentation that had been compiled by both us and Westinghouse. That documentation included Westinghouse's safety evaluation check list (SECL)97-189, and our 10 CFR 50.59 evaluation completed for the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown. It was determined that, during a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown, the CCW supply temperature may exceed the previously analyzed 95~ F and reach a maximum of 120~ F.

As part of design change 12-DCP-855, the necessary technical reviews and plant modifications were completed to allow the CCW to AEP:NRC:1260GB Page 2 supply temperature to reach 1204 F. In this letter, the 10 CFR 50.59 evaluation that was completed for 12-DCP-855 is referred to as the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown safety evaluation.

Upon reviewing the Westinghouse cooldown calculation, the NRC questioned two of the inputs used in the analysis. One question dealt with the total CCW flow. In the analysis, the total CCW flow rate was 8,000 gpm (i.e., 4.0E6 lb./hr.). This value was taken directly from table 9.5-3 of the UFSAR. The NRC expressed concern that there was no margin between the analysis input value and the UFSAR value. To address the concern, a 10 CFR 50.59 evaluation was expedited to revise the UFSAR to allow 9,000 gpm of total CCW flow through a single CCW heat exchanger. As discussed above, the need for this 10 CFR 50.59 review had already been recognized and committed to as part of the investigation of CR 97-2378.

important to recognize that the analysis input of 8,000 gpm was It is taken directly from the UFSAR and that the 9,000 gpm value is to be added to the UFSAR to show that margin exists in the cooldown analysis.

Another question dealt with the CCW cooling flow to the residual heat removal (RHR) heat exchangers. This value was 5,000 gpm in the cooldown analysis. The NRC inquired whether instrument uncertainty had been applied to the flow value. The NRC commented that at other plants it was not uncommon to see a flow uncertainty on the order of 1,000 gpm. To address the concern, we provided the NRC with information showing that the uncertainty in the flow indication loop was only on the order of p200 gpm. Additionally, the CCW flow to the RHR heat exchanger is on our critical parameters list. Per a previous commitment made to the NRC in our letter AEP:NRC:1260G3, dated December 2, 1997, instrument uncertainty will be incorporated into the operating procedures for each value on the critical parameters list. This commitment has a due date of December 31, 1998. Thus, the issue about uncertainty in the CCW cooling flow rate to the RHR heat exchangers is already scheduled to be addressed under a pxevious commitment. to the NRC.

In addition to reviewing the actual cooldown calculation, the NRC also reviewed the supporting documentation. Included in that documentation was SECL 97-189 completed by Westinghouse to support operation of the CCW system at. the elevated supply temperature of 1204 F. In their SECL, Westinghouse stated the control valve.

regulating the CCW flow to the letdown heat exchanger will go full open during a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown. The NRC noted that Westinghouse assumed, with the valve full open, a total CCW flow of 1,000 gpm through the letdown heat exchanger. The inspector pointed out that table 9.5-2 of the UFSAR lists the design CCW flow to the letdown heat exchanger as 984 gpm and questioned why the apparent discrepancy was not called out in the safety review. The following explanation was provided. The actual design cooling flow rate for the letdown heat exchanger is 492,000 lb./hr. - a mass flow rate, not a volumetric flow rate. This is shown in table 9.2-3 of the UFSAR. The flow rate of 984 gpm, as shown in table 9.5-2 of the UFSAR, is the volumetric flow rate that corresponds to 492,000 lb./hr. at standard conditions. The flow rate of 1,000 gpm, as shown in the SECL, is the volumetric flow rate that corresponds to 492,000 lb./hr. at the CCW return temperature expected during a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown. Thus, the volumetric flow rates of 984 gpm and 1,000 gpm are both correct as they are both derived from the design mass flow rate specified for the letdown heat exchanger.

to AEP:NRC:1260G8 Page 3 As a follow up question to the one above, the NRC asked how much CCW flow is expected to pass through the letdown heat exchanger with the control valve full open. A preliminary calculation determined with the control valve full open, approximately 1,400 gpm would pass through the letdown heat exchanger. The NRC further questioned why the flow rate, which is greater than the value listed in the UFSAR, was not evaluated in the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown safety evaluation. To address the immediate concern, a 10 CFR 50.59 evaluation was expedited to evaluate twice the design cooling flow rate through the letdown heat exchanger (approximately 2,000 gpm). As discussed above, the need for this 10 CFR 50.59 review had already been recognized and committed to as part of the investigation of CR 97-2378. Although the letdown heat exchanger cooling flow rate of 1,400 gpm was not specifically addressed in the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown safety review, a bounding analysis had been completed and a commitment to revise the UFSAR already existed. It is important to recognize that, although the increased CCW flow rate to the letdown heat exchanger is a consequence of the higher CCW supply temperature, it is not an input in the cooldown analysis.

ATTACHMENT 2 TO AEP:NRC:1280G8 FLOW CHART DEPICTING INFORMATION PROVIDED IN ATTACHMENT 1

r hccacbsienc 2 so AEP/NRCI126008 . QUESTION: The ANSWER: The CCW ihw to the RHR heat cakuhtion assumes exchangers is on the critical parameters gsL All 5000 g pm of CCW procedures which reference the 5000 gpm used coohng flow to the RHR in the analysts wfll be revised to account for heat exchanger. Was instrument uncertainty. This ls a gbbal hstrument uncertainty comnitment to the NRC which is due by the appfied? end of 1998.

NRC Inspector QUESllON; The cooldown analysis assumes RESOLUTION: The Reviewed 50.59 review to aflaw a total COW lhwof 8000 gpm. The FSAR Westinghouse aflaws a total CCW flaw of 809 gpm through 9000 g pm through the

'0 Plant Cooldown the COW heat exchanger. Wiry is there no I CCW heat exchanger Analysis I was expedited.

margin between the osculation hpul and the I phot design basis? I (Comphted 1/23/98)

I NRC/AE,INSPECTION I August- September CAL ITEM ¹3 I 36 Hour Cooldown I 1997 I I I

I I NRC Inspector I Reviewed QUESTION: Westinghouse stated that the I Westinghouse COW control valve for the letdown heat I RESOLUTION: As shawn h the FSAR, the SECL and Our exchanger wgl go fufl open. Using a CCW flaw I I design flaw for the letdown heat exchanger Safety rate of 1000 gpm to the letdown heat is 492,000 ib/hr. This is equfiratent to 984 flow exchanger, they cahuhted the magnum Evaluation letdown temperature. The FSAR states that the gprn at 76 'F and 1000 gpm at 120'F.

Westinghouse correcfly used 492,000 Ib/hr CCW Ihw to the letdown heat exchanger Is 984 to cahutate Ihe temperature.

gpss Why wasnl this addressed in lhe safety evahation?

QUESTION: If the ANSWER With the control RESOLUTION: The control valve does go 50.59 review to aihw valve full open, the CCW fufl open, what wifl be twice design flaw rate wg be approxhnatety the CCW fhw rate to through the letdown 1400 gpm..This Is greater the letdown heat than the design fhw rate heat exchanger was exchanger? shown h the FSAIL expefited.

(Completed 1/23lg The maxhxrm aflawabfe ihw through the CCW heat exchanger is 9000 gpm.

Subsequent to the hspection, a Completed 12/2N97, A part F convnitment was technkst review was completed made In corxfitian report 97-to determine a maximum 2378 to hcorporate the aflowabh caofing ihw rate for maximum flaw fimits hto the each component served by FSAIL CCW. (CR 97-2378) The maximum aflawable Commitment made: 11/23I97 llaw through cooflng fhw Commitment due: Apnl 1999 through the letdown heat exchanger Is twice design (l.e. 984,000 lb/br).

ted 12/20/97.