ML20211J072

Submits Response to NRC GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions

ML20211J072
Person / Time
Site:	Davis Besse
Issue date:	09/30/1997
From:	Jeffery Wood CENTERIOR ENERGY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2488, GL-96-06, GL-96-6, TAC-M96803, NUDOCS 9710070308
Download: ML20211J072 (5)

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M01 N State Route 2 - 419-249-2300 John K. Wood oak Hartor.oH 43449 FAX: 419-3218337 Vce Prescers. Nuclear -

Does-Besse Docket Number 50-346 License Number NPF 3 Serial Number 2488 4

-September 30, 1997

- United States Nuclear Regulatory Commission Document Control Desk

. Washington, D. C. _20555-0001

-Sub}^ct: Response to NRC Generic Letter 96-06: Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions (TAC No. M96803)

Ladies and Gentlemen:

On September 30,1996, the Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 96-

06. That letter requested licensees, such as those for the Davis-Beve Nuclear Power Station (DBNPS) Unit Number 1, to address the following generic issues:

(1) Cooling water systems serving the containment air coolers (CACs) may be exposed to the hydrodynamic effects of water hammer during either a loss-of-coolant accident (LCCA) or a main steam line break (MSLB). These cooling water syvems were not designed to withstand the hydrodynamic effects of water hammer and corrective actions may be needed to satisfy system design and operability requirements. Licensees are to dstermine if their plant's CACs cooling water systems are susceptible to water hammer during postulated accident conditions.

(2) Cooling water systems serving the containment air coolers may experience two-phase flow conditions during postulated LOCA and MSLB scenarios. The heat removal assumptions for design basis accident scenarios were based on single phase flow conditions. Corrective actb :3 may be needed to satisfy system design and operability requirements. Licensees are to determine if their plant's CACs are susceptible to two-phase flow conditions during postulated accident conditions.

(3) Thermally-inuen 'verpressurization of isolated water-filled piping sections in containment could: 1) jeopardize the ability of accident-mitigating systems to perform their safety functions, and 2) could also lead to a breach of containment integrity via bypass leakage. Corrective fy actions may be needed to satisfy system operability requirements. Lesees are to determine if

- the piping systems which penetrate their plant's containment are susceptible to thermal g;

expansion of fluid so that.pverpressurization could occur.

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Seria! Number 2488 i Page 2

' On January 28,1997 Toledo Edison (TE) provided by letter (Serial Number 439) an interim response to GL 96-06. On February 28,1997, TE provided by letter (Serial Number 2442) a summary report describing the actions taken to that date and the results of those actions. On July 28,1997 TE provided by letter (Serial Number 2473) an update on the status of TE's progress in resolving the issues of GL 96-06 and informing the NRC that plans for closure on the issue were delayed due to an unexpected forced outage that occurred because of a plant trip on May 4,1997.

The following provides a report on the resolution of post-LOCA thermal overpressurization of containment penetrations, and the water hammer and two phase flow effects on the CAC trains.

Thermal Overpressurization of Pining Penetrations l The following table summarizes the current status of and planned actions for the thirteen containment penetrations that TE had idemified previously as potentially susceptible to post-LOCA thermal overpressurization. These are the final planned actions concerning these penetrations.

Penetration Name Status / Planned Actions 14 RCS Letdown Meets ASME Code Faulted Stress Allowables and I ASME Code pressure limitations under faulted conditions. No further action.

56 Reactor Coolant Pump Seal Meets ASME Code Faulted Stress Allowables and Return ASME Code pressure limitations under faulted conditions. No further action.

12 Component Cooling Water Modification completed. Installed Bypass Check to the Control Rod Drive Valve during May 4,1997 outage. No further action.

Mechanisms 21 Demineralized Water to AOV Globe valve-Providt > inherent relicf to Containment prevent overpressurization. No further action.

32 Reactor Coolant Drain to AOV Diaphragm Valve-Provides inherent relief to Reactor Coolant Drain Tank prevent overpressurization. No further action.

49 Refueling Canal Fill Was partially drained. Procedure changes to assure piping remains partially drained will prevent poteatial overpressurization.

74C Pressurizer Auxiliary Spray Meets DBNPS Interim Stress Allowables. Procedure changes to assure piping remains partially drained during plant operation will prevent potential overpressuriration.

1 Pressurizer Sample Line Meets ASME Code Faulted Stress Allowables, however, procedure changes will assure weekly sampling ia coordinated with valve stroke testing to assure the penetration is isolated with the fluid at a higher temperature than possible after a design bases accident.

13 Containment Noimal Sump Meets DBNPS Interim Stress Atlowables. Install Relief Valve during iIRFO to achieve full ASME code compliance.

Docket Number 30-346 License Number NPF-3 Serial Number 2488 Page 3 Penetration Name Status / Planned Actions 3 Component Cooling Water , Meets ASME Code Faulted Stress Allowables. Install to Containment Dypass Check Valve during iIRFO to achieve full ASME code compliance.

4_ Component Cooling Water Current documented LLRT leakage will prevent over.

Return from Containment pressurization. Install Bypass Check Valve during 11RFO to nehieve full ASME code compliance.

47a Core Flood Tank Sample Meets DBNPS Interim Stress Allowables. Install Line Bypass Check Valve during 11RFO to achieve full

,, ASMB Code compliance.

48 Pressurize Quench Tank AOV Globe valve providing inherent relief to prevent-Outlet overpressurization in the interim. Further engineering evaluation has determined that installation cf a Bypass Check Valve during IIRFO to ar$ieve full ASME code compliance is the preferred long term solution for this penetration.

l The planned actions described in the table above will be implemented prior to startup from the -

Eleventh Refueling Outage (11RFO)in May,1998.

l This concludes the extent of modifications to address the containment penetrations that may be affected by post-LOCA thermal overpressurization. With the implementation of the modifications and procedure changes cited above during the 11RFO, TE consiSrs this issue closed.

1 Containment Air Cooler Water Hammer and Two Phase Flow Effects

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A thermal-hydraulic evaluation _was performed on the service water piping of the CACs. The thermal hydraulic evaluation results were used to evaluate the acceptability of the system piping stresses and the support /res'raint loads. The results show that the CAC-related piping meets ASME Code Section III allowable stresses. The structural review determined that the pipe support loads are within the present restraint capacities. All CAC-related pipe anchors have been evaluated and, except for anchor A160, all loads are within the anchor capacities. Anchor A160 has one weld that potentially could become over-stressed, however, the stresses are within the DBNPS interim stress allowables for continued operation. A more rigorous evaluation is currently being performed and will be completed, along with any potential modifications to the anchor, prior to startup from the 11RFO,' scheduled for May 1998.

As reported to the NC in TE letter dated January 28,1997, (Serial Number 2439), a review of the CACs design and openition during the bounding design basis LOCA conditions indicate that the CACs are not susceptible to continued two-phase flow. Even if two-phase flow conditions should develop during a loss of power scenario, single phase flow would be restored at approximately the time the CACs are credited in the containment analysis. The CACs were also evaluated for any detrimental affects from the waterhammer loads on the adjacent piping and found to be acceptable.

Therefore, the CACs remain operable, a

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- : Docket Number 50-346- ,

c Lictnse Number NPF-3 :  !

, . Serial Number 2488 ?

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. Summarv o j With the actions taken to date, the containment penetrations potentially subject to thermal  !

Loverpressurir.ation remain operable. The procedure and design changes planned for implementation -;

. during the 11RFO will ensure full ASME Code compliance.  !

4 Analysis indicates that although waterhammer transients may occur, they _will not cause piping or -

restraint over-stress, except as discussed previously for anchor A160. The CAC system is not j susceptible to significant heat transfer reductions due to two-phase flow. Therefore, the CAC system'will remain operable during the bounding design basis accident conditions.

Toledo Edison will provide a summary letter to the NRC detailing all actions taken to ensure compliance with the reauirements of Generic Letter 96-% within 60 days after startup from the l lRFO Should you have any questions or require additional information, please contact

- Mr. James L Freels, Manager - Regulatory Affairs, at (419) 321-8466.

Ve ly yours, l K/laj l

- cc: A. B. Beach, Regional Administrator, NRC Region III

- A. G. Hansen, DB-1 NRC/NRR Project Manager

- S. Stasek, DB-1 NRC Senior Resident Inspector

• Utility Radiological Safety Board '

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- Page 5

RESPONSE

TO NRC GENERIC LETTER 96-06 FOR T11E DAVIS BESSE NUCLEAR POWER STATION UNIT NUMBER 1 This letter is submitted pursuant to 10 CFR 50.54(f) and contains information pursuant to NRC Generic Letter 96-06," Assurance of Equipment Operability and Containment Integrity During Design - Basis Accident Conditions" for the Davis-Besse Nuclear Power Station, Unit Number 1.

By:

JM[ood, Vice Presid'ent - Nuclear Sworn to and subscribed before me this 30th day of September,1997.

Am M Notary Public, State ofDhio - Nora Lynn Flood My-commission-expires September 4, 2002.

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