NRC-98-0094, Provides Addl Info Re Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions, Per 980611 Telcon W/Nrc

Provides Addl Info Re Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions, Per 980611 Telcon W/Nrc

ML20236H860
Person / Time
Site:	Fermi
Issue date:	06/30/1998
From:	Gipson D DETROIT EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-98-0094, CON-NRC-98-94 GL-96-06, GL-96-6, NUDOCS 9807070395
Download: ML20236H860 (6)

Text

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Douglas R. Gipson Senior Vice President, Nuclear Generation fermi 2 ,

04% .Wrt h INxie llwy , Newport, Michigan 4Slui l Tel: 313JMI>201 Fax: 313JM4172 )

Detroit Edsson 7

June 30,1998 NRC-98-0094 l

l U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555

References:

1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43

2) NRC Generic Letter 96-06: Assurance of Equipment l Operability and Containment Integrity During Design-Basis Accident Condition.c, dated September 30,1996 )

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3) Detroit Edison Letter to NRC," Detroit Edison 30-Day Response to NRC Generic Letter 96-06," NRC 96-0118, i dated Octe.,ber 30,1996  ;

4) Detroit Edison Letter to NRC," Detroit Edison 120-Day Respense to NRC Generic Letter 96-06," NRC 97-0003, )

dated. January 28,1997 l

5) NRC Letter to Detroit Edison," Request for Additional Infor. nation Related to the Fermi-2 Response to Generic Letter 96-06," dated September 9,1997 (TAC No. M96811) h

6) Detroit Edison Letter to NRC,"AdditionalInformation Related to l Detroit Edison Response to NRC Generic Letter 96-06," NRC 0'l11, dated October 17,1997

7) NRC Letter to Detroit Edison," Request for Additional

! Information (RAI) Related to the Fermi-2 Response to Generic

, Letter 96-06, Assurance of Equipment Operability and l Containment Integrity During Design-Basis Accident Conditions" dated April 6,1998 (TAC No. M96811) 9807070395 980630 PDR ADOCK 05000341 1 P PDR A ige Energy company l

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l USNRC l NRC-98-0094 l Page 2

• 8 l 8) NRC Meeting with Nuclear Energy Institute (NEI) and l Industry Representatives to Discuss Generic Letter 96-06 Waterhammer and Two-Phase Flow Issues, Washington D C, dated May 27-28,1998

Subject:

Additional Infonnation Related to Detroit Edison Response to NRC Generic Letter 96-06 i

The purpose of this letter is to provide Detroit Edison response to Reference 7, which requested Detroit Edison to provide additional information regarding Fermi 2 response to Generic Letter 96-06. Reference 7 stated the licensee's response indicated that the drywell coolers are not required for accident mitigation and that cocling water to the drywell coolers, provided by the Reactor Building Closed Cooling Water (RBCCW) system, is isolated automatically on high drywell pressure.

However, cince no positive measures have been taken to assure that RBCCW will be l isolated upon event initiation and remain isolated, additional information is requested j as described in Reference 7.

A teleph,one conference (telecon) was held on June 11,1998 between Detroit Edison and the NRC staff (Mr. Andrew Kugler, NRC Project Manager - Fermi 2, and Mr. I James Tatum, NRC Technical Reviewer for Generic Letter 96-06) to discuss Detroit Edison's response to this RAI. It was concluded that Detroit Edison's response as discussed in the telecon and provided below is sufficient and the response to specific

~

Questions 1 through 8 of Re1erence 7 is not required. I 1

Fermi 2 plant design automatic ally isolates the Reactor Building Closed Cooling Water (RBCCW) system header supplying cooling water to the drywell coolers on high drywell pressure (1.68 psig) or Reactor Pressure Vessel low water level 2 signal following a design basis accident and the drywell remains isolated thereafter. Fermi 2 does not depend on the operation of the drywell coolers for the accident mitigation. l As such, Fermi 2 procedu;e prohibits the reopening of the cooling water supply to j the subject coolers even if the condition subsequently clears, during and following ,

the design basis accident. Fermi 2 Emergency Support Procedure on drywell coohng  !

water restoration prr,vides the following caution "If Drywell Pressure reaches 20 i psig AND Torus water temperature exceeds 120

• F, do not operate the Drywell {

l Cooling Fans (even if the condition subsequently clears)." Thus, the waterhammer  !

and two phase flow are not likely to occur even if the worst case single failure  !

criterion were to be considered at the initiation of the design basis accident. l Therefore, no furthec action is required at Fermi 2 at the above provides the l necessary assurance that RBCCW system headers to the drywell are isolated upon l event initiation and remain isolated.

Additionally, for the cases imunded by recirculation line break (DBA Large Break l LOCA) to 0.1 sq. ft. line break ir. side the containment, the combination of 20 psig

USNRC l NRC-98-0094 Page 3 drywell pressure and 120

• F suppression pool temperature would be reached within a short duration before the operator can restore the drywell cooling. In the cases of breaks smaller than 0.1 sq. ft., the combination of 20 psig drywell pressure and 120'F is not possible and the voiding will not occur due to (i) the low thermal input due to the break during the coasting of the fan (short duration) when the coolant flow to the coolers is impaired due to the RBCCW pump stopping, (ii) the thermal expansion of the coolant piping inside the drywell due to the increased ambient temperature, and (iii) the relief valve pressure setting of 150 psig on the supply header being much higher than the saturation pressure corresponding to the drywell temperature. Hence, the present administrative controls are deemed adequate to prevent possible damage due to the waterhammer and two phase flow phenomena.

Recently, an NRC meeting was held with Nuclear Energy Institute (NEI) and industry representatives to discuss Generic Letter 96-06 Waterhammer and Two-Phase Flow Issues (Reference 8). The following conclusion was arrived at in the meeting: for systems that are automatically isolated and shut down post-accident and prohibited from use post-accident, the waterhammer and two phase flow are not f feasible and no further action is required. For those plants where the containment fan coolers will not be relied upon during accident conditions, the NRC expects licensees to submit sufficient information to provide assurance that appropriate mechanisms are in place to prohibit the use of the coolers. Fermi 2 design and current administrative controls in place prohibit use of these coolers.

Further, in the Reference 8 meeting, it was concluded that if the cooling were to be i

restored any time later in the accident, adequate administrative controls must be in place in the fonn of technical guidance either embedded in the procedures or with the technical support center staff to monitor key parameters and to take appropriate sequence of actions to reduce the possibility of damage due to the waterhammer and two phase flow phenomena.

Therefore, if Fermi 2 design configuration were to change in the future, adequate administrative controls would be in place prior to returning the coolers to service post-accident following a design basis accident to assure that waterhammer and two phase flow do not occur. If needed in the future, guidance would be embedded in the Emergency Support Procedure concerning the containment temperature, cooling water system pressure, and cooling system inventory that need to be considered prior to the returning of the coolers to service, to ensure containment integrity.

1 In response to Question 9 of Reference 7, a simplified system sketch is attached to facilitate review. Please note that the seven divi;ional drywell coolers are not shown individually but are grouped with other nonessential loads in the drywell. In addition, the relief valve on the drywell supply header is also not shown on the sketch.

l L_________________.________. -

NRC-98-0094 Page 4 If you have any questions, please contact Mr. Norman K. Peterson, Director-Nuclear Licensing at (734) 586-4258.

! Sincerely, O

Attachment ec: B. L. Burgess G. A. Hanis A. J. Kugler Regional Administrator, NRC Region III i

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USNRC NRC-98-0094 Page 5 l

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I, DOUGLAS R. GIPSON, do hereby affirm that the foregoing statements are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

/M DOUGLAS R. GIPSON Senior Vice President, Nuclear Generation On this O ay o klL 1001998 before me personally appeared Douglas R.

Gipson, being first duly sddrn and says that he executed the foregoing as his free act and deed.

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