NSD-NRC-97-5164, Forwards Rev 1 of WCAP-14837, AP600 Shutdown Evaluation Rept & Response to NRC Shutdown Risk Assessment Question Received on 970121 Which Completes W Action Related to NRC Requests for Analyses & Procedures

Forwards Rev 1 of WCAP-14837, AP600 Shutdown Evaluation Rept & Response to NRC Shutdown Risk Assessment Question Received on 970121 Which Completes W Action Related to NRC Requests for Analyses & Procedures

ML20148J069
Person / Time
Site:	05200003
Issue date:	06/06/1997
From:	Mcintrye B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20148J075	List: ML20148J081 NSD-NRC-97-5164, Forwards Rev 1 of WCAP-14837, AP600 Shutdown Evaluation Rept & Response to NRC Shutdown Risk Assessment Question Received on 970121 Which Completes W Action Related to NRC Requests for Analyses & Procedures
References
NSD-NRC-97-5164, NUDOCS 9706160168
Download: ML20148J069 (3)

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, u gh Pennspvania 15230 0355 NSD-NRC-97-5164 DCP/NRC0897 Docket No.: STN-52-003 June 6,1997 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T. R. QUAY

SUBJECT:

SUBMITTAL OF WCAP-14837 REVISION I AND RESPONSE TO NRC SHUTDOWN RISK ASSESSMENT QUESTION RECEIVED 1/21/97

References:

1.

Letter DCP/NRC0809,"AP600 SilUTDOWN EVALUATION REPORT (WCAP-14837) AND RESPONSE TO RAI 440.53", dated 4/15/97.

2.

Letter from NRC to Westinghouse, "NRC COMMENTS ON Tile AP600 STANDARD SAFETY ANALYSIS REPORT (SSAR) CIIAPTER 15 ACCIDENT ANALYSES", dated 1/21/97.

Dear Mr. Quay:

Enciosure 1 is Revision 1 of WCAP-14837, "The AP600 Shutd< wn Evaluation Report," for NRC review. As noted in Reference 1, this report is updated to include:

1. Section 4.8.5 for loss of normal residual heat removal system (RNS) scenarios.

2. Changes to sections 2.1.2.2 and Table 8.2-1 to reflect the addition of actuating fourth-stage ADS valves on low (empty) RCS hot leg level on a twcr aut-of-two basis (this change was a result of the loss of RNS analysis).

3. Changes to page 4.10-3 to reflect the change ofIRWST gutter drain return from nonsafety-related to safety-related.

4. Changes to Section 8.1 to reDect references 8.1-8 and 8.1-9, issuing RAI 440.120 response Revision 1 and responding to NRC question number 53 of Reference 1 regarding shutdown risk f

assessment.

l That response is attached for NRC review and completes Westinghouse action related to NRC Y[]

requests for 1) analyses and procedures addressing inadvertent isolation of the normal residual heat removal system (RNS) which could be caused by spurious containment high radiation signal resulting from radiation release spikes during outage activities and 2) explanation of how an RNS pipe break outside containment is mitigated if the containment radiation isolation signal does not actuate to isolate RNS and terminate the break.

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DCP/NRC0897 June 6,1997 l

This submittal completes the Westinghouse action related to Open Item Tracking System items 2295, l'

2296,3007, and 4185 related to completion of WCAP-14837 Revision 1 and item 4524 for the response to Question 53 of Reference 1. The NRC is requested to review this Shutdown Evaluation l

Report revision and the attached response to Question 53.

l Please contact Robin K. Nydes (412-374-4125) is you have any questions regarding this transmittal.

(44A/K.

Brian A. McIntyre, Manager Advanced Plant Safety and Licensing jml Attachment Enclosure cc:

William liuffman, NRC (SL, SE)

Summer Sun, NRC (ll, IE)

Marie Pohida, NRC (IL,1E)

Gene lisii, NRC (2L,2E)

Nicholas J. Liparulo, Westinghouse (w/o Enclosure) r nnA pr J

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( to letter DCP/NROdFMK l

NRC Question 53 To address Open Item 1002 in the DSER regarding the need of automatic containment isolation for non-safety related penetrations, Westinghouse responded, in a letter dated October 10, 1996, that normal RHR and CVS charging isolation valves will isolate on a containment radiation signal permitting these non-safety systems to continue to operate when in service unless there is an elevated radiation level present.

The staff is concerned that a spurious j

containment high radiation signal caused by radiation release spikes during outage activities may inadvertently isolate the normal RHR.

Westinghouse is requested to discuss its analysis and/or plant procedures to justify the mitigation strategies for this inadvertent isolation of the RHR event during the refueling outage.

1 For a postulated RHR pipe break outside the containment, the conteinment radiation isolation signal will not actuate to isolate the RHR and terminate the break.

Explain how this LOCA outside the containment is mitigated and discuss the need for automatic isolation of the RHR based on a low reactor water level signal.

Westinghouse Response 1

The signal that isolates the normal reridual heat removal system (RNS) l lines which penetrate containment on a high-2 containment radiation signal is manually blocked below the P-11 pressure interlock prior to l

the RNS being placed into service for shutdown decay heat removal.

i This minimizes the probability of a spurious signal that could cause a loss of shutdown cooling.

In the unlikely event of a loss of RNS during shutdown, the passive safety systems are designed to mitigate the consequences of the event as shown in section 4.8.5 of the l

Shutdown Evaluation Report (WCAP-14837 Revision 1).

A postulated break of RNS piping outside containment is not a credible l

event.

The AP600 RNS design pressure is 900 psig, with an ultimate rupture strength greater than the normal RCS operating pressure.

The maximum expected operating pressure for the RNS is 450 psig, and for the majority of the time when the system operates (>90%), the RCS pressure is at atmospheric conditions.

Due to the low pressures associated with RNS operation and the high design pressure of the system, a break of the RNS piping is not credible.

l l

In the event of a leak in the RNS piping outside containment, the operators would be alerted to the leak outside containment by the presence of high radiation alarms in the auxiliary building.

The operators would follow the guidance similar to that provided in Emergency Response Guideline AECA-1.1, LOCA Outside Containment, to identify and isolate the leak by manually isolating the RNS.

Offsite doses resulting from such a leak would be bounded by those for the analyses provided in SSAR Section 15.7, Radioactive Releases from a Subsystem or Component.

The passive safety systems would be available to provide safety-related core cooling and decay heat removal as

required, t

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