ML20210T655

Provides Results of Confirmatory Analyses for FSAR Accidents Which Utilize Either EES or Reheater Section of Steam Generator for DHR

ML20210T655
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/06/1987
From:	Brey H PUBLIC SERVICE CO. OF COLORADO
To:	Berkow H NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20210T661	List: ML20210T655 ML20210T686 ML20210T965 ML20266B243 ML20266B257 ML20266B267 ML20266B271
References
P-87053, TAC-63576, TAC-66574, NUDOCS 8702180266
Download: ML20210T655 (3)

Text

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2420 W. 26th Avenue, Suite 1000, Denver, ralorado 80211 February 6, 1987 Fort St. Vrain Unit No. 1 P-87053 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.

20555 Attention: Mr. H. N. Berkow, Director Standardization and Special Projects Directorate Docket No.

50-267

SUBJECT:

Confirmatory Analyses for Reactor Cooldown from 83.2%

Power for Various FSAR Accidents

REFERENCE:

1) PSC Letter, Warembourg to Berkow, dated December 30, 1986 (P-86683)

2) LER 86-026, dated August 17, 1986 (P-86587)

3) LER 86-020, dated August 11, 1986 (P-86513)

4) PSC Letter, Warembourg to Berkow, dated December 30, 1986 (P-86682)

Dear Mr. Berkow:

The purpose of this letter is to provide the results of the confirmatory analyses for FSAR accidents, not previously reevaluated in Reference 1, which utilize either an EES or a reheater section of a steam generator for decay heat removal.

The need for the confirmatory analyses was identified as a result of References 2 through 4, which identified limitations of the cooling capabilities of the steam generator EES and reheater sections.

PSC committed to reanalyze other FSAR accidents which rely on EES or reheater sections for decay heat removal to confirm the adequacy of the systems and I

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P-87053 Paga 2 February 6, 1987 flow paths utilized for shutdown cooling and to verify that no additional restrictions exist which further limit power beyond that previously identified in Reference 1 (83.2% power). (EE-22-0007) analyzes FSAR accidents which utilize steam generator EES sections for decay heat removal (see Tables IA and 18 of Attachment 1).

The results of these analyses confirm that the applicable cooling modes utilized will provide adequate decay heat removal for reactor cooldown following unlimited reactor operation at power levels not to exceed 83.2% power.

Fuel temperatures remain below the FSAR fuel temperature limit of 2900 degrees F in all cases. (GA 909258) analyzes FSAR accident conditions of concern which utilize a steam generator reheater section for decay heat removal.

The specific accident analyzed is the FSAR steam generator tube leak accident (as discussed in FSAR Section 14.5.3, Cases 2 and 5).

This accident involves a steam / water ingress event resulting from a steam genere tor EES tube or subheader rupture, followed by malfunctions of the Plant Protective System (PPS) resulting in automatic isolation and dumping of the wrong (intact) loop. The FSAR postulates that the loop with the intact steam generator is isolated, the contents of its EES section are dumped to the steam / water dump tank and restoration of cooling with the intact loop cannot be achieved. Therefore cooldown must proceed using the intact reheater in the leaking steam generator.

This represents a highly improbable set of circumstances in that it is postulated that multiple failures of the safety related PPS, as stated in FSAR Section 14.5.3.2, occur resulting in isolation of the intact loop.

It is also postulated, given these PPS failures, that the loop containing the intact EES section cannot be recovered.

The results of the Attachment 2 analysis confirm the adequacy of this postulated cooling mode in which the reheater of one loop supplied with condensate and two helium circulators in the same loop driven by feedwater provide sufficient heat removal capability following shutdown from 100% power. The analysis also confirms the adequac cooldown utilizing feedwater to the intact (isolated and dumped) y of EES with feedwater driving two circulators in the associated loop. The latter.is the preferred cooldown method following a steam / water ingress with wrong loop dump.

The analyses for both of the above described cooling modes are based on an initial 30-minute cooldown with feedwater continuing to supply the leaking EES and steam continuing to drive two circulators in the associated loop. This is the basis for the FSAR analysis of this accident.

PSC considers the results of these confirmatory analyses for the steam generator EES and reheater sections to be sufficient to demonstrate the capability of the steam generator EES and reheater sections to provide adequate core cooling for FSAR accidents other

P-87053 Page 3 Fcbruary 6, 1987 than those evaluated in Reference 1 (1'.e., Safe Shutdown Cooling and Appendix R Fire Protection Cooldown) from 83.2% power.

If you have any questions, please contact Mr. M. H. Holmes at (303) 480-6960.

Very truly yours, We H. L. Brey, Manager Nuclear Licensing and Fuels Divsion HLB /CB:jmt Attachments