ML20210A748

Requests Concurrence to Start Up & Operate Facility Through Graduated Rise to Power Up to 100% of Rated Power,Subj to Listed Constraints. Fort St Vrain 1987 Power Ascension Plan Encl

ML20210A748
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/30/1987
From:	Robert Williams PUBLIC SERVICE CO. OF COLORADO
To:	Berkow H NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20210A753	List: ML20210A748 ML20210A757
References
P-87038, TAC-63576, NUDOCS 8702090005
Download: ML20210A748 (5)

Text

i h Public Service- lch 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 'iO211

=_

R.O. WILUAMS, JR.

Et$Nk"bTioNS January 30, 1987 Fort St. Vrain Unit No. 1 P-87038 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:

Request for Authorization to Operate FSV with Specified Power Restrictions

REFERENCES:

1) PSC Letter, Williams to Berkow, dated January 15, 1987 (P-87002)

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

3) PSC Letter, Walker to Berkow, dated December 10, 1985 (P-85460)

4) NRC Letter, Denton to Walker, dated February 7, 1986 (G-86062)

Dear Mr. Berkow:

The Public Service Company of Colorado (PSC) hereby requests that the Nuclear Regulatory Commission (NRC) provide concurrence to start up and operate Fort St. Vrain (FSV) through a graduated rise to power up to 100% of rated power, subject to the constraints as delineated in this letter.

Acci I 10 8702090005 870130 7 PDR ADOCK 0500 P

. P-87038 Page 2 January 30, 1987 Power Ascension Plan Attached for NRC information and review is a copy of the " Fort St.

Vrain 1987 Power Ascension Plan," including a pictorial representation of plant startup activities as a function of power level. In brief, upon-receipt of permission from the NRC to restart, and upon completion of pre-critical testing including testing of the Steam Line Rupture Detection / Isolation System (SLRDIS), criticality would be achieved. Startup and testing activities in the startup range (0-2% power) would then proceed as described in the Power Ascension Plan. The reactor power would then be raised above 2%

into the Low-Power range to less than 10% reactor power for nuclear dryout, training, testing, and other activities as described in the Power Ascension Plan. Ten percent (10%) rated reactor power is considered to be an appropriate hold point for assessment of PSC's startup activities by NRC Region IV.

Upon receipt of authorization from NRC Region IV to exceed 10%, PSC would continue to raise the power level in an orderly manner through the Low-Power range into the Power range up to, but not to exceed, 35% of rated reactor power. As justified below, 35% power is considered to be an appropriate restriction level for FSV operation under provision of the existing Technical Specifications while awaiting NRR approval and issuance of the Technical Specification change proposed by Reference 1.

Upon receipt of NRR approval, PSC would continue in accordance with the Power Ascension Plan up to, but not to exceed, 82% of rated reactor power. Shutdown cooling has been analyzed to be safe and adequate from power levels above 82% as summarized in the justification below and detailed in References 1 and 2.

The alternatives for supporting a power level above 82% have not been evaluated at this time. Additional analyses will certainly be required and further plant modification may be required to justify power operation up to 100% power. At the present time, PSC considers the 100% operation goal to be a longer term goal (beyond the next refueling). Therefore, when operation at a power level above 82% is pursued, PSC will provide the NRC with the necessary supporting analysis and justification.

Justification for 35% Power Operation PSC has previously requested permission for interim operation with a 35% power restriction during 1986 (Reference 3). This previous request conservatively assumed that in the event of an Environmental Qualification Design Basis Event, all electrical equipment exposed to the resulting harsh environment would fail. Shutdown cooling of the reactor would be accomplished using only electrical equipment located in a mild environment supplying firewater to the PCRV liner cooling system. As a result of this previous request, the NRC authorized interim operation of FSV at 35% of full reactor power through May 31, 1986 (Reference 4). The NRC's Safety Evaluation Report (Enclosure to Reference 4) found that following a High Energy Line Break (HELB) with the firewater system utilized for cooling the

- , P-87038 Pag: 3 January 30, 1987 PCRV liner from power levels of 35% or less, a permanent Loss of Forced Circulation (LOFC) accident can be experienced without significant damage to any of the fission product barriers, including the fuel particle coatings. This evaluation also concluded that re-establishing liner cooling in the 35% power case after it had been shut off for a prolonged period of time was acceptable. (

PSC affirms that there have been no adverse changes to the basic assumptions and analyses transmitted by Reference 4. The most significant change enhances safety. It is that two redundant trains of Safe Shutdown equipment necessary for forced circulation Safe Shutdown Cooling will have been environmentally qualified to the requirements of 10CFR50.49. -

Two other related improvements will have been implemented before criticality is achieved. One improvement is installation of the Steam Line Rupture Detection / Isolation System (SLRDIS). SLRDIS has the capability to automatically and quickly shut off escaping steam or hot water in the event of a HELB within the Turbine or Reactor Building. This would result in substantially lower building temperatures than those presented in Reference 3, thus permitting much earlier operator access to the affected building to perform any f required manual actions. The other beneficial modification pertinent L~

to this discussion is the environmental qualification of the helium circulator brake and shaft seal system controls. This will essentially eliminate primary coolant leakage along the shaft of a stopped circulator following a HELB, resulting in a considerable reduction in the radiological doses 3 (which were already extremely low)previously presented in Reference .

For the purpose of analysis of the above mentioned permanent LOFC accident, no credit is taken for any of the multiple and redundant methods of removing core decay heat utilizing forced circulation of the helium. If all forced circulation is arbitrarily assumed to be permanently unavailable, the previously reviewed permanent LOFC accident analysis described above is considered to be bounding for reactor power levels up to 35%. This analysis was approved by the NRC in Reference 4. For this reason 35% power is considered to be an appropriate rc:triction level for FSV while awaiting NRC's Office of Nuclear Reactor Regulation (NRR) approval and issuance of the Technical Specification change proposed by Reference 1, PSC completion of the modification to install a new six-inch vent line on each main steam loop header, and finalization of identified EQ related issues.

Justification for 82% Power Operation NRR approval of the Technical Specification change request submitted with Reference 1 will be required prior to exceeding the 35%

restriction level, since the existing Technical Specifications require that at least one reheater section be available to support shutdown heat removal, which includes Safe Shutdown Cooling. As discussed in Reference 1, the reheaters have limited capability to support Safe Shutdown Cooling (i.e. supplied with firewater from one firewater pump after a 1-1/2 hour Interruption of Forced

P-87038

. Pag? 4 January 30, 1987 Circulation). PSC must also complete the plant modification to add a six-inch vent pipe to atmosphere from the outlet of each of the two economizer-evaporator-superheater (EES) sections of the steam generator. These vents (one per loop) provide a discharge path to atmosphere for the .once-through cooling mode (Reference 2) and prevent the redundant EES sections from being incapacitated simultaneously by a postulated HELB located in the common discharge piping. While the six-inch vents are required before exceeding 35% i power, the modification is currently expected to be completed prior to criticality.

As described in References 1 and 2, Safe Shutdown Cooling can be satisfactorily performed with one firewater pump supplying one of the two redundant EES sections after a 1-1/2 hour Interruption of Forced Circulation from reactor power levels up to 87.5%. Safe Shutdown Cooling utilizes only equipment on the Safe Shutdown list, all of which are seismically qualified to withstand a Design Basis Earthquake (DBE), and where required, are environmentally qualified to withstand a HELB inside the Turbine or Reactor Building. The analyses performed in regards to fire protection (i.e., in conformance to 10CFR50, Appendix R) confirm the adequacy of the

" Appendix R" shutdown cooling water flow paths from reactor power levels up to 83.2% (Attachment 1 to Reference 2).

Some of the recent concerns regarding the capability of an EES section or a reheater section to adequately support Safe Shutdown Cooling were also raised regarding use of these heat exchangers for cooldown from certain other accidents previously identified in the FSAR. These other accidents (i.e., other than those that result in Safe Shutdown Cooling) have now been reevaluated to determine the effects of the revised secondary coolant analytical models on the accident consequences. Results of these reanalyses assure that the steam generator heat transfer sections are capable of supporting decay heat removal and do not require power level limitations more restrictive than the 83.2% analyzed in the Appendix R case in Attachment 1 to Reference 2. The results of these reanalyses are being submitted separately to the NRC in the immediate future.

Justification for 100% Power Operation As mentioned above, the alternatives for supporting a power level above 82% have not been evaluated at this time. Further analyses will certainly be required and additional plant modification may be required to justify power operation up to 100% power. At the present time, PSC considers the 100% operation goal to be a longer term goal (beyond the next refueling). Therefore, when operation at a power level above 82% is pursued, PSC will provide the NRC with the necessary supporting analysis and justification.

PSC Restart Reauest Summary PSC formally requests that NRC grant permission for FSV to be operated at reactor power levels up to 82% of rated power, subject to the 10% hold point and 35% power restriction level as discussed in this letter.

i

- . P-87038 Page 5 January 30, 1987 PSC hereby comits to an 82% power restriction until such time as NRC Staff approval is secured for operation at a power level above 82%.

PSC understands that the NRC has the option of confirming by order that the 35% power restriction and the 82% power restriction are not to be exceeded without prior NRC approval.

If you have any qtiestions or comments, please call Mr. M. H. Holmes at (303) 480-6960.

Very truly yours, R. O. Williams, Jr.

Vice President, Nuclear Operations R0W/TRM:jmt Attachments cc: J. E. Gagliardo Region IV, NRC R. E. Farrell NRC Senior Resident Inspector, FSV