ML20214U731
| ML20214U731 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 06/01/1987 |
| From: | Robert Williams PUBLIC SERVICE CO. OF COLORADO |
| To: | Calvo J NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| References | |
| P-87202, TAC-54373, NUDOCS 8706110316 | |
| Download: ML20214U731 (11) | |
Text
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2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 R.O. WILLIAMS, JR.
June 1, 1987 N L A PE DONS Fort St. Vrain Unit No. 1 P-87202 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555 Attention: Mr. Jose A. Calvo, Director Projects Directorate IV Docket No.
50-267
SUBJECT:
Request for Appendix R Exemptions Under 10 CFR Part 50.12
REFERENCES:
See Attachment 1 4.
Dear Mr. Calvo:
The Public Service Company of Colorado (PSC) hereby requests an exemption from all of 10 CFR Part 50, Appendix R,
except Sections III.G and III.J. under 10 CFR Part 50.12, as noted on page 9 of of Reference 1.
BACKGROUND PSC has been working with the NRC staff to implement the applicable provisions of 10 CFR Part 50, Appendix R, at Fort St.
Vrain (FSV).
As stated in Reference 15, there is a common understanding between PSC and the NRC staff as to the applicable fire protection regulatory criteria for FSV. PSC has consistently applied its best efforts to implement the fire protection regulatory criteria applicable to FSV.
- However, from time to time problems have arisen in implementing certain fire protection criteria which did not explicitly take the unique features of the FSV gas-cooled reactor into account.
PSC and the NRC staff have been successful in resolving these specific problem areas.
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P-87202 Page 2 June 1, 1987 PSC has never received a comprehensive NRC safety evaluation on its complete 1978 Branch Technical Position BTP-APCSB 9.5-1, Appendix A fire protection program submittals, References 9 and 10, even though the NRC staff in Reference 12 explicitly stated that there were no outstanding fire protection issues on Fort St. Vrain as of 1980.
Without this BTP-APCSB 9.5-1, Appendix A safety evaluation, an inspector could take the position that PSC does not comply with the criteria of 10 CFR 50.48 (b) necessary to be exempted from compliance with the entire 10 CFR 50, Appendix R, except Sections III.G III.J and III.0 (the NRC granted PSC an exemption from Section III.0 of 10 CFR 50 Appendix R in Reference 14).
The requested exemption will officially. establish the regulatory criteria for the FSV High Temperature Gas Cooled Reactor's fire protection program, and promote a clear understanding of that criteria for any future NRC inspection of the FSV fire protection program.
REQUEST FOR EXEMPTION FROM 10 CFR50, APPENDIX R Therefore, PSC requests an exemption from all of 10 CFR Part 50, Appendix R, except Sections III.G and III.J.
PSC also requests that the NRC's exemption approval define the regulatory criteria for Fort St. Vrain's fire protection program to be:
Appendix A to -Branch Technical Position BTP-APCSB 9.5-1, Rev1, except those sections superceded by Sections III.G and III.J of 10 CFR 50, Appendix R PSC letter dated August 17, 1984 (Reference 15)
Sections III.G and III.J of 10 CFR Part 50, Appendix R, with specified exemptions noted in the FSV Appendix R Evaluation Reports AUTHORITY FOR EXEMPTION PSC's application for this exemption from 10 CFR 50, Appendix R conforms to the requirements of 10 CFR 50.12, based on the special circumstances noted below:
Since FSV will comply with a set of regulatory criteria specifically established for this gas-cooled
- reactor, the requested exemption will not present an undue risk to the public health and safety.
P-87202 Page 3 June 1, 1987 10 CFR 50.12(a)(2)(iii)
FSV's compliance with the entire Appendix R at this time would result in an undue hardship, due to the limited financial and manpower resources available at this time, and would not result in a substantial increase in safety margins for public health and safety compared with other similar vintage plants.
The differences between Light Water Reactors and High Temperature Gas-Cooled Reactors were not considered when this regulation adopted.
was Given the design features of FSV High Temperature Gas-Cooled Reactor, PSC submitted, on August 17, 1984 (Reference 15),
specifically tailored set of regulatory guidance to permit FSV a
to meet the fire protection provisions of Section III.G of 10 CFR 50, Appendix R and the intent of the criteria in Section III.L of 10 CFR 50 Appendix R, as they apply to FSV.
The existence of a evaluation reports on FSV " Appendix A" fire protectionsubstantial number of
- features, with only minor review left for a comprehensive safety evaluation, is a special circumstance, which was also not considered when this regulation was adopted.
BASIS FOR EXEMPTION 10 C FP.- Part 50.48 (b) defines the applicability of 10 CFR Part 50, Appendix R as:
"Except for the requirements of Sections III.G, III.J and III.0, the provisions of Appendix R to this part shall not be applicable to nuclear power plants licensed to operate prior to January 1,1979, to the extent that fire protection features proposed or implemented by the licensee have been accepted by the NRC as satisfying the provisions of Appendix A
to Branch Technical Position BTP APCSB 9.5-1 reflected in staff fire protection safety evaluation reports issued prior to the effective date of this rule, or to the extent that fire protection features were accepted by the staff in comprehensive safety evaluation reports issued before Appendix A to Branch Technical Postion BTP APCSB 9.5-1 was published in August, 1979."
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P-87202 Page 4 June 1, 1987 FSC reviewed the existing correspondence with the NRC on fire protection. This shows the history of the fire protection issues and the intent of both parties.
Appropriate quotations are noted in.
The NRC has issued various fire protection safety evaluation reports for Fort St. Vrain, including:
SUBJECT SER DATE REF
-Fire protection actions June 18, 1976 2
Three Room Control Complex Feb. 10, 1978 4
Alternate Cooling Method Feb. 10, 1978 4
Watcr Spray Protection System July 13, 1978 6
IActi/ACM cable routing June 6, 1979 11 The last four of these SERs evaluated their subject areas under Appendix A to BTP 9.5-1 Rev.
I criteria.
The NRC has noted in Reference 7 "that the remaining fire protection review required of FSV is of limited scope" due to the various SERs previously issued.
The NRC has never issued a comprehensive evaluation of PSC's completed " Appendix A" fire protection program as submitted to the NRC on October 12, 1978 (Ref 9) and November 12 1978 (Ref. 10).
10 CFR Part 50, Appendix R became effective February 17, 1981, but according to the NRC in Reference 12, was not applicable to Fort St.
Vrain (except for Sections III.G, III.J, and III.0), since there were
.r "no outstanding fire protection issues on Fort St. Vrain."
Although References 7 and 12, indicated that there were only minimal or no disputed fire protection issues outstanding on FSV, the NRC requested PSC to compare its fire protection program to the 10 CFR Part 50, Appendix R criteria on October 6,1981 (Reference 13).
The NRC explained in Reference 13 that:
"The NRC staff had intended in its original proposal for Appendix R that the requirements be applicable only for the resolution of unresolved disputed fire protection features."
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P-87202 Page 5 June 1, 1987 FSV's fire protection program meets the NRC's intent stated in Ref.
13 for granting relief from this regulation.
The NCC has issued various SERs on FSV " Appendix A" fire protection matters and has indicated that there were no significant outstanding disputed issues requiring the application of the provisions of Appendix R for resolution (Reference 12).
On June 4,
1984, the NRC noted that Appendix R, Sections III.L and III.0 did not apply to FSV (Reference 14).
The NRC stated that PSC's FSV Appendix R fire protection program need not comply with the requirements of 10 CFR Part 50, Appendix R, other than Sections III.G and III.J. As a result of this direction, PSC submitted their Appendix R Evaluation comparison based only on Appendix R, Sections III.G and III.J.
As can be seen from the information contained herein, it appears that the NRC staff and PSC mutually understand the applicability of 10 CFR Part 50, Appendix R to the licensing basis of Fort St. Vrain.
However, to remove all doubt on this issue, PSC requests the exemption contained herein.
If you have any questions concerning this matter, please contact Mr.
M. H. Holmes at (303) 480-6960.
Very truly yours,
!/
R. O. Williams, Jr.
Vice President, Nuclear Operations R0W/ RAS:bac Attachments l
l 1.
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P-87202 Page 6 June 1, 1987 cc: Regional Administrator, Region IV Attn: Mr. J. E. Gagliardo, Chief Reactor Projects Branch Mr. R. E. Farrell Senior Resident Inspector Fort St. Vrain
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Attachment l' to P-87202 Page 1 REFERENCES 1.
PSC Letter, Williams to Calvo, dated January 16, 1987 (P-87013) 2.
NRC Letter, Denise to Walker, dated June 18, 1976 (G-76046) 3.
NRC Letter, Denise to Fuller, dated October 28, 1977 (G-77076) 4.
NRC Letter, Denise to Fuller, dated February 10, 1978 (G-78010) 5.
NRC Letter, Denise to Fuller, dated March 31, 1978 (G-78029) 6.
NRC Letter, Speis to Fuller, dated July 13, 1978 (G-78074) 7.
NRC Letter, Speis to Fuller, dated September 22, 1978 (G-78096) 8.
PSC Letter, Fuller to Ganmill, dated September 28, 1978 (P-78158) 9.
PSC Letter, Fuller to Gammill, dated October 13,1978(P-78167)
- 10. PSC Letter, Fuller to Gammill, dated November 13, 1978 (P-78182)
- 11. NRC Letter, Gammil to Millen, dated June 6,1979 (G-79103)
- 12. NRC Letter, Eisenhut to All Power Reactor Licensees with Plants Licensed Prior to January 1,
1979, dated November 24, 1980 (G-80208)
- 13. NRC Letter, Eisenhut to Lee, dated October 6, 1981 (G-81181)
- 14. NRC Letter, Wagner to Lee, dated June 4,1984 (G-84176)
- 15. PSC Letter, Lee to Johnson, dated August 17, 1984 (P-84281)
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t to P-87202 Page 1 4
CORRESPONDENCE HISTORY A review of the fire protection correspondence shows the NRC intent 4
and direction concerning the application of Appendix R to FSV.
It also shows that PSC exercised due diligence in addressing the NRC's fire protection requirements.
The following applicable quotations are included herein to reflect the history of this issue:
On June'18, 1976, the NRC wrote in their Safety Evaluation (Reference 2; Page 48):
" Based on the fire protection action proposed by PSCo and discussed previously in this report, we conclude that the overall fire protection plan described by PSCo is acceptable.
We conclude that the fire protection improvement action to be taken for Stage 1, in conjunction with the other action to be taken for Stage 1 operation, provides acceptable defense-in-depth for protection against fires during this relatively short period of operation."
On October 28,
- 1977, the NRC wrote in their Safety Evaluation j
(Reference 3; page 4-5):
"We conclude that the fire protection requirements Stage 2 operation have been satisfactorily completed.
Satisfactory equipment installations, test, technical specifications, and i
operating procedures are completed for the upgraded fire protection systems in accordance with the criteria of the June 1976 safety evaluation report.
With respect to fire protection, the plant could proceed to 100 percent of rated power during Stage 2 operations.
Stage 3 operation will l
require satisfying the consnitments for fire protection in the June 1976 staff report and the development of an acceptable plan for substantial confonnance with other NRC fire protection guidelines, which may include Appendix A to NRC Branch Technical Position 9.5-1.
The licensee has proposed a design for meeting Stage 3 requirements which is currently beir.g reviewed."
On February 10, 1978, the NRC wrote (Reference 4):
"During the past eighteen months, we have reviewed the fire protection system for the Fort St. Vrain three room control complex and the interim version of the alternate cooling method and have found both acceptable for Stage 2
operation."
to P-87202 Page 2 On March 31, 1978, the NRC wrote (Reference 5):
".. we find that the proposed floor penetration seal design using Dow Corning silicone foam material is an acceptable three hour fire barrier to be used in the Fort St. Vrain three-room complex."
On July 13,1978, the NRC wrote (Reference 6):
" Based on our review, we conclude that the water spray fire protection system, in conjunction with the previously accepted fire protection measures discussed above, are acceptable at Fort St. Vrain for Stage 3 operation."
On September 22, 1978, the NRC advised PSC (Reference 7) to submit its comparison of the Fort St.
Vrain fire protection provisions against the applicable guidelines of Appendix A to Branch Technical Position 9.5-1 and a fire hazards analysis for the plant by October 15, 1978, to enable the NRC staff to complete its review by December, 1978.
In this letter, the NRC noted:
"that the remaining fire protection review required of FSV is of limited scope since the fire protection for the three room complex and adjacent congested electrical cable areas has already been evaluated and accepted by the staff" and "an. alternate cooling system,... has been previously found acceptable by the staff, has been installed..."
On September 28, 1978, PSC comitted to submit this requested comparison by October 15, 1978, and the fire hazards analysis 30 days later as requested by the NRC (Reference 8).
t On October 13, 1978, PSC submitted (Reference 9),
its " Fire
'l Protection Program Review for Fort St.
Vrain Nuclear Generating Station in Response to Branch Technical Position 9.5-1,"
as requested.
On November 13, 1978, PSC submitted (Reference 10 Section 4.0) its
" Fire Protection Program Review for Fort St. Vrain Nuclear Generating l
Station in lleponse to Branch Technical Position 9.5-1,"
which i
contained the Fire Hazards Analysis, as promised.
to P-87202 Page 3 1
On June 6,
- 1979, the NRC wrote in their Safety Evaluation Report (Reference 11; page 16):
i "We have reviewed the design and implementation of the IACM/ACM conversion cable routing. The requirement to be satisfied was that no single event would simultaneously result in failure of the ACM and the primary systems for cooling down the plant. We have concluded from our review that the cable routing through an independent and separate duct bank between the new diesel generator power supply and the equipment necessary to cool down the plant satisfies the above requirement and is therefore acceptable.
"In addition, we have reviewed the design, installation and operational requirements of the Alternate Cooling Method and conclude that the ACM can provide all necessary functions to assure safe plant shutdown and emergency cooling under the degraded conditions caused by a large electrical fault or fire in the electrical system."
On November 24, 1980, the NRC transmitted a revised 10 CFR 50.48 and a new Appendix R to all Reactor Power Licensees with Plants Licensed prior to January 1, 1979 (Reference 12).
This letter set forth the requirement that all plants had to meet Sections III.G.
III.J, and III.0 in their entirety.
It also stated:
"The second category of Appendix R provisions applicable to the fire protection features of your facility consists of requirements concerning "open" items of previous NRC staff fire protection reviews of your facility. An open item is t
defined as a fire protection feature that has not been previously approved by the NRC staff as satisfying the provisions of Appendix A to Branch Technical Postion BTP PCSB (sic) 9.5-1, as reflected in a staff fire protection
)
safety evaluation report.
The fire protection features of your facility that are in this category must satisfy the specific requirements of Appendix R by the dates established i
l by Paragraph 50.48(c), unless an exemption from the Appendix R
requirements on those features is approved by the Comission.
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to P-87202 Page 4
" Enclosure 2 is a sunnary list of the open items concerning the fire protection features of your facility based on a review of our records." noted:
"No outstanding fire protection issues on Fort St. Vrain."
On October 6,1981, the NRC wrote (Reference 13):
"The technical requirements of Appendix R to 10 CFR Part 50 are now being used as guidelines in our evaluation of the fire protection programs for plants under review for operating licenses.
"It has been our recent practice to perform the fire protection reviews for OL plants using the provisions of Appendix R.
Accordingly, as a part of your overall fire protection program submittal, we request that you include a
comparison of your fire protection program to Apperdix R to 10 CFR Part 50.
Specifically identify and justify any deviation from Appendix R."
to Reference 13 is the Federal Register publication of the " Fire Protection Program for Operating Nuclear Power Plants" final rule.
On page 76602 of Federal Register Vol. 45, No. 225 of Wednesday, November 19, 1980, it states:
"The NRC staff had intended in its original proposal for Appendix R that the requirements be applicable only for the resolution of unresolved disputed fire protection features."
On June 4,
- 1984, the NRC (Reference 14):
wrote of FSV's Appendix R requirements "We agree that Section III.L does not apply to FSV, however,Section III.G does apply...Section III.0 also does not apply to FSV since the plant does not have reactor coolant pumps or a lubricating oil system for helium circulators.
Thereforms no exemptions are needed for Sections III.L and III.O."
.