ML20138E392
| ML20138E392 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 10/11/1985 |
| From: | Butcher E Office of Nuclear Reactor Regulation |
| To: | Lee O PUBLIC SERVICE CO. OF COLORADO |
| References | |
| TAC-59787, NUDOCS 8510250011 | |
| Download: ML20138E392 (6) | |
Text
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October 11, 1985 DISTRIBUTION:
DOCKET FILE ACRS 10 Docket No. 50-267 NRC PDR Gray File L PDR HBerkow ORB # 3 Rdg PWagner HThompson Mr. O. R. Lee, Vice President 0 ELD Electric Production EJordan Public Service Company of Colorado BGrimes P. O. Box 840 JPartlow Denver, Colorado 80201 KHeitner PMKreutzer
Dear Mr. Lee:
EJButcher
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON SCHEDULE EXTENSION, 10 CFR 50.49 We are reviewing your September 24, 1985 request for special consideration and a schedule extension to the equipment qualification rule,10 CFR 50.49.
Enclosed is a request for additional information we need to complete this review. We request that you provide this information to us within 7 days of your receipt of this letter so that we can complete our review on a schedule compatible with the deadlines in the rule.
The information requested in this letter affects fewer than 10 respondents; therefore OMB clearance is not required under P.L.96-511.
Sincerely,
/s/
Edward J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
See next page
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Mr. O. R. Lee Public Service Company of Colorado Fort St. Vrain cc:
C. K. Millen Albert J. Hazle, Director Senior Vice President Radiation Control Division Public Service Company 4210 East lith Avenue of Colorado Denver, Colorado 80220 P. O. Box 840 Denver, Colorado 80201 J. W. Gahm Nuclear Production Manager Mr. David Alberstein,14/159A Public Service Company of Colorado GA Technologies, Inc.
P. O. Box 368 P. O. Box 840 Platteville, Colorado 80651 Denver, Colorado 80201 J. K. Fuller, Vice President Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201 Senior Resident inspector U.S. Nuclear Regulatory Comission P. 0. Box 640 Platteville, Colorado 80651 Kelley, Stansfield & 0'Donnell Public Service Company Building Room 900 550 15th Street Denver, Colorado 80202 Regional Administrator, Region IV U.S. Nuclear Regulatory Comission Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Comissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street Denver, Colorado 80651
REQUEST FOR ADDITIONAL INFORMATION FORT ST. VRAIN REQUESI FOR SCHEDULE EXItNSION ON 10 CFR 50.49 1.
What is the capability of the reactor building confinement and louvers to withstand a steamline break and still function effectively to reduce the radiological consequences of DBA-17 2.
What is the effect of the moisture released by the break on the effectiveness of the reactor building exhaust filters?
3.
What are the safety benefits from operating the plant at partial power during the period November 30, 1985 to March 31, 1986?
4 How many equipment items and total components will be qualified as of November 30, 1985?
How many additional equipment items and total components will be qualified by March 31, 1986?
5.
What actions would be required to return Fort St. Vrain to power operation from a permanent loss of forced circulation cooling at partial power?
6.
Identify and discuss the reasons for the equipment used at Fort St.
Vrain being considerably different from the equipment used in Light Water Reactors (e.g.; transmitters, valve operators, temp sensors, etc.).
7.
Allowing 4 minutes for leak termination, the accident environmental profiles for Fort St. Vrain are not exceptionally higher than those seen in PWR containments (400 F - 500*F). What is the basis for stating that equipment qualification information from LWR industry is not available or applicable to Fort St. Vrain?
8.
How does the acceptability or unacceptability of the 4-minute isolation time impact the following specific deficiencies in the Fort St. Vrain EQ program?
Aging effects on equipment?
Post-accident operability time?
Field walkdown for equipment verification?
Identification of the equipment to be included in the qualification program?
Review and evaluation of the completeness of EQ documentation and files?
9.
Does the installation of an automatic leak detection / isolation system adversely impact any aspect of the present assumptions of the EQ program regarding resulting accident environment? (i.e., do the resulting l.
accident environmental conditions become more severe than presently estimated?)
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- 10. Provide sufficient detail for the NRC to evaluate the adequacy of the proposed leak detection / isolation system or the schedule for its submittal. Provide assurance that the system will be fully environmentally qualified at time of installation.
- 11. Redundant, safe shutdown, forced circulation cooling water systems using fire water were considered qualified prior to early 1985. Assuming a i
4-minute isolation time is acceptable, identify the requirements that were modified that led the PSC to conclude that now "not all of this equipment can be considered fully qualfied to the requirements of 10 CFR 50.49"?
l 12.
In a letter dated August 20, 1985 PSC stated that completion of all major l
aspects of EQ program review would be accomplished by September / October 1985 with the exception of steam detection / isolation system f r.stallation. The schedule is now being revised to March 1986. What assurance is there that the new schedule is realistic?
- 13. The request for a schedule extension for meeting the requirements of 10 CFR 50.49 is based, in part, upon relying solely on the FSV-PCRV Liner Cooling System (LCS) for decay heat removal (using fire water as the cooling medium) until such time as equipment qualification concerns are resolved on those systems and components used for forced He circulation decay heat removal. The following coments pertain to this mode of operation:
a)
While decay heat removal via the LCS is analyzed in the FSV-FSAR as D3A #1, the licensing basis of the plant assumes several lines of defense (steam drive and water drive for the He circulators) prior to reaching a LCS only decay heat removal situation. Therefore, it 1
i appears that the request to rely solely on the LCS for decay heat i
removal assumes these lines of defense are not necessary, which would appear to be beyond the licensing basis of the plant and not in accordance with Technical Specification requirements. Justification to support such operation needs to be provided.
l b)
Section 14.10 and Appendix D of the FSV-FSAR indicate that reliance I
on the LCS alone for decay heat removal could lead to plant damage and fuel failures at decay heat levels associated with high power operation. Therefore, the potential for plant damage, fuel failure and fission product release should be discussed as part of the extension request.
In addition, the extent of plant cleanup or repair after such an event should also be discussed.
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c)
The FSAR presents only an analysis of LCS decay heat removal using the Reactor Plant Cooling Water System; however, the PSC 10 CFR 50.49 extension request is based upon LCS decay heat removal using fire water cooling. Analysis of LCS performance 4
(temperatures, adequacy of cooling water inventory, etc.) should be provided to demonstrate that operation of the LCS in the fire water mode is within the bounds of the FSAR-DBA #1 analysis.
14 The 10 CFR 50.49 extension request indicates that only the LCS with fire water cooling and a Steam Line Rupture Detection / Isolation System need be qualified to support continued operation of the plant and stay within the bounds of the FSV-FSAR DBA #1 analysis. Our review of the DBA #1 analysis contained in Section 14.10 and Appendix 0 of the FSV-FSAR indicates that there are many other plant systems and components 4
i which are assumed to operate in the DBA #1 analysis whose perfonnance could affect the outcome of this event. These are listed below:
l Equipment assumed operating in the FSV DBA #1 analysis:
o At least one He purification system train for depressurization.
This includes:
He and charcoal cooling systems.
o Reactor Building ventilation system. This includes:
One air handling unit Moisture separator Two out of three HEPA and charcoal filters Filter exhaust fans Particulate, I and noble gas exhaust monitors.
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o At least one loop of the Liner Cooling System.
This includes:
l Control room actuation of PCRV water flow control l
Control room actuation of LC5 cover gas pressure increase.
o Reactor plant cooling water system.
o Reserve shutdown system. This includes:
Control and pressurization systems.
The status of the qualification of all the above systems and components under steam line or feedwater line break conditions should be described.
This also includes their attendant AC power, DC power, instrument air supply, instrumentation, cooling, etc.
For any of the above which are not qualified, the impact on the DBA #1 analysis and conclusions should be described.
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- 15. Does the Alternate Cooling Method diesel generator provide power to:
I o
LCS instrumentation?
o Reactor Plant Cooling Water System?
t o
He Purification System and LCS valve operations and controls?
i o
He Purification System He and Charcoal Cooling System?
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- 16. Are there procedures in place to support decay heat removal with the LCS on fire water cooling? Have the operators been trained in this mode l
of operation? Is all equipment required to perform in this mode of j
decay heat removal periodically tested?
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