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NUCLEAR REGULATORY COMMISSION
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***** October 11, 1985
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Docket No. 50-267 OCT 2 41985 !
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Mr. 0. R. Lee, Vice President
Electric Production
Public Service Company of Colorado
P. 0. Box 840
Denver, Colorado 80201
Dear Mr. Lee:
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,
etie
Edward J. Butcher, Acting Chief
Operating Reactors Branch #3
Division of Licensing
Enclosure:
As stated
cc w/enclosure:
See next page
851178
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Mr. 0. 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 11th Avenue
of Colorado Denver, Colorado 80220
P. 0. 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. 0. Box 368
P. 0. 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 Commission
P. 0.Box 640
Platteville, Colorado 80651
Kelley, Stansfield & O'Donnell
Public Service Company Building
Room 900
550 15th Street
Denver, Colorado 80202
Regional Administrator, Region IV
U.S. Nuclear Regulatory Commission
Office of Executive Director
for Operations
611 Ryan Plaza Drive, Suite 1000
Arlington, Texas 76011
Chairman, Board of County Commissioners
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 Sr. VkAIN REQUEST FUR SCHEDULE EXIENSION 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-1?
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?
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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
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
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"?
12. In a letter dated August 20, 1985 PSC stated that completion of all major
aspects of EQ program review would be accomplished by September/October
1985 with the exception of steam detection/isolation system
installation. 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 comments pertain to this
mode of operation:
a) While decay heat removal via the LCS is analyzed in the FSV-FSAR as
DBA #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
appears that the request to rely solely on the LCS for decay heat
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.
b) Section 14.10 and Appendix D of the FSV-FSAR indicate that reliance
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
(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 D of the
FSV-FSAR indicates that there are many other plant systems and components
which are assumed to operate in the DBA #1 analysis whose performance
could affect the outcome of this event. These are listed below:
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.
o At least one loop of the Liner Cooling System. This includes:
- Control room actuation of PCRV water flow control
- Control room actuation of LCS 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:
o LCS instrumentation?
o Reactor Plant Cooling Water System?
o He Purification System and LCS valve operations and controls?
o He Purification System He and Charcoal Cooling System?
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
of operation? Is all equipment required to perform in this mode of
decay heat removal periodically tested?
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