ML20205T398
| ML20205T398 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 04/01/1987 |
| From: | Brey H PUBLIC SERVICE CO. OF COLORADO |
| To: | Berkow H NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, P-87118, NUDOCS 8704070282 | |
| Download: ML20205T398 (6) | |
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2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 April 1, 1987 Fort St. Vrain Unit No. 1 P-87118 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555 Attention: Mr. H. N. Berkow, Director Standardization and Special i
Projects Directorate Docket No.
50-267
SUBJECT:
Generic Letter 83-28, Action Item
2.2 REFERENCES
- (Please see Attachment 1)
Dear Mr. Berkow:
This letter is in response to a NRC telephone request on March 9, 1987, that PSC provide a written discussion explaining how Fort St.
Vrain's definition of safety-related equipment meets the intent of the definition as stated in the footnote to Action Item 2.2 of NRC's Generic Letter 83-28 (Reference 1). This letter is supplemental to information submitted in 1983 in Reference 2.
The definition of safety-related in the footnote to Generic Letter 83-28 Item 2.2 reads as follows:
" Safety-related structures, systems, and components are those that are relied upon to remain functional during and following design basis events to ensure: (1) the integrity of the reactor coolant boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, and (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines of 10 CFR Part 100."
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P-87118 Page 2 April 1, 1987 As discussed below, the FSV safety classification system meets the intent of the Generic Letter 83-28 definition of " Safety Related."
FSV SAFETY-RELATED DEFINITION Historically the FSV equivalent of " safety-related" involved the equivalent of the seismic Class I list (now FSAR Table 1.4-1) at the time of the construction permit application in accordance with the AEC requirements at that time. After the construction permit review was underway, tornadoes became of national concern to the AEC, which led to the development of the Safe Shutdown list for use following a Maximum Tornado or Design Basis Earthquake. This is now FSAR Table 1.4-2, " List of Structures, Systems, and Components Required for Safe Shutdown of the Plant," and its use is described in FSAR Section 10.3.9, "Cooldown With Safe Shutdown List Equipment Items Following Design Basis Earthquake or Maximum Tornado." These two lists formed the bases for FSV's safety-related equipment list at the time of issuance of the Operating License.
Subsequently, as a result of reviews following the Brown's Ferry Fire, the " Alternate Cooling Method (ACM)" equipment has been added to the safety-related list.
As a natural result of this evolutionary process, the present FSV definition of " safety-related items" appears as follows in PSC Administrative Procedure G-1 " Definitions":
" Safety-Related Items Those plant systems, structures, equipment, and components which are identified in the FSAR and as detailed and supplemented by applicable P & I, IB and IC diagrams, E and E-1203 schematic diagrams, the Cable Tab and SR-6-2 and SR-6-8 lists to include the following:
a)
Class I per the updated FSAR, Table 1.4-1.
b)
Safe shutdown components per the updated FSAR, Table 1.4-2.
c)
Alternate cooling method (ACM) equipment."
FSV SAFETY-RELATED DEFINITION MEETS INTENT OF GENERIC LETTER DEFINITION A review of the safety-related equipment items of FSAR Tables 1.4-1 and 1.4-2 gives assurance that the intent of the Generic Letter 83-28 definition of " Safety Related" structures, systems, and components is served.
Item 1,
equipment " relied upon to remain functional during and following design basis events to ensure : (1) the integrity of the reactor coolant boundary," is encompassed by:
P-87118 Page 3 April 1, 1987 The PCRV and the steam generators (this includes the safety valves and PCRV liner)
The PCRV closed loop cooling water system and service water piping to heat exchangers for PCRV liner cooling system.
Item 2,
equipment " relied upon to remain functional during and following design basis events to ensure: (2) the capability to shut down the reactor and maintain it in a safe shutdown condition," is equivalent to FSAR Table 1.4-2
" List of Structures, Systems, and Components Required for Safe Shutdown of the Plant."
Item 3,
equipment " relied upon to remain functional during and following design basis events to ensure:
(3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines of 10CFR Part 100,"
is achieved by several items of equipment on the above lists. Examples of equipment for " preventing" accidents from these list are:
The Reactor Building below the refueling floor and the Turbine Building below the operating floor - this protects Safe Shutdown equipment within from accident damage by tornado missiles or building collapse from earthquake.
The core support structure protects against core collapse.
Safe Shutdown list equipment for mitigating accidents include:
The Plant Protective System this system limits and takes corrective action for conditions resulting from accidents and plant control system failure.
The standby diesel generators and batteries - these are for accidental loss of normal electric power sources.
Inlet and outlet secondary coolant system piping from the PCRV up to and including first isolation valves - this is for isolating steam generators to mitigate steam / water inleakage and/or primary coolant outleakage resulting from steam generator leaks or ruptures.
USE OF NON-SAFETY RELATED E0VIPMENT FOR DESIGN BASIS ACCIDENTS Equipment that is non-safety related may be justified for use during and/or following design basis events for one or more of the following reasons:
1.
Due to the very low probability of the accident.
2.
Due to the accident mitigation equipment having to be in normal service at the onset of the accident (if the plant is to operate atall).
P-87118 Pag; 4 April 1, 1987 3.
Due to the accident not affecting the accident mitigation capability of the equipment.
4 The accident mitigation equipment operation capability is a small fraction of the normal inservice capability of the equipment (i.e.,
the accident mitigation capability is demonstrated by the normal operational capability).
5.
The equipment can mitigate consequences of an accident, but its operation is not necessary for the consequences to be acceptable.
The boiler feed pumps are non-safety related equipment items which would be used to supply the water-turbine drives of helium circulator (s) following Design Basis Accident Number 2 (DBA-2). This l
1s the rapid depressurization accident resulting from hypothetical simultaneous rupture of two ASME Code, safety related, seismically qualified, closures in series.
The considerable evidence that the NRC approves of the use of non-safety related boiler feed pumps for this purpose at FSV is reviewed and summarized in Reference 3.
The l
numerous documents referenced therein clearly indicate that the use of non-safety related boiler feed pumps at FSV to provide adequate primary coolant circulation in the event of DBA-2 has been approved by the NRC.
Further, the documentation demonstrates that the NRC took into account that the rapid depressurization accident was not originally a Design Basis Accident, and that DBA-2 requires multiple l
coincident failures and is, therefore, highly unlikely to occur.
Recent information presented in FSAR Section 14.11.1.4 indicates the probability of DBA-2 is less than 1 E-07 per year.
Furthermore, the AEC did not associate depressurization accidents with earthquakes or tornadoes, since the primary coolant system pressure boundary, as l
well as the PCRV penetration secondary closures, are designed to i
preclude rupture due to either the Design Basis Earthquake (DBE) or l
the Maximum Tornado (Reference 4).
The unreasonableness of comrounding the extremely low probability DBA-2 accident with a very I
l low probability natural occurrence (DBE or Maximum Tornado) is a l
basis for the NRC approving non-safety related feed pumps for DBA-2 mitigation. However, boiler feed pump operability is assured through l
the Technical Specifications, as well as being assured by the practical consideration of being operating immediately prior to the I
hypothetical rapid depressurization from full power. The additional l
expense associated with upgrading these pumps to safety-related l
status would not result in a commensurate increase in the protection of the health and safety of the public.
1 l
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l l
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W P-87118 Pag 2 5 April 1, 1987 CONCLUSION It is concluded that FSV equipment safety classification criteria meets the intent of the Generic Letter 83-28 definition of " Safety Related" equipment.that was developed subsequent to the Operating License review of Fort St. Vrain.
If you have any questions on this subject, please contact Mr. M. H.
Holmes at (303) 480-6960.
Very truly yours, h
l H. L. Brey, Manager Nuclear Licensing and Fuels Division HLB /TRM:jnt I
cc: Regional Administrator, Region IV Attention: Mr. J. E. Gagliardo, Chief Reactor Projects Branch Mr. R. E. Farrell Senior Resident Inspector Fort St. Vrain
to P-87118 Page 6 REFERENCES
- 1) USNRC Generic Letter 83-28 " Required Actions Based on Generic Implications of Salem ATWS Events," July 8, 1983
- 2) PSC Letter Lee toEisenhut(NRC),
Subject:
Response to Generic Letter 83-28, November 4, 1983 (P-83359)
- 3) PSC Letter Williams to Berkow (NRC),
Subject:
Use of Non-Safety Grade Equipment.for DBA-2 Cooling, October 8,1986 (P-86406)
- 4) Safety Evaluation by the Division of Reactor Licensing, U.S.A.E.C, in the Matter of Public Service Co. of Colorado, Fort St. Vrain Nuclear Generating Station, Docket No. 50-267, January 20, 1972.