ML20207G599
| ML20207G599 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/12/1988 |
| From: | Miltenberger S Public Service Enterprise Group |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NLR-N88128, NUDOCS 8808240172 | |
| Download: ML20207G599 (4) | |
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Pubhc Service Elec.ric and Gas Cornpany Steven E. Mittenberger Pubhc Service Electric and Gas Company P.O. Box 236. Hancocks Bridge. NJ 08038 609-339 4199 v ce nescem a,J cw wc' ear oncer August 12, 1988 NLR-N88128 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REQUEST FOR AMENDMENT - SUPPLEMENTAL INFORMATION FACILITY OPERATING LICENSE NPF-57 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Public Service Electric and Gas Company (PSE&G) requested a change to Technical Specification Table 3.3.3-3 in an August 13, 1987 transmittal from C. A.
McNeill, Jr. which would increase the response time for the High Pressure Coolant Injection (HPCI) system f rom 27 seconds to 35 seconds.
Several discussions with the NRC staff have resulted in a verbal request to provide additional information which further quantifies the impact of the proposed change. contains the necessary j
supplemental information which shoul.d address the staff concerns.
Should you have any additional questions or comments, please do not hesitate to contact us.
Sincerely,
,{
Attachment (90{
8808240172 880812 PDR ADOCK 05000354 g (
P PDC
Document Control Desk 2
8/12/88 C
Mr.
G.
W. Rivenbark USNRC Licensing Project Manager Mr. G. W. Meyer USNRC Senior Resident Inspector Mr. W. T.
Russell, Administrator USNRC Region I Mr.
D. M. Scott, Chief Bureau of Nuclear Engineering Department of Environmental Protection 380 Scotch Road Trenton, NJ 08628
In response to verbal NRC staff questions, the following additional information supplements the request for amendment, submitted on August 13, 1987, which proposed increasing the High Pressure Collant Injection (HPCI) system response time from 27 seconds to 35 seconds.
Of specific concern was the impact of the proposed change on the limiting peak clad temperatures (PCTs) calculated in the performance evaluation for the Emergency Core Cooling Systems (ECCS) during a Loss of Coolant Accident (LOCA)
(UFSAR Section 6.3), and the decrease in reactor vessel water level during Feedwater (FW) transients (UFSAR Section 15.6).
UFSAR Section 6.3 discusses the Emergency Core Cooling Systems, including the HPCI system, and eveluates the performance of these systems across the entire break spectrum.
UFSAR Table 6.3-6 summarizes the types of single failure cases considered in the ECCS performance evaluation.
As discussed in UFSAR Section 6.3.1.1.2, the worst caso single failure is an assumed loss of the Channel A de source (which results in the HPCI system being unavailable because the HPCI system injection path requires operation of de-powered, Channel A, solenoid-operated valves) since all other single failure cases would result in tne availability of more ECCS equipment.
As indicated in UFSAR Sections 6.3.3 and 6.3.3.3, this scencrio is the worst caso scenerio across the entire break spectrum, from a double-ended recirculation pump suction break to an instrument line break and all intermediate-sized breaks between these two cases.
The results of the ECCS performance evaluation (UFSAR Table 6.3-3 and UFSAR Figure 6.3-14) represent the effect of a worst case single failure, a channel A de power source failure, in conjunction with any pipe break across the entire break spectrum.
Increasing the time delay for the HPCI system does not affect the results of the limiting LOCA analysis over the entire break spectrum presented in UFSAR Section 6.3 be use the analysis assumes the HPCI system never operates.
TL Channel A de source failure, the worst case single failure as du, cussed above, eliminates the HPCI system f rom the accide:.t analysis since the injection path requires operation of dc-rowered, Channel A solenoid-operated valves.
With regard *.o non-limiting single failures, the two single failure cases identified in UFSAR Table 6.3-6 which take credit for the HPCI system all have additional ECCS equipment available to mitigato the consequences of the accident.
The only difference in t.he standby dicsol generator (SDG) single failure from the Channel A de source failure is the availability of the HPCI system.
Thus, if an infi"ite response time is assumed for the HPCI system, the ECCS performance results would be the same as those presented for the Channel A de source failure.
The LPCI single failure results in the availability of an additional core spray loop than that available in the Channel A de source failure, even with an assumed loss of or infinite time delay in the HPCI system.
Thus a delay in the HPCI response Page 1 of 2
time would not create the potential for a non-limiting single failure to become a limiting case.
Hence, a change in the HPCI system response time does not change the fact that the Channel A de power source failure is still the limiting accident scenerio over the entire break spectrum and the PCT results presented-in UFSAR Section 6.3 bound the proposed change and their revision is not necessary.
The second accident evaluation discussed in the amendment request is the loss of feedwater (FW) transient.
As discussed in UFSAR Section 15.6.6.2.3, a FW break outside primary containment is a special case of the LOCA break spectrum evaluated in UFSAR Section 6.3 since the break is isolable.
The single failure of 1
either the HPCI or the Reactor Core Isolation Cooling (RCIC) system would not prevent sufficient flow from maintaining the-core covered thus assuring adequate core cooling and preventing 4
fuel rod cladding failure.
As a result, UFSAR Section 15.6.6.3.1 concludes that the FW accident scenerio is less limiting than the cases evaluated in UFSAR Section 6.3 since the accident scenerio evaluated in UFSAR Section 6.3 results is so:".e fuel uncovery.
To evaluate the impact of the proposed change in HPCI response time on this conclusion, PSE&G has calculated that an 8 second delay in the 5600 gpm, HPCI system results in a water inventory reduction of approximately 750 gallons.
This reduction in inventory results in a 3 to 4 inch drop in reactor vessel water level which is still above Level 1 (32.2 inches above the top of the active fuel.)
Since the fuel is maintained covered, the i
conclusions reached in Section 15.6.6.3.1 are still valid.
In conclusion, PSE&G has reviewed the proposed change against the limiting and non-limiting accident transients across the entire break spectrum and has reached the same conclusion as contained in the referenced amendment request.
Namely, increasing the HPCI response time is not a significant hazards consideration since the accident scenerios which utilize the HPCI systems (1) have never been limiting events, and (2) do not pose the possibility
^
of becoming limiting events as a result of the change.
In addition, the proposed change does not affect the limiting PCT values contained in UFSAR Section 6.3 nor reduce the reactor i
vessel water level below Level 1 during a FW transiont.
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