ML20006E276
| ML20006E276 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 02/09/1990 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20006E269 | List: |
| References | |
| NUDOCS 9002220495 | |
| Download: ML20006E276 (10) | |
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t' PROPOSED TECHNICAL SPECIFICATION '
CHANGES REGARDING RHR PUMP OPERABluTY
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' 3.5 (cont'd) 4.5 (cont'd)
- 2. From and after the date that one of the Core Spray 2. When it is dets n 6 sed that ore Core Spray System is Systems is made or found inoperable for any reason, inoperable, the operable Core Spray System, and both continued reactor operation is permissible during the LPCI subsystems, shall be venfied to be operable succeeding 7 days unless the system is made operable immediately. The isnaic6 g Core Spray System shall be earlier, provKled that during the 7 days all active _
venfied to be operable daily thereafter.
components of the other Core Spray System and the LPCl System shah be operaNe.
- 3. LPCI System testing shall be as specified in 45.A.1a, b, c,
- 3. Both LPCI subsystems of the RHR System shall be d, f and g except that each RHR pump shall deliver at least operable whenever irradiated fuel is in the reactor and prior 9,900 gpm against a system head cwwdrg to a to reactor startup from a cold condition, except as reactor vessel to pnmary containment differential pressure i specified below. of greater than or equal to 20 psid.t - l l a. From the time that one of the LPCI subsystems is a. When it is determined that one LPCI subsystem is L made or found to be inoperable for any reason, inoperable, the operable LPCI subsystem and both _
continued reactor operation is permissible dunng the Core Spray Systems shall be venfied to be operable succeeding 7 days unless that subsystem is made immediately and daily thereafter.
operable earlier provided that during these 7 days the operable LPCI subsystem and both Core Spray Systems shall be operable.* I l
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i tFor the remainder of Cycle 9, RHR pumps *N and
- LPCI subsystem "N may be inoperable for a 14 "C" shall each deliver at least 8,910 gpm against a l day period. This temporcry LCO exists until the system head correspordig to a reactor vessel to end of Cycle 9.
primary containment differential pressure of greater g ,
than or equal to 20 psid.
l Amendment No. J4, Q(1%1g j 114
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3.5 (cont'd) ' 4.5 (cont'd)
- 3. Should one RHR pump and/or one RHRSW pump of the 3. ' When one contamment cookng subsystem loop becomes components required in 3.5.B.1 above be made or found inoperable, the operable loop shall be verified to' be inoperable, continued reactor operation is perrrussible only operable immediately and daily thereafter.-
- during the n% 30 days provided that dunng such 30 . days all' remammg actwo components .of ' the containment coolmg mode are operable.
- 4. . Should one of the contamment coohng subsystems become inoperable, contmoed reactor' . operation is permissable for a penod not to exceed 7 days, unless such subsystem is sooner made operable provided that during
, . such 7 days all achve ccirycneits of the other
] containment cooling subsystem are operable. *
- 5. if the requirements of 3.5.8 cannot be met, the reactor -
shall be placed in a co?d condition within 24 hr.
- 6. Low power physics testing and reactor operator training shall be permitted with ' reactor coolant temperature
<212 F with an inoperable component (s) as specified in 3.5.B above.
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l Containment Cooling subsystem "A" may be inoperable for a 14 day period. This temporary LCO condition exists until the end of Cycle 9.
l Amendment No. X9tfM6
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_ , s& ; SAFETY EVALUATION FOR ' .
4 PROPoYE5TEURRICETPEUiPTDATION
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a 0 2 New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT i Docket No. 50-333 l l
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SAFETY EVALUATION
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Section 1 - DESCRIPTION OF THE PROPOSED CHANGES -
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-The proposed. changes to the James A. FitzPatrick Technical Specifications revise Specifications 3.5.A.3.a. 4.5.A.3.a, on page 114, and 3.5.B.4 on page 116. Specifically, notes are added as follows:
1 Specification 3.5.A.3.a Page 114 Add an asterisk (*) at the end of this specification and add a note to read:
- LPCI subsystem "A" may be Inoperable for a 14 day period. This temporary LCO condition exists until the end of Cycle 9.
k Specification 4.5.A.3.a Page 114 Add a dagger (t) at the end of this specification and add a note to read: 1 tFor the remainder of Cycle 9, RHR pumps "A" and "C" shall each deliver at least 8,910 gpm against a system head corresponding to a ;
reactor vessel to primary containment differential pressure of greater ;
than or equal to 20 psid. l 1
i Specification 3.5.B.4 Page 116 Add an asterisk (*) at the end of this specification and add a note to read:
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- Containment Cooling subsystem "A" may be inoperable for a 14 day period.- This -1 temporary LCO condition exists until the end of Cycle 9. .,
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'Sectlon 11 PURPOSE OF THE PROPOSED CHANGES
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.-1 The FitzPatrick plant has four Residual Ept Removal (RHR) pumps, with RHR pumps "A" '
. and "C" comprising one Low Pressure Coolant (LPCI) subsyste_m and RHR pumps *B" and 3 "D" comprising the other LPCI subsystem. These pumps are tested in accordance with j Technical Specification 4.5.A.3 to assure that adequate emergency core cooling capacity is - 1 available. During recent surveillance tests of the "A" and "C" RHR pumps, decreasing trends !
in Indicated pump differential pressure were noted. Based upon this data, a comprehensive 1 analysis of the entire RHR system.was performed. After investigating Instrumentation, 1
- valving, and other possible causes for the negative trend, the source of the problem appears to be internal to the RHR pumps. !
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x - , SAFETY EVALUATION Page 2 of 5 These' pumps need to be disassembled for an intemal inspection. This effort requires' extensive work such as: motor removal; pump disassembly; and close inspection of impeller, '
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? seals, bearings, and casing for wear and degradation.
These' pumps _ are needed during the 1990 refueling outage to operate in the ' shutdown cooling mode of RHR (the redundant RHR pumps, *B" and "D," motors are being overhauled during the outage for environmental qualification). Therefore, it is desirable to refurbish these pumps during power. operation prior to the outage. Since the refurbishment takes longer than the current Technical Specifications allowed outage time of 7 days, the Authority requests that a one time extension of the LPCI and Containment Cooling subsystems Limiting Condition for Operation'(LCO) from 7 days to 14 days be granted. This will provide sufficient time to refurbish the pumps prior to their use during the outage.
The next surveillance test for these RHR pumps is scheduled for the week of February 25, 1990. Based upon the past performance of these pumps, the Authority is concerned that ;
one or both of them may not meet the current Technical Specification acceptance criteria of-9,900 gpmi Should~a pump not pass the surveillance test, then plant shutdown would be ,
required. Based on the General Electric safety evaluation (Reference 5), the Technical Specification flow requirement can be reduced to 8,910 gpm and stili continue to meet the acceptance criteria for ECCS requirements.
Imptomentation of the proposed Technical Specification changes will 1) allow plant operation with a reduced RHR flow rate until such time that RHR pumps "A" and "C" can be refurbished (approximately a 3 week period), and 2) provide 14 days to overhaul the pumps during power operation.
j p Section lli IMPACT OF THE PROPOSED CHANGES f
The allowed outage time for the RHR and Core Spray subsystems were determined such. l "that the average risk rate for repair would be no greater than the basic risk considering single failure (l.e., loss of a subsystem) should be less than 30 days " (Technical Specification h BASES 3.5.A). The seven day LCO was then conservatively selected based upon .
judgements of the reliability of the remaining systems (i.e., both Core Spray subsystems and the other LPCI subsystem) and the expected fix and repair time. The proposed one time extension of the LPCI. and. Containment Cooling LCO to 14 days does not exceed the t analytically determined allowable repair period of 30 days,
' Analyses have been performed by the General Electric Co. (Reference 5) which demonstrate t
- that the RHR system is capable of performing its intended function (s) at the reduced flow rate of 8,910 gpm. These analyses considered the following functions of the RHR system: 1) LPCI Injection (post LOCA and following an Appendix R fire event); 2) suppression pool cooling; 3) shutdown decay heat removal; and 4) containment spray performance.
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- 5 5 SAFETY EVALUATION' Page 3 of 5 I
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The design of the RHR pumps is based on the higher flow rate requirements 'of LPCI following g"
a large break LOCA. Sensitivity studies on the FitzPatrick plant LOCA analysis (Reference 4) demonstrate an increase in the fuel peak clad temperature (PCT) by 88 F for a 10% ,
- below the allowable limit of 2200 F. Therefore, the FitzPatrick plant would still meet'all requirements of 10 CFR 50.46 and 10 CFR 50, Appendix K, with over 500 F margin assuming a 10% reduction in RHR pump flow rate. i LPCI system performance during postulated Appendix R events is not as limiting as the LOCA i scenarios.' The worst case Appendix R event is a catastrophic fire in the control room which -
assumes that RHR pump *B" is used in the LPCI mode to replench core. inventory. The resultant increase in FCT is less than G0 F assuming an RHR flow rate of 8,910 gpm. 1 Therefore, the fuel peak cladding temperature for this event remains significantly below the 1500 F temperature at which cladding damage may be expected.
Use'of the RHR system for suppression pool cooling, shutdown cooling, and containment -
spray _ cooling is unaffected by RHR pump flow rates at 8,910 gpm.
Recent probabilistic risk analyses show that the likelihood of the large break LOCA is approximately 10 3 er p year. With individual ECCS subsystems having a probability of failure in the range of 10'3 per year, it is highly unlikely that a LOCA would occur during the 14 day
_ period with Insufficient ECCS mitigation capability. .
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- Section IV EVALUATION OF EXIGENT SITUATION
~ As stated in Section ll above,- the Authority requests that this Application for Amendment be processed on an exigent basis under the provisions of 10 CFR 50.91(a)(6), since insufficient i time exists to allow for a 30 day public comment period.
10 CFR 50.91(a)(6) defines an exigent situation as:
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" . a licensee and the Commission must act quickly and that time does not permit the Commission to publish a FEDERAL REGISTER notice allowing 30 days for prior public comment, ..."
s The proposed changes are required prior to the next scheduled surveillance test of the RHR D pumps on February 25,1990. Since that date is less than 30 days from the date of this application, insufficient time is available to permit a 30 day comment period.
10 CFR 50.91(a)(6)(vi) requires the Authority to address the following concern before the NRC can make a determination that an exigent condition exists:
L *The Commission will require the licensee to explain the exigency and L why the Uconsee cannot avoid it."
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. -C SAFETY EVALUATION
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The degradation of the RHR pumps is determined based upon recent performance trends.
, The extent of the degradation and root cause was not identified until January 1990. It should
, be noted that the RHR pumps have not yet failed to meet any Technical Specification 0 requirement. This proposed change is based upon past performance trends.
m 'Section V EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the proposed Amendment would not involve a significant hazards consideration as stated in 10 CFR 50.92, since it would not:
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1, involve a significant increase in the probability or consequences of an accident -
1 previously evaluated.
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The RHR system is designed to mitigate the consequences of analyzed accidents and is normally in the standby mode. This system cannot initiate accidents and the proposed changes have no effect on the probability of occurrence of previously evaluated accidents.
.The one time extension of the LPCI and Contairraent cooling LCO reduces the level of redundancy.in the number of low pressu.e systems available to mitigate the consequences of an accident. During the time that one subsystem is out of service, the redundant subsystem as well as both Core Spray subsystems will be available to mitigate an accident.
The effect of a reduction of the RHR purcp flow rates has been fully analyzed. These analyses demonstrate that the consequances of postulated accidents remains well within the acceptable limits established in the FitzPatrick FSAR and applicable federal regulations. The 88 F expected increase in peak clad temperature is not significant with respect to the existing 600 F margin to the 2200 F acceptance criteria. ,
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- 2. create the possibility of a new or different kind of accident from any accident i previously evaluated. l
-1 The proposed changes do not involve hardware changes at the FitzPatrick plant. No !
actions taken as a result of the proposed changes can initiate any type of accident. q
- 3. Involve a significant reduction in a margin of safety.
Technical Specifications currently allow for one LPCI and Containment Cocling subsystem to be out of service for up to seven days. During this time, redundant systems (two Core Spray subsystems and the other RHR system) are still available to
- mitigate the consequences of an accident. The proposed one time extension to 14 l
. days does not significantly affect the level of safety afforded by the ECCS systems. 1 l
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' Attachment ll= 4 a SAFETY EVALUATION Page 5 of 5 1
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- The effect of a 10%~ reduction in the RHR pump flow rate has been fully analyzed.. .]
Although the calculated fuel PCT has increased by .88 F, this is not significant with
- respect to the 600 F margin to the ECCS acceptance criteria of 2200 F. ;
-i Section VI IMPLEMENTATION OF THE PROPOSED CHANGE 1
Implementation of the proposed changes will not impact the ALARA or Fire Protection 1 Programs at the FitzPatrick plant, nor will the changes impact the environment. l 1
Section Vll CONCLUSION a-l
' The change, as proposed, does not constitute an unreviewed safety question us defined in 10 CFR 50.59. That is, it: ,
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- a. will not change the probability nor the consequences of an accident '
malfunction of equipment important to safety as previously evaluated in tha
Safety Analysis Report 1
- b. will not increase the possibility of an accident or malfunction of a different type !
from any previously evaluated in the Safetv Analysis Report; H
- c. will not reduce the margin of safety as defined in the basis for any technical l specification;' I
- d. does not constitute an unreviewed safety question; and
- o. Involves no significant hazards consideration, as defined in 10 CFR 50.92.
L Section Vill REFERENCES 2
-1. James A. FitzPatrick Nuclear Plant Updated Final Safety Analysis Report.
- 2. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972, and Supplements.
- 3. James A. FitzPatrick Nuclear Power Plant SAFER /GESTR LOCA . Loss of Coolant Accident Analysis, NEDC-31317P, dated October,1986.
L 4. Sensitivity of the James A. FitzPatrick Nuclear Power Plant Safety Systems L Performance to Fundamental System Parar?!ars, Proprietary General Electric P
Report, MDE 83-0786, dated July,1986.
L 5. General Electric Co., Nuclear Safety Evaluation For a 10% Decrease in LPCI l Flow, JAF-SE 90-024, dated February 5,1990.
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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT 4
- Docket No. 50-333 J c;
DPR 59 '
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