ML20093C135
| ML20093C135 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 10/06/1995 |
| From: | Walter MacFarland PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9510120218 | |
| Download: ML20093C135 (3) | |
Text
.
. Writ::r G. MacFtrl:nd, IV, P.E.
Vce President -
_g Limerick Generating Station
=agr PECO ENERGY
= ge -
Sanatoga. PA 19464 0920 610 718 3000 Fax 610 718 3008 Pager 1800 352 4732 #8320 October 6,1995 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85
' U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
Umerick Generating Station, Units 1 and 2 Main Steam Safety Relief Valve / Emergency Core Cooling System Action Plan On September 11,1995, Limerick Generating Station (LGS), Unit 1, was manually scrammed in response to an unexpected opening of a Main Steam Safety Relief Valve (MSRV) when the valve could not be closed within two minutes. During the response to this event, Operations observed indications of suction strainer fouling on the Residual Heat Removal (RHR) pump being used for suppression pool cooling. Inspection of the MSRV revealed that steam erosion due to p"ot disc leakage resulted in failure of the pilot valve which equalized pressure across the main disc and caused the valve to open. Inspection of the RHR pump suction strainer identified a brown, fibrous naterial covering approximately 70% of the stralner. Chemical amilysis identified the material as polypropylene fibers, and iron and zinc oxide corrosion products. The polypropylene fibers are not a constituent of any permanent Primary Containment equipment.
On September 13,1995, the NRC sent a team of three inspectors to LGS to review the details of these two events, including PECO Energy Company's identification of the causes and corrective actions. On September 21,1995, the NRC Team conducted their inspection exit meeting, and requested that PECO Energy provide the NRC with a letter describing the details of the following action plans for LGS, Units 1 and 2: 1) MSRV tailpipe temperature,2) Emergency Core Cooling l
System (ECCS) pump suction strainer differential pressure, and 3) suppression pool water cleanliness. The Attachment to this letter provides the detal!s of these action plans.
If you have any questions or require additional information, please contact us.
Very truly yours,
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(AdvG, dos 2 ;>
Attachment cc:
T. T. Martin, Administrator, Region I, USNRC w/ attachment N. S. Perry, USNRC Senior Resident inspector, LGS
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Attachment "
- Decket Nos. 50452 and 50453 October 6,1995 Page 1 of 2 Main Steam Safety Rollef Valve (MSRV) Tall Pipe Tamparature Action plan l
For any M8RV temperature 22257 (Alert Level):
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Operations shall log the affected MSRV taH pipe temperature every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> while operating at l
power (and every six hours during startup).
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Engineering shall trend the affected MSRV tau pipe temperature versus time and project when
~ he temperature is expected to reach 275*F. This projection shall be based on historical trends t
l (of the Unit 1 "M" MSRV from March 1994 to September 1995) and Industry experience. In L
addition, preparations for a planned outage to replace the affected MSRV wRI be initiated.
For any MSRV temperature 22607 (Action Level):
Based on tall pipe temperature trends and suppression pool heat input, Engineering shall provide specific recommendations of when to schedule a planned outage to replace the affected MSRV, i.e., before the MSRV tall pipe temperature is expeted to reach 2757. ',n the event of -
sudden and significant temperature increases, operabHity at the affected MS'N would be evaluated along with potential mitigating actions.
Basis:
The alert level of 2257 was selected since it represents a minor MSRV leak (l.e., approximately 20 lbm/hr).
The action level of 2507 was selected since it represents a more significant MSRV leak (i.e.,
500-1000 lbm/hr). However, based on historical tail pipe temperature trends, there is adequate time to schedule a planned outage to replace the affected MSRV. The Unit 1 "M" MSRV lift event occurred after the tar pipe temperature gradually incressed to 2957 over an 18 month period of time due to severe phot stage leakage and erosion. In particular, the rise in tail pipe temperature from 250T to 2757 took over nine (9) months. Additionally, from a suppression pool heat-up standpoint, 250T is a conservative action level based on analysis that if all 14 MSRV's leaked on a particular unit et 1000 lbm/hr per MSRV, the resultant heat input would be well within the capabHity of the suppression pool cooling system.
The temperature of 2757 was selected since this temperature provides adequate margin to l
2957 based on historical tan pipe temperature trends for conducting a planned orderly shutdown to replace the affected MSRV. For example, the Unit 1 "M" MSRV tar pipe temperature rise from 2757 to 2957 took six months.
Sudden and significant increases in MSRV taH pipe temperatures are atypical. The phot stage steam erosion of the significance that led to the Unit 1 "M" MSRV lifting event would take a long period of time to occur.
l The temperature levels specified above apply for MSRV phot stage leakage which la the worst L;
case MSRV leakage condition, and one which In the most severe condition could potentially lead to spurious MSRV actuation. Since Limerick Generating Station (LGS) as currently configured cannot distinguish between plot stage versus main seat leakage, all MSRV leakage shall
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i Attachmerd Docket Nos. 50 352 and 50-353 October 6,.1996 Page 2 of 2
_ conservatively be assumed to be pNot stage leakage and the above plan shall be followed. A.
design change to distinguish between pNot stage versus main seat leakage shall be evaluated. - if
. such a design change is determined to be feasible, and la successfully incorporated, this.MSRV
. tan Pipe Temperature Action Plan shall only apply to MSRV phot stage leakage.
Emergency Cars 0="-.a Sy". fECCS) Pumn Bar^'
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l Current ECCS/ Reactor Core Isolation Cooling (RCIC) Pump, Valve and Flow (PV&F) operabNity l
tests shall trend suction strainer differential pressure (DP) values versus baseline suction strainer j
values. An increase of 0.5 paid for the suction strainer wHl currently be the alert level. At this level, Engineering shall perform an evaluation to project when the affected ECCS suction strainer DP wHl reach the maximum allowable DP based on DP versus time trends, and shall recommend
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appropriate actions.
i The maximum allowable DP across an ECCS/RCIC pump suction strainer is that required to j
satisfy design not positive suction head requirements at rated flow. The specific design calculations are currently being reviewed by Nuclear Engineering Division. The finalized action i
values wNI be avaNable by November 1,1995. This acceptance criteria shall be incorporated into each ECCS/RCIC PV&F operabHity test.. If any value is exceeded, the associated ECCS shall be declared inoperable and the appropriate Technical Specifications (TS) action followed.
4 In special circumstances where extended operation of an ECCS pump is required, such as suppression pool cooling mode cf the Residual Heat Removal (RHR) system, additional suction
' strainer data trending wNI be evaluated and recommended by Engineering.
i Suppression Pool Watar Cleanlinema Action Plan i
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Chemistry shall sample and trend suppression pool water for fibrous content on a monthly basis.
This fiber sampling will be used for Information only. Actions shall only be taken on the basis of j
suction strainer DP as described above.
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