ML20011F010
| ML20011F010 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 02/14/1990 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20011F011 | List: |
| References | |
| NUDOCS 9002230444 | |
| Download: ML20011F010 (8) | |
Text
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10 CPR 50.90 o,
C PHILADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955 65 CHESTERBROOK BLVD.
WAYNE. PA 19087 5691 pts) sso sooo February 14, 1990 F
Docket No. 50-277 Liciise Hc DPR-44 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.
C.
20555
SUBJECT:
Peach Bottom Atomic Power Statlers, Unit 2 Technical Specifications Change Request
Dear Sir:
Philadelphia Electric Company (PCCC) hereby submits Technical Specifications Change Request No. 30-04, in accordance with 10 CFR 50.90, requesting an emergency isn.porary amendment to the Technical Specifications (Appendix A) of Operatlng License No. DPR-44 for Peach Bottom Unit 2.
The current alttation at Peach Bottom Unit 2 with one Automatic Depressurization System (ADS) valve inoperable, conforms to the 10 CFR 50.914a)(5) description of an emergency situation.
To prevent a plant shutdown in compliance with a Limiting Condition for Operation, PECo requests that this application be processed as an emergency change as permitted by 10 CFR 50.91(a)(5) and that.a Temporary Walver of Compliance be granted until the license amendment is approved.
A safety evaluation supporting this request is attached.
Attachment I contains information describing this change and supporting a finding of No Significant Hazards.
contains the revised Technical Specification page.
Currently, one of the five ADS valves is inoperable.
Correcting the problem requires a shutdown to access equipment in the drywell.
In accordance with Technical Specification 3.5.E.3, an orderly shutdown must be initiated at 1355 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.155775e-4 months <br /> today, February 14, 1990, and reactor pressure must be reduced to at least 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Because a midcycle outage is scheduled to begin on March 3, 1990 (seventeen days from now) and because, until that time, the unit is limited to approximately f
74% of rated power due to Feedwater heater tube leaks, continued
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operation until the midcycle outage with one inoperable ADS valve f 'O\\
9002230444 900214' PDR ADDCK 05000277 PNU p
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%Do;um:nt Control Deck FCbruory 14, 1990 TSCR 90-04 Pcg2 2 is justified.
The demand on the ADS system in the event of a Loss of Coolant Accident (LOCA) is reduced by the lower reactor power level.
The attached Safety Evaluation demonstrates that the four remaining operable ADS valves fulfill the required depressurization capability as described in the Updated Final Safety Analysis Report.
PECo proposes that this situation satisfies the emergency criteria set forth in 10 CFR 50.91(a)(5) because i
failure to act in a timely manner would result in a shutdown of the unit.
PEco has applied for this amendment in a timely fashion and could not have foreseen this emergency situation.
On the morning of February 13, 1990 NRC Region I and NRC Headquarters personnel were contacted with respect to relief from the Technical Specification shutdown requirement.
Although the inoperability of the valve was traced back to February 7, 1990, identification of this situation at that time would not have allowed sufficent time for the normal or exigent processing of a license amendment.
The abnormal results of a routine surveillance test conducted on February 7, 1990 were determined to be acceptable relative to ADS valve operability.
However, on February 12, 1990 after observing similar results when performing the same test on another ADS logic channel, it was determined that the results on February 7, 1990 were misinterpreted.
PECo promptly initiated the compensatory measures required by the Technical Specifications and contacted the NRC.
In no way has PECo intended to abuse the emergency provision of the NRC regulations by failing to make a timely application for a license amendment nor could PECo have avoided the emergency situation.
Therefore, PECo respectfully requests that the NRC process this amendment application in an emergency manner and that the NRC grant a Temporary Waiver of Compliance for the period of time until the amendment is issued.
If you have any questions regarding this matter, please contact us.
V.ry truly yours, ft G. A. Hunger, Jr.
Director Licensing Section Nuclear Services Department Attachments cc:
J. J. Lyash, USNRC Senior Resident Inspector W.
T. Russell, Administrator, Region I, USNRC
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i CdMMONWEALTH OF PENNSYLVANIA :
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COUNTY OF PHIIADELPHIA l'
Corbin A. McNeill, Jr., being first duly sworn, deposes and says:
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i That he is Executive Vice President of Philadelphia Electric Company; j
i-the Applicant herein; that he has read the Technical Specifications Change Request No.:90 04, and knows the contents thereof; and that the statements and r
mattera set forth therein are true and correct to the best of his knowledge, information and belief.
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i Executive Vice President I
Subscribed and sworn to before me this /dday l-of 1990.
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i Notary Public I-tJ01 AtuAL SEAL SUSAN C.11tORNTOtt, Notary Public City of Philadelphia, Phila. Countf I My Commlssion Cxpires IAdy 4,1992 I"
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' ATTACHMENT 1 i
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i PEACll BOTTOM ATOMIC POWER STATION
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UNIT 2 i
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Docket No. 50-277 l
License No. DPR-44
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TECHNICAL SPECIFICATION CRANGE REQUEST l
NO. 90-04 i
" Emergency' Extension of Automatic Depressurization System (ADS) Limiting Condition for Operation Shutdown
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i Requirement" t
i Supporting Information-for Changes t
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Dockot No. 50-277 t
License No. DPR-44 f
Description of Changes:
PECo proposes to add a note to page 131 of the Technical Specifications stating that the seven-day Limiting Condition for p
Operation (LCO) of Specification 3.5.E.2 is extended to 11:59 p.m. on March 3, 1990 provided that the Reactor Core Isolation i
Cooling (RCIC) System is operable.
The RCIC operability
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provision is proposed as a compensatory measure since RCIC can be used for depressurization and high pressure coolant make up.
The purpose of this temporary amendment is to avoid a plant shutdown /startup cycle and the deleterious effects on the
_ primary system pressure. boundary materials and plant equipment which are subject to thermal transients and additional duty during each shutdown /startup cycle.
PECo is not prepared to begin all of the maintenance activities scheduled for the approaching midcycle outage ahead of schedule.
Necessary personnel and materials are not fully available nor properly staged to support an early start of the outage.
Thus, PECo is requesting emergency Technical Specification relief to avoid either a shutdown /startup cycle tc repair the ADS valve and resume power operation prior to the midcycle outage, or the early 1
start of the outage for which PECo is not fully prepared.
During the period that this Technical Specification change is in effect, one of the five ADS valves will be Dockot NO. 50-277 License No. DPR-44 I
considered inoperable.
Therefore, the compensatory measures required by the Technical Specifications, weekly testing of the High Pressure Coolant Injection System and the operable ADS valve t..
I actuation logic channels, will be carried out.
Additionally, RCIC-system operability is being made a condition of the ADS LCO.
If another ADS valve, HPCI or RCIC become inoperable during this time-period the shutdown required by Technical Specification
. 3.5.E.3 Will be initiated _and reactor pressure will be reduced to at least 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
No-Significant Hazards' Consideration Determination 1
PECo proposes that this temporary Technical Specification change does not involve Significant Hazards
- Considerations' based on the following.
1 )-
The proposed revision does not involve a significant increase in the probability or consequence of an accident previously evaluated.
Having one ADS valve out of service is not related to any initiating sequence of an accident scenario.
The existing condition with four ADS valves operable with
= low. pressure Core Standby Cooling Systems (CSCS) in conjunction with High Pressure Coolant Injection allows - -
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DockGt No. 50-277 g
License No. DPR-44 h
f-for single-failure' proof core cooling for intermediate and small break LOCA's.
One inoperable ADS valve does not affect the operability of the four remaining ADS
[-
_ valves.
SAR LOCA analyses were based on maintaining 4 f
In addition, Safe Shutdown will I
also be assured even with the K valve inoperable because for Safe Shutdown method D, only 2 of the 3 provided SRVs are required.
Calculation ME-384 has determined that only 2 SRV's are required for a low pressure CSCS to remove decay heat after reactor shutdown.
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The proposed revision does not create the possibility of a new-or different kind of accident from any accident previously evaluated.
A single ADS valve out of service has already been evaluated in the SAR.
The failure mode of the inoperable ADS valve (i.e., an electrical open circuit)
'does.not interfere with the pressure relief valve 0
function.
Ability to achieve core cooling during small and intermediate break LOCA's is not affected, and therefore the possibility of a different type of accident is not created.
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'*e DockGt No. 50-277 License No. DPR-44 j-.
iii)'The proposed revision does not involve a significant reduction in a margin of safety.
The current'LCO analysis assures 4 ADS valves are operable.
The large break LOCA Peak Clad Temperature (PCT) margin is unchanged since the ADS does not
,I initiate during this scenario.
The small and intermediate break PCT margin is unchanged since 4 ADS valves are operable.
The current small and intermediate break PCT is on the order of 1500 F, (Reference 1) which is significantly below the 2200 F PCT limit.
In addition,_if another ADS valve, HPCI or RCIC becomes p'
inoperable, an orderly shutdown will be initiated per Tech. Spec.
3.5.E.3.
Conclusion r
The Plant Operations Review Committee and the Nuclear
-Review Board have reviewed this proposed change to the-Technical Specifications and determined that it does not involve an Unreviewed Safety Question and will not endanger the health and.
safety of the public.
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