ML19210B419

Tech Spec Change Request 66 to DPR-50,App A,Re Updating of Valves Nameplate Data.Certificate of Svc Encl

ML19210B419
Person / Time
Site:	Three Mile Island
Issue date:	11/18/1977
From:	Herbein J METROPOLITAN EDISON CO.
To:
Shared Package
ML19210B418	List: ML19210B417 ML19210B419
References
NUDOCS 7911080633
Download: ML19210B419 (4)

Text

.

UNITD STATES CF A:iERICA NUCLEAR REGULATCRY CC:GIISSICII IN THE MATTER OF DCCKET NO. 50-289 LICENSE lio. DPR-50 METECPCLITA'I DISCN CC:GMPI This is to certify that a copy of Technical Specificaticn Change Request No. 66 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, en the date given belov, been filed with the U. S. Nuclear Regulatory Cc::: mission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States nail, addressed as follows:

Mr. Welden 3. Arehart Mr. Harr/ B. Reese, Jr.

Board of Supervisors of Board of County Consissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Read Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITXI EDISCN CCMPXTY

/

j f I

/ s

?y / l v '

/ 71*ce President Dated: November 18, 1977 O

1! ^U O 1ID'IJGb u833 191108

METROPOLITAN EDISO:T CCMPANY JERSEY CENTRAL PC'JER & LIGHT CCMPANY AND PENNSYLVA'TIA ELECTRIC COMPA'rf THREE MILE ISLA'iD ITUCLEAR STATIO:T UNIT 1 Operating License No. DPR-50 Decket :io. 50-289

echnical Specification Change Reauest '70. 66 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request ,

proposed replacement pages for Appendix A are also included.

METRCPOLITAN EDISON CCMPANY 3' 7

/ Vice President A

Sworn and subscribed to me this le day of A) & , 1977 Notary Publif NOTARY PUBUC Reaciq. Serirs County, Ps.

vy Commiss.cn Escirts N=.19,133 i fsuo C

Three Mile Island Nuclear Station, Unit 1 Operating License No. LPR-50 Docket No. 50-289 Technical Snecification Chance Recuest No. 66 -

The licensee requests that the attached revised page replace page 3-2 of the existing Technical Specification.

Reasons for Chance Recuest As stated in Event Report TT-Oh/lT, an error was discovered during an updating of the valves nameplate data. The incorrect value of the valves cc=bined relief capacity was not calculated frc= values as listed on the valve na:eplate.

Safety Analysis' Justifying Change Recuest This change has been reviewed and the RCS peak pressure was reco=puted.

The CADD analysis for the vorst case T4L3 accident analyzed in support of LII-1 operation with the high pressure trip set at 2h05 psig and the pressuriner safety valves set at 25C0 psig used a relief rate through the pressuriser code safety valves of 86 lbn/sec at 2500 psis, or 309,600 lb=/hr/ valve. The resulting RCS peak pressure was 27h9 5 psia.

• The analysis was repeated, assuming a relief rate through the code relief valves of 78 lbs/sec at 2500 psig, or 280,800 lb=/hr/ valve. The resulting RCS peak pressure is 275h psia. The increase of k psia agrees with .the esticate of 5 psia =ade by "MPR Associates, Inc.". This analysis supports operation of TMI-1 vith the 2kO5 psig H.P.T. setpoint and the 2500 pais code safety valve setpoint. This analysis was performed assuming a trip string pressure delay tine of 500 ms which is characteristic of the 59 PH sensor originally installed at TMI-1.

The same analysis was also perfor ed assuming a string pressure delay time of h50 ms, which is characteristic of the Rosemount sensor which was installed at TMI-1 during the outage. This analysis resulted in peak RCS pressure of 2749.8 psia at the pump discharge. This is 0.3 psi more than the 2749 5 psia peak pressure calculated criginally, assuming 86 lbm/sec relief rate at 2500 psig and a 500 s string delay time.

Therefore, with a taxi =un possible change of 7 psig: (i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased; (ii) the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created; (iii) the margin of safety defined in the bg. sis for any technical specification is not reduced.

15o6 190

^

Bases The limitation en power operation with one idle RC pu=p in each loop has been imposed since the ECCS cooling performance has not been calculated in accordance with the Final Acceptance Criteria requirements specifically for this node of reactor operation. A time period of 2h hours is allowed for operation with one idle RC pump in each loop to effect repairs of the idle pump (s) and to return the reactor to an acceptable combination -

of operating RC pumps. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for this = ode of ope 1ation is acceptable since this code is expected to have considerable targin for the peak cladding te=perature limit and since the likelihood of a LOCA within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is considered very re=ote.

A reactor coolant pu=p or decay heat renoval pu=p is required to be in operation before the boron concentration is reduced by dilution with makeup water. Either pump will provide mixing which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor.

Cne decay heat renoval pump will circulate the equivalent of the reactor coolant syste= volume in one half hour or less.

The decty. heat re= oval systes suction piping is designed for 300 F and 370 psig; thus, the systes can remove dec'ay heat when the reactor coolant system is below this tenperature. (2, 3)

One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the su= of the available heat sources which are pump energy, pressurizer heaters , and reactor decay heat. (h) Both press"-4-a- aode safety valves are required to be in service prior to criticality to confor= to the systes design relief capabilities. The code safety valves prevent overpressure for a rod withdrawal or feedvater line break accidents. (5) The pressurizer code safety valve lift set point shall be set. at 2500 psig t15 allowance for error and each valve shall be capable of relieving 230,300 lb/h of saturated steam at a ,l pressure not greater than three percent above the se: pressure.

References (1) FSAR, Tables 9-10 and h-3 through h-7 (2) FSAR, Sections h.2 5 1 and 9 5.2.3 (3) FSA3, Section h.2 5.h _

(h) FSA3, Sections L.3.10.h and h.2.h (5) FSAR, Section h.3 7 r m /

ILuo IO'I /

,..