ML19210A543

Corrections to Tech Spec Change Request 30 Supporting Licensee Request to Change OL DPR-50,App a Re Rod Position Limits.Certificate of Svc Encl

ML19210A543
Person / Time
Site:	Three Mile Island
Issue date:	01/20/1976
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19210A534	List: ML19210A533 ML19210A543
References
NUDOCS 7910300583
Download: ML19210A543 (5)

Text

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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Corrections to Technical Specification Change Request No. 30 These Technical Specification Change Request Corrections are submitted to correct the 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 these corrections, replacement pages for Change Request No. 30 are also included.

MSTROPOLITAN EDISON COMPANY By /s/ R. C. Arnold Vice President-Generation Sworn and subscribed to me this 20th day of Januaz7 , 1976 Lawrence L. Lawyer Notary Public 1488 026 7910800f83

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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND M e.:#: ;

PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Corrections to Technical Specification Change Request No. 30 These Technical Specification Change Request Corrections are submitted to correct the 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 these corrections, replacement pages for Change Request No. 30 are also included.

METROPOLITAN EDISON COMPANY G , .

By / /) .4 ~ 0 _J(

Vics President-Generation Sworn and subscribed to me this day of , 1976

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CW1%$5i2 l' Y UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of egrrections to Technical Specification Change Request No. 30 to Appendix A of the Operating License for Three Mile Island Nuclear Station, Unit 1, dated Jant.sry 21, 1976, and filed with the U.S . Nuclear Regulatory Coc: mission January 21, 1976, has this 21st day January,1976, been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania, and of Dauphin County, Pennsylvania, by deposit in the United States Mail, addressed as follows:

Mr. Weldon B. Arehart, Chairman Mr. Charles P. Hoy, Chaiman Board of Supervisors of Board of County Commissioners of Londonderry Township Dauphin County R.D. #1, Geyers Church Road Dauphin County Courthouse Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY By /s/ R. C. Arnold Vice President-Generation 1488 028

l Metropolitan Edison Co. (Met-Ed)

Three Mile Island Nuclear Station Unit 1 (IMI-1)

Operating License No. DPR-50 Docket No. 50-289 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 30 The licensee requests that the attached changed pages replace pages 2-1, 2-2, 2-3, 2-5, 2-6, 2-7, 2-9, 3-16, 3-34, 3-35, 3-36, figures 2.1-1, 2 & 3, figures 2.3-1 & 2, and figures 3.5-2A thru F of the existing Technical Specifications.

REASON FOR PROPOSED CFANGE Thesc changes to technical specifications are necessary to ensure safe operation of TMI-l at rated power of 2535 MWe for the duration of Cycle 2 and are based on a Cycle 1 burnup of 440 + 10 EFPD.

Changes to the present technical specifications are necessary as a result of: the effects of introducing 56 fresh batch 4 fuel assemblies combined with relocation of once bucaed Batch 2 and 3 fuel assemblies; use of the B&W-2 CHF correlation with a 95/95 confidence level, and extended pressure application to 1750 psi; use of a RC flow equal to 106.5% of Cycle 1 design flow; and ECCS Final Acceptance Criteria (FAC).

The 56 batch 4 fuel assemblies are not in general the technical specifi-cation limiting assemblies. Their presence combined with relocation of the once burned batch 2 and 3 assemblies produces a redistribution of fuel and assemblies which results in changed core physics and thermal-hydraulic calculations. Further burnup and the cycle 2 locations of the batch 2 and 3 assemblies results in these assemblies being the limiting assemblies thermally and mechanically. Other factors that were considered in the derivation of the Cycle 2 specification limits are the slight dif ferences between the new and once burned fuel assemblies. These minor differences are reduced active length, slightly higher pellet density, and improved flow characteristics for the new assemblies compared to the burned assemblies.

In addition to Fuel changes, the use of the B&W-2 CHF correlation combined with the asemed minimus Flow of 106.5% have had an influence on these proposed specifications. Use of this correlation and flow more realistically predict core pet.ormance but still provide conservative technical specifica-tion limits.

Met-Ed submitted revised technical specifications based on FAC guidelines in our Technical Specification Change Request 17 (August 8,1975) . Additional ECCS supporting information was provided in our letters of April 19, 1975, July 9,1975, July 15,1975, and October 23, 1975. The attached changed pages for TMI-1 Cycle 2 operation include changes that were requested in Change Request 17 which apply to Cycle 2 operation. The number 17 beside the margh.a1 bars indicatec those changes that were requested in Change Request 17 and all other marginal bars indicate Cycle 2 changes. All appropriate cycle 2 Technical Specifications were developed based on FAC guidelines.

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The low pressure (1800 psig) and variable low pressure (1175 Tout - 5103) trip setpoint shown in Figure 2 3-1 have been established to maintain the DNB ratio greater than or equal to 13 for those design accidents that result in a pressure reduction (3.h).

Due to the calibration and instru=entatio: errors, the safety analysis used a variable low reactor coolant system pressure trip value of (11.75 Tout - 51h3).

d. Coolant outlet te=perature The high reactor coolant outlet temperature trip setting linit (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant te=peratures in the operating range.

The calibrated range of the temperature channels of the RPS is 520 to 620 F. The trip setraf it of the channel is 619 F. Under the verst case environ =ent, power supply perturbations, and drift, the accuracy of the trip string ist,lF. This accuracy was arrived at by su= sing the worst case accurecies of each module. This is a conservative method of error analysis since the normal procedure is to use the root mean square method.

Therefore, it is assured that a trip vill occur at a value no higher than 620F even under vorst case conditions. The safsty analysis used a high temperature trip set point of 620F.

The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drift, linearity, repeatability, etc. This does not imply that the equipment is restricted to operation within the calibrated range. Additional testing has de=onstrated that in fact, the temperature channel is fully operational approxinately 10% above the calibrated range.

Since it has been established that the channel vill trip at a value of RC outlet temperature no higher than 620F even in the vorst case, and since the channel is fully operational approximately 10% above the calibrated range and exhibite no hysteresis or foldover characteristics, it is concluded that the instrument design is acceptable.

e. Reactor building pressure The high reactor building pressure trip setting li=it (k psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident , even in the absence of a low reactor coolant systes pressure trip.

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