ML19262A472

Tech Spec Change Request 25 Supporting Licensee Request to Change DPR-50,App a Re 2750 Psig RCS Pressure Safety Limits. Certificate of Svc Encl

ML19262A472
Person / Time
Site:	Three Mile Island
Issue date:	11/17/1975
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19262A471	List: ML19262A470 ML19262A472
References
NUDOCS 7910300397
Download: ML19262A472 (5)

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METROPOLITAN EDISON COMPAh7 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 Technical Specification Channa Request No. 25 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 Ir. land Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appandix A are also included.

METROPOLITAN EDISON COMPANY

1. 1 By [

Vice PresGent-Gene' ration Sworn and subscribed to me this /7 M day of o%h , 1975 M -

vu, Notary Public

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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 Technical Specification Change Request No. 25 to Appendix A of the Operating License for Three Mile Island Nuclear Station, Unit 1, dated November 17, 1975, and filed with the U. S. Nuclear Regulatory Commission November 17, 1975, has this 17th day November,1975, been served on the chief executives of Londonderry Tcwnship, Dauphin County, Pennsylvania, and of Dauphin County, Pennsyl-vania, by deposit in the United States Mail, addressed as follows:

Mr. Weldon B. Arehart, Chairman Mr. Charles P. Hoy, Chairman Board of Supervisors of Board of County Commissioners of Londonderry Township Dauphin County R.D.#1, Geyers Church Road Dauphin County n arthouse Middletown, Pennsylvania 17957 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY By /s/ R. C. Arnold Vice President - Generation 1486 279

THREE MILE ISLAND NUCLEAR STATION UNIT 1 (IMI-1)

OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 25 The licensee requests that the attached changed pages replace pages 2-4 and 4-8 of the existing Technical Specifications.

Please note that page 4-8 has been requested to be changed in our change request No. 7 Amendment No.1 and change request No. 19 dated October 28, 1975 and August 23, 1975 respectively.

REASON FOR PROPOSED CHANGE The proposed change will provide clarification to the 2750 psig reactor coolant system pressure safety limit. Removing specification 2.2.2 will ensure that the settings for the code safety valves are not construed to be additional reactor coolant system pressure safety limits.

According to Articles N-910.5, N910.6 and N910.7 of the ASME Boiler and Pressure Vessel Code,Section III, Article 9, Winter 1968:

"The required relie ing capacity shall be secured by the use of at least two pressure relief devices... pressure setting of at least one of the pressure relief devices connected to any vessel or system shall not be greater than the design pressure of the vessel (at design temperature) which it protects. Additional pressure relief devices may have higher pressure settings but in no case shall these settings be such that the total accumulated pressure exceed 110% of the design pressure as set forth in N-910.4. ..In determining the setting pressures and discharge capacities required to comply with these rules, full account shall be taken of the pressure drop on both inlet and discharge sides of the pressure relief devices at full discharge conditions."

It is clearly the intent of the ASME Code to ensure that the appropriate safety limit (2750 psig in this case) is not exceeded. Therefore, as the code safety valves are set in accordance with the code as noted on proposed changed page 4-8 and the basis of proposed changed page 2-4, this setting ensures that the safety limit of 2750 psig is not exceeded.

SAFETY ANALYSIS JUSTIFYING CHANGE The proposed change does not alter the design capabilities of the system or the required setpoints of the pressurizer code safety valves. Should the transient continue to increase the pressure, the two code safety valves will insure that the safety limit is not exceeded. The proposed change does not involve an unreviewed safety question nor does it cause undue risk to the health and safety of the public.

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2.2 SAFE"Y LIIIITS - REACTOR SYSTEM PRESSURE m.

Anplicability Applies to the limit on reactor coolant system pressure.

Objective

'To maintain the integrity of the reactor coolant syste= and to prevent the release of significant amounts of firsion product activity.

Snecification

~2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel.

Bases The reactor coolant system (1) serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere. In the event of a fuel cladding failure, the reactor coolant system is a barrier against thr. release of fission products. Establishing a system pressure limit helps to assure the integrity of the reactor coolant system. The maxi =wn transient pressure allowable in the reactor coolant system pressur essel under the AS:E Code,Section III, is 110 percent of design pressure. 2 The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under A:ISI Section B31.7 is 110 percent of design pressure. Thus, the safety

%, limit of 2~5 established.{ sig The (110settings percent for of the the2500 psighigh reactor designessure pressure) has been trip (2355 psig) ard the pressurizer code safety valves (2h35 psig)( have been established in accordance with ASME Boiler and Pressurizer Vessel Code,Section III, Article o, Winter,1968 to assure that the reactor coolant system pressure safety limit is not exceeded. The initial hydrostatic test was conducted at 3125 psig (125 percent of design pressure) to verify the integrity of the reacter coolant system. Additional assurance that the reactor coolant system pressure does not exceed the safety lim t is provided by setting the pressuriser electromatic relief valve at 2255 psig. b)

REFEREIICES (1) FSAR, Section h (2) FSAR, Secticn h.3.10.1 jfQf }gl (3) FSAR, Section h.2.h (h) FSAR, Table h-1 mv C=b

TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency

1. Control Rods Rod drop time of all Each refueling shutdown full length rods

2. Control Rod Movement of each rod Every two weeks, when reactor Movement is critical

3. Pressurizer Safety Setpoint 50% each refueling period Valves

4. Main Steam Safety Setpoint 25% each refueling period Valves

5. Refueling System Functional Start of each refueling period Interlocks

6. Main Steam (See Section 4.8)

Isolation Valves

7. Reactor Coolant Evaluate Daily, when reactor coolant System Leakage system temperature is greater than 525 F

8. Air Treatment See Section 3.15 See Section 4.12 Systems

9. Spent Fuel Cooling Functional Each refueling period prior to System fuel handling

10. Intake Pump House (a) Silt Accumulation- Each refueling period F loo r Visual inspection of Intake (Elevation 262 Ft Pump House Floor 6 in.) (b) Silt Accumulation Quarterly Measurement of Pump House Flow

• The setpoint of the pressurizer code safety valves shall be in accordance with ASNE Boiler and Pressurizer Vessel Code,Section III, Article 9, Winter,1968.

1486 282 4-8