ML20205H078

Proposed Tech Specs Adding LCO Action Statements,Making SRs More Consistent with NUREG-1430,correcting Conflicts or Inconsistencies Caused by Earlier TS Revs & Revising SFP Sampling

ML20205H078
Person / Time
Site:	Three Mile Island
Issue date:	04/01/1999
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20205H074	List: ML20205H069 ML20205H078
References
RTR-NUREG-1430 NUDOCS 9904080048
Download: ML20205H078 (14)

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I Enclosure 2 TMI-l License and Technical Specification Revised Pages l

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i 9904080048 990401 PDR ADOCK 05000289 i P PDR ' j i i i

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i 1.4.2 REACTOR PROTECTION SYSTEM The reactor protection system is described in Section 7.1 of the Updated FSAR. It is that combination of protection channels and associated circuitry which forms the automatic system that protects the reactor by control rod trip. It includes the four protection channels, their associated instrument channel inputs, manual trip switch, all rod drive control protection trip breakers, and activating rciays or coils.

l 1.4.3 PROTECTION CHANNEL l 1 A PROTECTION CHANNEL as described in Section 7.1 of the updated FSAR (one of three l or one of four independent channels, complete with sensors, sensor power supply units, amplifiers, and bistable modules provided for every reactor protection safety parameter) is a j combination ofinstrtunent channels forming a single digital output to the protection system's l coincidence logic. It includes a shutdown bypass circuit, a protection channel bypass circuit and a reactor l

trip module.

1.4.4 REACTOR PROTECTION SYSTEM LOGIC This system utilizes reactor trip module relays (coils and contacts)in all four of the protection I

channels as des:ribed in Section 7.1 of the updated FSAR, to provide reactor trip signals for l de-energizing the six control rod drive trip breakers. The control rod drive trip breakers are arranged to provide a one-out-of-two-times-two logic. Each element of the one-out-of-two-times-two logic is controlled by a separate set of two-out-of-four logic contacts from the four reactor protection channels.

1.4.5 ENGINEERED SAFETY FEATURES SYSTEM This system utilizes relay contact output from individual channels arranged in three analog sub-systems and two two-out-of-three logic sub-systems as shown in Figure 7.1-4 of the updated FSAR. The logic sub-system is wired to provide appropriate signals for the actuation ofredundant engineered safety features equipment on a two-of-three basis for any given  ;

parameter.

1.4.6 DEGREE OF REDUNDANCY The difference between the number of operable channels and the number of channels which, when tripped, will cause an automatic system trip.

1.5 INSTRUMENTATION SURVEILLANCE 1.5.1 TRIP TEST A TRIP TEST is a test oflogic elements in a protection channel to verify their associated trip action.

1-3 Amendment No. W, W

1 a.l.12.3 If the reactor vessel head is installed and Tavg is $332 F, High Pressure

Injection Pump breaker shall not be racked in unless

l a. MU-V16A/B/C/D are closed with their breakers open, and MU-V217 is Closed and Pressurizer level is $220 inches. If pressurizer level is

>220 inches, restore level to 5220 inches within I hour

Or "

b. Emergency plant condition exists that requires High Pressure injection. 4 3.1.12.4 The PORV Block Valve shall be OPERABLE during HOT STANDBY, STARTUP, and POWER OPERATION:

a. With the PORV Block Valve inoperable, within I hour either:

1. restore the PORV Block Valve to OPERABLE status or

2. close the PORV (verify closed) and remove power from the PORV i

3. otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With the PORV Block Valve inoperable, restore the inoperable valve to OPERABLE status prior to startup from the next COLD SHUTDOWN unless the COLD SHUTDOWN occurs within 90 Effective Full Power Days (EFPD) of the end of the fuel cycle. If a COLD SHUTDOWN occurs within this 90 day period, restore the inoperable valve to OPERABLE status prior to startup for the next fuel cycle.

Bases If the PORV is removed from service while the RCS is below 332 F, sufTicient measures are incorporated to prevent severe overpressurization by either eliminating t! s high pressure sources or flowpaths or assuring that the RCS is open to atmosphere.

The PORV setpoints are specified with tolerances assumed in the bases for Technical Specification 3.1.2.: Above 287 F (275 F + 12 F), the PORV setpoint has been chosen to limit the potential for inadvertent discharge or cycling of the PORV. Other action such as removing the power to the PORV has the same effect as raising the setpoint which also satisfies this requirement. There is no upper limit on this setpoint as the Pressurizer Safety Valves (T.S. 3.1.1.3) provide the required overpressure relief.

Below 263 F (275 F - 12 F), the PORV setpoint is reduced to provide the required low  !

temperature overpressure relief when high pressure sources and flowpaths are in service.

There is no lower limit on the pressure actuation specified as lower setpoints also provide this same protection.-

3-18e Amendment No.%449,167,186, I

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d In both cases, the 275 F 12 F setting is specified to reflect the nominal value which allows for )

normal variations in the temperature setpoint while maintaining the tolerances a::3ured in the bases l

for T.S. 3.1.2. Either pressure actuation setpoint is acceptable within the temperature rege between 2,63 F and 287 F.

During normal plant heatup and cooldown, with RCS temperatures less than 332 F and the makeup pumps running, the high pressure injection valves are closed and pressurizer level is maintained less than 220 inches to allow time for action to prevent severe overpressurization in the event of any single failure.

The PORV Block Valve is required to be OPERABLE during the HOT STANDBY, STARTUP, and POWER OPERATION in order to provide isolation of the PORV discharge line to positively control potential RCS depressurization.

For protection from severe overpressurization during HPI testing, refer to Section 4.5.2.1.c.

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3-18f Amendment No. 446

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3.3 EMERGENCY CORE COOLING, REACTOR BUILDING EMERGENCY  !

COOLING AND REACTOR BUILDING SPRAY SYSTEMS l Applicability: Applies to the cperating status of the emergency core cooling, reactor l '

building emergency cooling, and reactor building spray systems.

Obiective: To define the conditions necessary to assure immediate availability of the .

emergency core cooling, reactor building emergency cooling and reactor builder spray systems. I Specification 3.3.1 The reactor shall not be made critical unless the following editions are met:  !

3.3.1.1 Iniection Systems l

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a. The borated water storage tank shall contain a minimum of 350,000 gallons of l water having a minimum concenication of 2,500-ppm boron at a temperature j not less than 40 F. Specification 3.0.1 applies.

b. Two makeup pumps are operable in the engineered safeguards mode powered from independent essential buses. Specification 3.0.1 applies.

c. Two decay heat removal pumps are operable. Specification 3.0.1 applies.

d. Two decay heat removal coolers and their cooling water supplin are operable.

(See Specification 3.3.1.4) Specification 3.0.1 applies.

e. Two BWST level instrument channels are operable.  !

f. The two reactor building sump isolation valves (DH-V6A/6B) shall be remote-manually operable. Specification 3.0.1 applies.

3.3.1.2 Core Floodina System a.Two core flooding tanks each containing 940 30 ff ofborated water at 600 2 i 25 psig shall be available. Epecification 3.0.1 applies. I

b. Core flooding tank boron concentration shall not be less than 2,270-ppm boron.

Specification 3.3.2.1 applies.

c. The electrically operated discharge valves from the core flood tank will be assured open i by administrative control and position indication laraps on the engineerec' safeguards status panel. Respective breakers for these valves shall be open and conspicuously l nurked. Specification 3.0.1 applies.

d. One core flood tank pressure instrumentation channel and one core flood tank level instrumentation channel per tank shall be operable. Specification 3.3.2.1 applies.

3-21 l

Amendment No-2!, 98,178,203

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a. Core flood tank (CFr) vent valves CF-V3A and CF-V3B shall be closed and the breakers to the CFT vent valve motor operators shall be tagged open, except when adjusting core flood tank level and/or pressure. Specification 3.0.1 applies.

l 3.3.1.3 Reactor Buildina Sorav System and Reactor Buildina Emernency Coolina System l

The following components must be OPERABLE:

i a, Two reactor building spray pumps and their associated spray nozzles headers and l ' two reactor building emergency cooling fan- mi associated cooling units (one in each train). Specification 3.0.1 applies.

l l b. %e sodium hydroxide (NaOH) tank shall be maintained : 8 ft. M inches lower l than the BWST level as measured by the BWST/NaOH tank differential pressure indicator. The NaOH tank concentration shall be 10.015 weight peicent (%).

Specification 3.3.2.1 applies.

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c. All manual valves in the discharge lines of the sodium hydroxide tank shall be locked open. Specification 3.3.2.1 applies.

l l 3.3.1.4 Coolina Water Systems - Specification 3.0.1 applies,

a. Two nuclear service closed cycle cooling water pumps must be OPERABLE. j

b. Two nuclear service river water pumps must ta OPERABLE.

l c. Two decay heat closed cycle cooling water pumps must be OPERABLE.  ;

l d. Two decay heat river water pumps must be OPERABLE.

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e. Two reactor building emergency cooling river water pumps must, be OPER ABLE.

3.3.1.5 Engineered Safeguards Valves and Interlocks Associated with the Systems in Specifications 3.3.1.1,3.3.1.2,3.3.1.3,3.3.1.4 are OPERABLE. Specification 3.0.1 applies.

3.3.2 Maintenance or testing sha!! be allowed during reactor oper#m on any component (s) in the makeup and purification, decay heat, RB emergency cooling water, RB spray, BWST level instrumentation, or cooling water systems which will not remove more than one train of each system from service. Components shall not be removed from service so that the affected system train is inoperable for more thar 72 consecutive hours. If the system is not restored to meet the requirements of Specification 3.3.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor si.all be placed in a HOT SHUTDOWN condition within six hours.

3.3.2.1 If the CFT boron concentration is outside oflimits, or the CFT pressure instrumentation. CFT level instrumentation, or RB spray NaOH addition j system is inoperable, res; ore the system to operable natus within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If I the system is not restored to meet the requirements of Specification 3.3.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in a HOT SHUTDOWN condition i within six hours.

3 22 Amendment No. 33, *?,98,137, !?4490

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ra 4.5.2 EMERGENCY CORE COOLING SYSTEM Apolicability: , olies to periodic testing requirer".ent for emergency core cooling systems.

Obiective: To verify tha the emergency core cooling systems are operable.

Specification 4.5.2.1 Hiah Pressure Injection

a. During each refueling interval and following maintenance or modification that affects system flow characteristics, system pumps and system high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable.

b. The test will be considered satisfactory if the valves (MU-V-14A/B

& 16A/B/C/D) have completed their travel and the make-up pumps are running as evidenced by system flow. Minimum acceptable injection flow must be greater than or equal to 431 gpm per HPI pump when pump discharge pressure is 600 psig or greater (the pressure between the pump and flow limiting device) and when the RCS pressure is equal to or less than 600 psig.

c. Testing which requires HPI flow thru MU-V16A/B/C/D shall be conducted only under either of the following conditions:

1) Tavg shall be greater the 332 F.

2) Head of the Reactor Vessel shall be removed.

4.5.2.2 Low Pressure Injection

a. During each refueling period and following maintenance or modification tint affects system flow characteristics, system pumps and high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable. The auxiliaries required for low pressure injection are all included in the emergency loading sequence specified in 4.5.1.

b. The test will be considered satisfactory if the decay heat pumps listed in 4.5.1.lb have been successfully started and the decay heat injection valves and the decay heat supply valves have completed their travel as evidenced by the control board component operating lights. Flow shall be verified to be equal to or greater than the flow assumed in the Safety Analysis for the single corresponding RCS pressure used in the test.

4-41 Amendment L.-19,57,6 , ! S,203 n

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c. When the Decay Heat System is required to be operable, the

, corre position of DH-V-19A/B shall be verified by observation within four hours of each valve stroking operation or valve maintenance, which affects the position indicator.

4.5.2.3 Core Flooding

a. During each refueling period, a system test chall be conducted to demonstrate proper operation of the system.

Verification shall be made that the check and isolation valves in the core cooling floediag tank discharge lines operate properly.

b. The test will be considered satisfactory if control board indication 6.are flooding tank level verifies that all valves have opened.

4.5.2.4 Component Tests

a. At intervals not to exceed 3 months, the components required foremergency core cooling will be tested.

b. The test will be considered satisfactory if the pumps and fans have been  !

successfully started and the valves have completed their travel as evidenced by the control board component operating lights, and a second means of verification, such as: the station computer, verification ofpressure/ flow, or control board indicating lights initiated by separate limit switch contacts.

Bases De emergency core cooling systems (Reference 1) are the principal reactor safety j features in the event of a loss of coolant accident. %; removal of he 2 from the core provided by these systems is designed to hmit cor: Jamage.

The low pressure injection pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the bypass valves in the borated water storage tank fill line. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank.

The minimum acceptable HPI/LPI flow assures proper flow and flow split between j injection legs.

With the reactor shutdown, the valves in each core flooding line are checked for operability by reducing the reactor coolant system pressure until the indicated level in the core flood tanks verify that the check and isolation valves have opened.

Reference (1) UFSAR, Section 6.1 " Emergency Core Cooling System" 4-42 Amendment No.57, 5*,119,157,167 l

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4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM I

1 Applicability Applies to testing of the reactor building cooling and isolation systems.

D&icctive To verify that the reactor building cooling systems are operab!c Spacification 4.5.3.1 System Tests

a. Reactor Buildina Sorav System

1. At each refueling interval and simultaneously with the test of the emergency loading sequence, a Reactor Building 30 psi high pressure test signal will start the spray pump. Except for the spray pump suction valves, all engineered safeguards spray valves will be closed.

Water will be circulated from the borated water storage tank through 1 the reactor building spray pumps and returned through the test line to the borated water storage tank.

The operation of the spray valves will be verified during thecomponent test of the Reactor Building Cooling and Isolation System.The test will be considered satisfactory if the spray pumps have been successfully started as evidenced by the control board component operating lights, and either the station computer or pressure / flow indication.

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2. Compressed air will be introduced into the spray headers to verify each spray nozzle is unobstnicted at least every ten years.

b. Reactor Buildina Coolina and Isolation Systems

1. During each refueling period, a system test shall be conducted to

l. demonstrate proper operation of the system.

l l 2. The test will be considered satisfactory if measured system flow is greater than accident design flow rate.

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4-43 Amendment No. #4198

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Enclosure 3 Certificate of Service for TMI-l Technical Specification Change Request No. 262 l

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1 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 Technical Specification Change Request No. 262 COMMONWEALTH OF PENNSYL\ .sNIA )

) SS:

COUNTY OF DAUPHIN )

This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License for Three Mile Island Nuclear Station, Unit 1. As a part of this request, proposed replacement pages for Appendix A to the License are also included. All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are true and correct to the best of my i knowledge.

GPU NUCLEAR INC.

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BY: hlUinss)m Vicep/esidesanjirector, TMI r Sworn and Subscribgo before me this day of //h .1999.

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% AM _ <

l Notary Publi

[ [ NotarialSeal Suzanne C. Miklosik. Notary Pubhc M C mmis on di es ov 2,'9h9 Membef.Pennsyl'anta Association of Notanes k...-........- . . . . - - - . - --