ML20134M121

Proposed Tech Specs,Incorporating Certain Improvements from Revised STS for B&W plants,NUREG-1430

ML20134M121
Person / Time
Site:	Three Mile Island
Issue date:	02/07/1997
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20134M114	List: ML20134M111 ML20134M121
References
RTR-NUREG-1430 NUDOCS 9702200100
Download: ML20134M121 (11)

Text

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3,7 UNIT ELECTRIC POWER SYSTEM l

Apolicability Applies to the availability of electrical power for operation of the unit auxiliaries.

Obiective To define those conditions of electrical power availability necessary to ensure:

a. Safe unit operation

b. Continuous availability of engineered safeguards Spe;g.ificatiort 3.7.1 The reactor shall not be made critical unless all of the following requirements are satisfied:

a. All engineered safeguards buses, engineered safeguards switchgear, and engineered safeguards load shedding systems are operable.

b. One 7200 volt bus is energized.

c. Two 230 kv lines are in senice.

d. One 230 kv bus is in senices. I I

e. Engineered safeguards diesel generators are operable and at least 25,000 gallons of fuel oil are 1 available in the storage tank.

f. Station batteries are charged and in senice. Two battery chargers per battery are in senice.

3.7.2 Tie enctor shall not remain critical unless all of the following requirements are satisfied:

a. Two 230 kv lines are in senice and capable of carrying auxiliary power to Unit 1, except as specified in Specification 3.7.2c below,

b. Both 230/4.16 kv unit auxiliary transformers shall be in operation except that within a period not to exceed eight hours in duration from and after the time one Unit I auxiliary transformer is made or found inoperable, two diesel generators shall be operable, and one of the operable diesel generators will be started and run continuously until both unit auxiliary transformers are in operation. This mode of operation may continue for a period not exceeding 30 days. l

c. Both diesel generators shall be operable except that from the date that one of the diesel generators is made or found to be inoperable 3-42 Amendment No.188' 9702200100 970207 yDR ADOCK 05000289 PDR

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Ior any reason, reactor operation is permissible for the succeeding seven days provided that the l redundant diesel generator is:  ;

1. verified to be operable immediately; . [

2. within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> either:  !

a. determine the redundant diesel generator is not inoperable due to a common .

mode failure or . 'I t

b. test redundant dicscl generator in accordance with surveillance requirement i 4.6.1.a. '

i In the event two diesel generators are inoperable, the unit shall be placed in hot shutdown in 12  !

hours. If one diesel is not operable within an additional 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period the pla. t shall be placed >

in cold shutdown within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereaAct. l l

With one dicsci generator inoperable, in addition to the above, verify that: All required  !

systems, subsystems, trains, components and devices that depend on the remaining OPERABLE l diesel generator as a source of emergency power are also OPERABLE or follow specifications l 3.0.1.  ;

d. If one Unit AExiliary Transformer is inoperable and a diesel generator becomes inoperable, the l  !

unit will be placed in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If one of the above sources of power is not  ;

made operable within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the unit shall be placed in cold shutdown within  ;

an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereaAer. I

c. If Unit 1 is separated from the system while carrying its own auxiliaries, or if only one 230 kv .

line is in senice, continued reactor operation is permissible pro ided one emergency dicsci J generator shall be started and run continuously until two transmission lines are restored.

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f. The engineered safeguards electrical bus, switchgear, load shedding,'and automatic diesel start _

systems shall be operable except as provided in Specification 3.7.2c above and as required for testing.

g. One station battesy may be removed from senice for not more than eight hours.

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. 3-43 a .

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Amendment No.98,4as

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TABLE 4.1-1 (Continued) i CALIBRATE REMARKS i CHANNEL DESCRIIrflON CHECK TEST -

27. Makeup Tank Level Channels D(1) NA F .(1) When Makeup and Purification Symem is in  !

operation.

28. Radiation Monitoring Systems

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a. RM-G6(FilBridge #1 Aux) W(1)(2) M(2) Q(2) (1) Using the installed check source when i

b. RM-G7 (FH Bridge #2 Main) W(IX2) M(2) Q(2) background isless than twice the  ;

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' expected increase in cpm which would

c. RM-G9 (FH Bridge-FH Bldg) W(IX3) M(3) E(3) result from the check source alone. }

d. RM-A2P(RB Atmospheric W(IX4) M(4) E(4) Background readings greater than this  !

e. RM-A21(RB Atmosphericiodine) W(IX4) M(4) E(4) value are sufficient in themsches to

f. RM-A2G(RB Atmosphericgas) W(1X4) M(4) E(4) to show that the monitor is functioning.

(2) RM-G6 and RM-G7 operability requirements are given in T.S. 3.8.1. Surveillances are required to be current only when handling irradiated fuel.

(3) RM-G9 operability requirements are given in T.S. .

3.8.1.

(4) RM-A2 operability requirements are given in T.S.

3.1.6.8.  ;

29. High and Low pressure N/A N/A F injection Sysems: -

Flow Channels 3 l i

• Includes only the monitors indicated under this item. Other T.S. required radiation monitors are included in specifications 3.5.5.2,4.1.3, Table 3.5-1 item C.3.f, and I Table 4.1-1 item 19e. i F

I Page 4-Sa Amendment No. 24,73,100,103,4%, M1, M2,145,197

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9 TABLE 4.1-1 (Continued)

CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS -

30. Borated Water Storage W NA F Tank LevelIndicator

31. Boric Acid Mix Tank

a. Level Channel NA NA F ,

b. Temperature Channel M NA F

32. Reclaimed Boric Acid Storage Tank

a. Level Channel NA NA F

b. Temperature Channel M NA F

33. Containment Temperature NA NA F

34. Incore Neutron Detectors M(1) NA NA (1) Check functioning; including functioning of computer '

readout or recorder readout when reactor pmver is greater than 15% l ,

35. Emergency Plant Radiation P.I(l) NA F (1) Battery check.

Instruments I

36. (DELETED) l NA R  ;

37 Reactor Building Sump NA Level  :

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Amendment No. N5  ;

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4.4.4 Hydronen Recombiner System Apolicability Applies to the testing of the hydrogen recombiner and associated controls.

Obiective To verify that the hydrogen recombiner and associated controls are operable. l 4.4.4.1 Specification

a. Perform a system functional test for each hydrogen recombiner each refueling interval as follows:

(1) Verify that the minimum heater sheath temperature increases to 2 700 F in 5 90 minutes.

(2) After reaching 700'F, increase the power to maximum for approximately 2 minutes and verify the power to be 2 60kW.

b. Visually examine the hydrogen recombiner enclosure and verify there is no evidence of 1 abnormal conditions each refueling interval.  !

c. Perform a resistance to ground test for each heater phase each refueling interval and verify that the resistance to ground for any heater is 210,000 ohms.

Bases The surveillance program described above provides high assurance that the hydrogen recombiner system will be available to perform its port-LOCA function of maintaining the containment hydrogen concentration below 4.1 volume percent. This system is not credited to mitigate any accident analyzed in Chapter 14 of the TMI-l FSAR. The frequency of the surveillance of the hydrogen recombiner system is based on the safety significance of the system. TMI I FS AR Section 6.5.3.1 indicates that the hydrogen recombiner system is not required until 9.0 days'following a LOCA. This is adequate time to place a l l hydrogen recombiner in service.

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4-38 (Page 4-38a deleted)

Amendment No.87,#8,MG , M8 ,

4.5 EMERGENGY LOADING SEOUENCE AND POWER TRANSFER. EMERGENCY CORE C_OOLING SYSTEM & REACTOR BUILDING COOLING SYSTEM PERIODIC TESTIND 4.5.1 Emergency Loading Sequence A_pplicability: Apphes to penodic testing requirements for safety actuation systems.

Obiective: To verify that the emergency loading sequence and automatic power transfer is operable.

r Specifications

1 i' 4.5.1.1 Seauence and Power Transfer Test

a. During each refueling interval, a test shall be conducted to demonstrate that the emergency loading sequence and power transfer is operable.

bL The test will be considered satisfactory if the following pumps and fans have been successfully started and the following valves have completed their travel on preferred l power and transferred to the emergency power. I

-M. U. Pump l

-D. H. Pump and D. H. Injection Valves and D. H. Supply Valves l

-R. B. Cooling Pump

-R. B. Ventilators

-D. H. Closed Cycle Cooling Pump

-N. S. Closed Cycle Cooling Pump l -D. H. River Cooling Pump. t l -N. S. River Cooling Pump ,

l -D. H. and N. S. Pump Area Cooling Fan -

-Screen House Area Cooling Fan l -Spray Pump. (Initiated in coincidence with a 2 out of 3 R. B.

30 psig Pressure Test Signal.)

-Motor Driven Emergency Feedwater Pump

c. Following successful transfer to the emergency diesel, the diesel generator breaker will be opened to simulate trip of the generator then re-closed to verify block load on the

! reclosure.

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4.5.1.2 c Seauence Test j

a. At intervals not to exceed 3 months, a test shall be conducted to demonstrate that the emergency loading sequence is operable, this test shall be performed on either preferred '

l power or emergency power.

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b. The test will be considered satisfactory if the pumps and fans listed in 4.5.1.lb have been successfully started and the valves listed in 4.5.1.lb have completed their travel. l

! 4-39 Amendment No. -70,78. 449, M7 i~

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'4 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM Aoolicability -

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

Objective ,

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To verify that the reactor building cooling systems are operable.

Specification 4.5.3.1 System Tests ,

a. Reactor Buildina Sorav System

1. At each refueling interval, a reactor building 30 psi high pressure test signal l will start the spray pump. Except for the spray pump suction valves, all I j engineered safeguards spray valves will be closed.

Water will be circulated from the borated water storage tank through 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 the component test of the R. B. cooling and isolation system.

The test will be considered satisfactory if the spray pumps have been successfully started. .

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2. Compressed air will be introduced into the spray headers to verify cach spray '

nozzle is unobstructed at least every ten years. '  ;

b. Reactor Buildina Coolina and Isolation Systems

1. During each refueling period, a system test shall be conducted to demonstrate proper operation of the system. A test signal will actuate the Reactor Buildmg Emergency Cooling System valves to demonstrate operability of the coolers.

2. The test will be considered satisfactory if the valves have completed their I expected travel.

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'l 4-43 Amendment No. M7,198 i

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l 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS i

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l 5-8 Revised (1-20-16)

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TABLE OF CONTENTS

Section Eagg 4

5 DESIGN FEATURES 5 4 5.1 SIIS 5-1 l 5.2 CONTAINMENT 5-2  !

! -5.2.1 REACTOR BUILDING 5-2  !

5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3  !

5.3 REACTOR 5-4 -

{- 5.3.1 REACTOR CORE 5-4 , j 5.3.2 REACTOR COOLANT SYSTEM 5-4 l 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 l 5.4.1 NEW FUEL STORAGE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 e i 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS (DELETED) 5-8 r 6 ADMINISTRATIVE CONTROLS 6-1 I 3

i 6.1 RESPONSIBILITY 6-1 6.2 . ORGANIZATION 6-1  !

l' 6.2.1 CORPORATE 6-1 6.2.2 UNIT STAFF 6-1  !

6.3 UNIT STAFF OUALIFICATIONS 6-3  !

6.4 TRAINING 6  !

i 6.5 . REVIEW AND AUDIT- 6-3 1 l 6.5.1 TECHNICAL REVIEW AND CONTROL 6-4  !

6.5.2 INDEPENDENT SAFETY REVIEW 6-5 I 6.5.3 AUDITS 6-7 l

} 6.5.4 INDEPENDENT ONSITE SAFETY REVIEW GROUP 6-8  !

6.6 REPORTABLE EVENT ACTION 6-10  !

6.7 SAFETY LIMIT VIOLATION 6-10 t

, 6.8 PROCEDURES AND PROGRAMS 6-11 a 6.9 REPORTING REOUIREMENTS 6-12 6.9.1 ROUTINE REPORTS 6-12 i 6.9.2 DELETED 6-14  ;

j 6.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6-17 l 6.9.4 ' ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6-18  !

6.9.5 CORE OPERATING LIMITS REPORT 6-19 l 6.10 BECORD RETENTION 6-20  !

I 6.11 RADIATION PROTECTION PROGRAM 6-22 6.12 HIGH RADIATION AREA 6-22 6.13 PROCESS CONTROL PROGRAM 6-23 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6-24  ;

6.15 DELETED 6-24 6.16 POST ACCIDENT SAMPLING PROGRAMS 6-24 l NUREG 0737 (II.B.3. II F.1.2) l 6.17 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-25 I v

Amendment No !!,",72, "'?, !29, !$0,.-4-7h

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5-9 Revised (1-20-76)

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- 6 3 UNIT STAFF OUALIFICATIONS i

l~ 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI /ANS j 3.1 of 1978 for comparable positions unless otherwise noted in the Technical Specifications.

,. Licensed operators shall also meet the requirements of 10 CFR Part 35, Individuals who do l not meet ANSI /ANS 3.1 of 1978, Section 4.5, are not cons.Jered technicians or maintenance i personnel for purposes of determmmg qualifications but are permitted to perform work for

! which qualification has been demonstrated.

'6.3.2 The management position responsible for radiological controls shall meet or exceed the qualifications of Regulatory Guide 1.8 of 1977. Each radiological controls i technician / supervisor shall meet or exceed the qualifications of ANSI-N 18.1-1971, i paragraph 4.5.2/4.3.2, or be formally qualified through an NRC approved TMl-1 Radiation i Controls training program. All radiological controls technicians will be qualified through j training and exammation in each area or specific task related to their radiological controls functions prior to their performance of those tasks.

6.3.3 The Shift Technical Advisors shall have a bachelor's degree or equivalent in a scientific c'r .

engineering discipline with specific trauung in unit design, response and analysis of transients and accidents. .

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6.4 TRAINING .

j 6.4.1 A retraining and replacement training program for the unit staff shall be mamtained under the direction of the plant trammg manager and shall meet or exceed the requirements and reco.==Mions of Regulatory Guide 1.8 of 1977. Licensed operator training shall also meet the requirements of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintamed and shall meet or exceed the requirements of Section 600 of the NFPA Code. l 6.5 REVIEW AND AUDIT

- 6.51 TECHNICAL REVIEW AND CONTROL The Vice President of each division within GPU Nuclear Corporation shall be responsible for ensuring the preparation, review, and approval of documents required by the activities described in 6.5.1.1 through 6.5.1.5 within his functional area of responsibility as assigned in the GPUN Resiew and Approval Matrix.

Implementing approvals shall be performed at the cognizant manager level or above.

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. Amendment Nos. M ,M M .M ,93.M8.44.44 M9. 444

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