Text

Proposed Improved Ts,Rev E
	ML20073G963
Person / Time
Site:	Hatch
Issue date:	09/23/1994
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20073G953	List: ML20073G951; ML20073G963;
References
	NUDOCS 9410050082
	Download: ML20073G963 (281)
	v • d • e

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l

] UNIT I IMPROVED TECHNICAL SPECIFICATIONS O

O 9410050082 940923 PDR ADOCK 05000321 P PDR

I Definitions l 1.1 >

1.1 Definitions !

O OPERABLE - OPERABILITY instrumentation, controls, normal or emergency I

(continued) electrical power, cooling and seal water, :

lubrication, and other auxiliary equipment that !

are required for the system, subsystem, division, j component, or device to perform its specified i safety function (s) are also capable of performing !

their related support function (s). '

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to i measure the fundamental nuclear characteristics of !

the reactor core and related instrumentation. t These tests are:

a. Described in Section 13.6, Startup and Power I Test Program, of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or j

c. Otherwise approved by the Nuclear Regulatory ;

Commission.

l t

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, O !

t i

RATED THERMAL POWER RTP shall be a total reactor core heat transfer f (RTP) rate to the reactor coolant of 2436 MWt. !

i REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS i TIME trip setpoint at the channel sensor until l de-energization of the scram pilot valve i solenoids. The response time may be measured by l means of any series of sequential, overlapping, or i total steps so that the entire response time is measured. !

O (continued) i HATCH UNIT 1 1.1-5 REVISION /I

Definitions 1.1 1.1 Definitions (continued)

SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:

a. The reactor is xenon free;

b. The moderator temperature is 68 F; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME consists RESPONSE TIME of two components:

a. The time from initial movement of the main turbine stop valve or control valve until 80%

of the turbine bypass capacity is established; and

b. The time from initial movement of the main

, turbine stop valve or control valve until initial movement of the turbine bypass valve.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

O HATCH UNIT 1 1.1-6 REVISION A

Feedwater and Main Turbine Trip High Water Level Instrumentation 3.3.2.2

[ SURVEILLANCE REQUIREMENTS

_------------------------------------NOTE-------------------------------------

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided feedwater and main turbine high '

water level trip capability is maintained.

SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL FUNCTIONAL TEST. 92 days I

SR 3.3.2.2.2 Perform CHANNEL CALIBRATION. The 18 months Allowable Value shall be s 56.5 inches.

i i

i SR 3.3.2.2.3 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including valve actuation.

O O

HATCH UNIT 1 3.3-21 REVISION A

PAM Instrumentation 3.3.3.1 3.3 INSTRUMENTATION 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation LC0 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1-1 shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

NOTES------------------------------------

1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more functions A.1 Restore required 30 days with one required channel to OPERABLE channel inoperable. status.

B. Required Action and B.1 Initiate action in Immediately associated Completion accordance with Time of Condition A Specification 5.6.6. l not met.

C. One or more Functions C.1 Restore all but one 7 days with two or more required channel to required channels OPERABLE status.

inoperable.

(continued)

O HATCH UNIT 1 3.3-22 REVISION

PAM Instrumentation 3.3.3.1 3

/ ACTIONS (continued)

\_.)*

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C Table 3.3.3.1-1 for not met. the channel.

E. As required by E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action D.1 and referenced in Table 3.3.3.1-1.

F. As required by F.1 Initiate action in Immediately Required Action D.1 accordance with and referenced in Specification 5.6.6. l Table 3.3.3.1-1.

O)

L SURVEILLANCE REQUIREMENTS

__-----------------------------------NOTES------------------------------------

1. These SRs apply to each Function in Table 3.3.3.1-1.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the other required channel (s) in the associated Function is OPERABLE.

SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK. 31 days SR 3.3.3.1.2 Perform CHANNEL CALIBRATION. 18 months l

(~h L)

HATCH UNIT 1 3.3-23 I REVISIONf[

PAM Instrumentation 3.3.3.1 Table 3.3.3.1-1 (page 1 of 1)

Post Accident A nitoring InstrLnentation CONDITIONS REFERENCED REQUIRED FROM REQUIRED FUNCTION CHANNELS ACTION D.1

1. Reactor Steam Dome Pressure 2 E

2. Reactor Vessel Water Level

a. -317 inches to -17 inches 2 E

b. -150 inches to +60 inches 2 E

c. O inches to +60 inches 2 E

d. O inches to +400 inches 1 NA

3. Suppression Pool Water Leval

a. O inches to 300 inches 2 E

b. 133 inches to 163 inches 2 E

4. Drywell Pressure

a. -10 psig to +90 psig 2 E

b. -5 psig to +5 psig 2 E

c. O psig to +250 psis 2 E

5. Dcywell Area Radiation (High Range) 2 F

6. Primary containment Isolation Valve Position 2 per t[,ag g flow E

7. Drywell H, Concentration 2 E

8. Drywell 0, Concentration 2 E

9. Suppression Pool Water Tenperature 2(C) E

10. Drywell Temperature in Vicinity of Reactor Level 6 J Instrtment Reference Leg

11. Diesel Generator (DG) Parameters

a. Output Voltage 1 per DG kA

b. Output Current 1 per DG NA

c. Output Power 1 per DG NA

d. Battery Voltage 1 per DG NA

12. RHR service Water Flow 2 E (a) Not required for isolation v.=Lves whose associated penetration flow path is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

(b) only one position indication channel is required for penetration flow paths with only one installed control room indication channel.

(c) Monitoring each of four quadrants.

O HATCH UNIT 1 3.3-24 REVISION A

ECCS Instrumentation 3.3.5.1

[D

%.)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. As required by G.1 Declare ADS valves I hour from Required Action A.1 inoperable. discovery of and referenced in loss of ADS Table 3.3.5.1-1. initiation capability in both trip systems A!!2 G.2 Restore channel to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from OPERABLE status. discovery of inoperable channel concurrent with HPCI or RCIC inoperable AND 8 days

(']s L

H. Required Action and H.1 Declare associated Immediately associated Completion supported feature (s)

Time of Condition B, inoperable.

C, D, E, F, or G not met.

'(--)

HATCH UNIT 1 3.3-37 REVISION A

ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS

_____________------------------------NOTES------------------------------------

1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 3.c and 3.f; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions other than 3.c and 3.f provided the associated Function or the redundant function maintains l initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.5.1.3 Perform CHANNEL CALIBRATION. 92 days SR 3.3.5.1.4 Perform CHANNEL CALIBRATION. 18 months SR 3.3.5.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months O

HATCH UNIT 1 3.3-38 REVISIONp'[

LOP Instrumentation 3.3.8.1 p)

U 3.3 INSTRUMENTATION 3.3.8.1 Loss of Power (LOP) Instrumentation LC0 3.3.8.1 The LOP instrumentation for each Function in Table 3.3.8.1-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, When the associated diesel generator (DG) is required to be OPERABLE by LC0 3.8.2, "AC Sources -- Shutdown."

ACTIONS

NOTE-------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Restore channel to I hour

x. inoperable for OPERABLE status.

Functions 1 and 2.

B. One or more channels B.1 Verify voltage on Once per hour l inoperable for associated 4.16 kV Function 3. bus is 1 3825 V.

C. Required Action and C.1 Declare associated DG Immediately l associated Completion inoperable.

Time not met.

HATCH UNIT 1 3.3-67 REVISION C

LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS

NOTE-------------------------------------

1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.

2. When a 4.16 kV Emergency Bus Undervoltage channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains initiation capability l (for Functions 1 and 2) and annunciation capability (for Function 3).

SURVEILLANCE FREQUENCY SR 3.3.8.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.8.1.2 Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.8.1.3 Perform CHANNEL CALIBRATION. 18 months h

SR 3.3.8.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months O

HATCH UNIT 1 3.3-68 REVISION i

i

i LOP Instrumentation 3.3.8.1

/ Table 3.3.8.1 1 (page 1 of 1)

(y Loss of Power Instrumentation REQUIRED CHANNELS SURVEILLANCE ALLOWABLE FUNCTION PER BUS REQUIREMENTS VALUE i

1. 4.16 kV Emergency Bus undervoltage (Loss of Voltage) i

s. Bus Undervoltage 2 SR 3.3.8.1.2 2 2800 V SR 3.3.8.1.3 SR 3.3.8.1.4

b. Time Delay 2 SR 3.3.8.1.2 SR 3.3.8.1.3 s 6.5 seconds SR 3.3.8.1.4

2. 4.16 kV Emergency Bus undervoltage (Degraded Voltage)

a. Bus Undervoltage 2 SR 3.3.8.1.2 2 3280 V SR 3.3.8.1.3 SR 3.3.8.1.4

b. Time Delay 2 SR 3.3.8.1.2 SR 3.3.8.1.3 s 21.5 seconds ,

SR 3.3.8.1.4

3. 4.16 kV Emergency Bus undervoltage (Annunciation)

a. Bus undervottage 2 SR 3.3.8.1.1 a 3825 V l

\ SR 3.3.8.1.2 SR 3.3.8.1.3 SR 3.3.8.1.4

b. Time Delay 2 SR 3.3.8.1.2 s 60 seconds l SR 3.3.8.1.3 SR 3.3.8.1.4 1

(

/e HATCH UNIT 1 3.3-68A REVISIONCg J

3 RCS P/T Limits 3.4.9

[ ) 3.4 REACTOR COOLANT SYSTEM (RCS)

%J 3.4.9 RCS Pressure and Temperature (P/T) Limits LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within limits.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1 Restore parameter (s) 30 minutes Required Action A.2 to within limits.

shall be completed if this Condition is AND 1 entered.

A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months O acceptable for

'\j Requirements of the continued operation.

LC0 not met in MODES 1, 2, and 3.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

O HATCH UNIT 1 3.4-21 REVISIONf[

l RCS P/T Limits 3.4.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE--------- C.1 Initiate action to Immediately Required Action C.2 restore parameter (s) shall be completed if to within limits.

this Condition is entered. AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LC0 not met in other operation. or 3 l -

than MODES 1, 2, and 3.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 -------------------NOTE--------------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify:

30 minutes

a. RCS pressure and RCS temperature are within the limits specified in Figure 3.4.9-1 and Figure 3.4.9-2; and

b. RCS heatup and cooldown rates are s 100 F in any I hour period.

l (continued)

O HATCH UNIT 1 3.4-22 REVISION A

/L i

I RCS P/T Limits l' 3.4.9 SURVEILLANCE REQUIREMENTS (contirued)

SURVEILLANCE FREQUENCY !

l SR 3.4.9.2 Verify RCS pressure and RCS temperature are Once within l within the criticality limits specified in 15 minutes j Figure 3.4.9-3. prior to l :

control rod !

withdrawal for ' i the purpose of achieving ,

criticality l (continued) !

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i O !I HATCH UNIT 1 3.4-)f St.1 A REVISION / E 4 i

l RCS P/T Limits j 3.4.9 !

SURVEILLANCE REQUIREMENTS (continued) ,

i SURVEILLANCE FREQUENCY l SR 3.4.9.3 --------------------NOTE------------------- !

Only required to be met in MODES 1, 2, 3, !

and 4 during startup of a recirculation !

pump. !

Verify the difference between the bottom 15 minutes !

head coolant temperature and the reactor !

pressure vessel (RPV) coolant temperature i is s 145 F. l j i

SR 3.4.9.4 -------------------NOTE-------------------- i Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation l pump.

Verify the difference between the reactor 15 minutes !

coolant temperature in the recirculation loop to be started and the RPV coolant i temperature is s 50 F. l t

l SR 3.4.9.5 -------------------NOTE-------------------- ;

Only required to be performed when !

tensioning the reactor vessel head bolting !

studs. .I Verify reactor vessel flange and head 30 minutes flange- temperatures are 2 76 F. l l

(continued) 22 HATCH UNIT 1 REVISION [g 3.4-2[

2 RCS P/T Limits 3.4.9 SURVEll. LANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.6 -------------------NOTE--------------------

Not required to be performed until 30 minutes after RCS temperature s 86*F in l MODE 4.

Verify reactor vessel flange and head 30 minutes flange temperatures are 2 76*F. l SR 3.4.9.7 -------------------NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature s 106*F in MODE 4. l Verify reactor vessel flange and head 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months flange temperatures are 2 76*F. l 9

O HATCH UNIT 1 3.4-282 _ REVISION

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1 RCS P/T Limits !

3.4.9 j

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E FPY 8 10 12 14 16 i

1200 ADJUSTED CORE BELTLINE, 1/4 T FLAW O 1

$ 1000 i E ,

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a.

K z 800 i

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U 5 a s00 A VERTICAL LIMIT LINE FOR PRESSURd A90VE 20% HYDROTEST (312 poet, 400 BASED ON 10CFR50 APPENDIX G REQUIREMENT OF (RTNDT + 90eF), l i

FLANGE REGION RTND7 = 160F SOLT PRELOAD TEMPERATURE OF ,

i 760F BASED ON RECOMMENDED 200 , (RTuor + 60eF) FOR 0.24-IN. FLAW IN CLOSURE FLANGE REGION, RTNDT = 180F 0

0 100 200 300 RPV METAL TEMPERATURE (oF)

Figure 3.4.9-1 (page 1 of 1)

Temperature / Pressure Limits for Inservice Hydrostatic and Inservice Leakage Tests HATCH UNIT 1 3.4- E S d k REV1SION/E

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l RCS P/T Limits 3.4.9 i

O l 1600 ; j VALID TO 16 EFFECTIVE FULL POWER YE ARS OF OPERATION 1400 g,

1200 O

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ADJUSTED CORE BELTLINE UJ 1/4 T FLAW, RTNDT = 108F 3 IRRADIATION SHIFT = 123*F

$ 800 -

E iy 600 9

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la E 400 FEEDWATER NOZZLE TEMPERATURE LIMIT FOR 1/4 T FLAW (BWR/6 0 '

RESULTS ADJUSTED TO 40 F RTNDTI 200 MINIMUM OPERATING TEMPERATUNE

/

/ OF 76*F BASED ON RECOMMENDED (RTNOT + 60'F) FOR 0.24-IN. FLAW IN CLOSURE FLANGE REGION, RTNDT = 15'F f 0 I I O 100 200 300 400 500 600 MINIMUM VESSEL METAL TEMPERATURE (OF)

Figure 3.4.9-2 (page 1 of 1)

Temperature / Pressure Limits for Non-Nuclear Heatup, -

Low Power Physics Tests, and Cooldown Following a Shutdown HATCH UNIT 1 3.4-J/Sd OJ REVISION [h i

RCS P/T Limits 3.4.9 O -

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VAllD TO 16 EFFECTIVE FULL POWER YEARS OF OPERATION 1400 l

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f 1200 ,

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IRRADIATION swift = ias'F ,

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400 l l FEEDWATER NOZZLE TEMPE RATURE LIMIT FOR 1/4 T FLAW (BWR/6 RESULTS ADJUSTED TO 40*F RTNOTI k -

200 l j MINIMUM OPERATING TEMPERATURE LIMIT OF 76'F FROM 10CFR50 APPENDIX G I REQUIREMENT THAT (Tugg = RTNDT + 80*F), j

% rLANcE RTuor is*F , j 0 100 200 300 400 500 600 l l

MINIMUM VESSEL METAL TEMPERATURE (DF) l I I a l Figure 3.4.9-3 (page 1 of 1)

Temperature / Pressure Limits for Criticality O

HATCH UNIT 1 3.4-K c% REVISION /f

Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCO 3.4.10 The reactor steam dome pressure shall be s 1020 psig, i

APPLICABILITY: MODES 1 and 2. l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within dome pressure to limit. within limit.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify rea. tor steam dome pressure is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s 1020 psig.

O HATCH VNIT 1 3.4-25 REVISION A

SGT System l

3.6.4.3

\

O SURVEILLANCE FREQUENCY l SR 3.6.4.3.1 Operate each required Unit I and Unit 2 31 days SGT subsystem for 2: 10 continuous hours with heaters operating.

j SR 3.6.4.3.2 Perform required SGT filter testing in In accordance l accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).

SR 3.6.4.3.3 Verify each required SGT subsystem 18 months actuates on an actual or simulated initiation signal.

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O HATCH UNIT 1 3.6-45 REVISIONh

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Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Ventilation Filter Testina Procram (VFTP)

The VFTP will establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2. Section 5a and at least once per 18 months or 1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, 2) following painting, fire or chemical release in any ventilation zone communicating with the system, or 3) after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation.

-- ---NOTES-----------------------------

1. Tests and evaluations have determined the impact on the Standby Gas Treatment (SGT) System filters of certain types of painting, buffing and grinding, and welding. The use of water based paints and the performance of metal grinding, buffing, or welding 'are not detrimental to the charcoal filters of the SGT System, either prior to or during operation. These activities will not require surveillance of the system upon their conclusion. This applies to all types of welding conducted at Plant Hatch, and tracking of the quantity of weld material used is not necessary.

2. For testing purposes, the use of refrigerants equivalent to those specified in ASME N510-1989 is acceptable.

a. Demonstrate for each of the ESF systems that an inplace test of the HEPA filters shows a penetration and system bypass

< 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, Section 5c and ASME N510-1989, Section 10, at the system flowrate specified below.

ESF Ventilation System Flowrate (cfm)

SGT System 3000 to 4000 Main Control Room Environmental 2250 to 2750 Control (MCREC) System i I

l (continued)

HATCH UNIT 1 5.0-11 REVISION A

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testino Praaram (VFTP) (continued)

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, Section 5d and ASME N510-1989, ,

Section 11, at the system flowrate specified below. '

ESF Ventilation System Flowrate (cfm)

SGT System 3000 to 4000 MCREC System 2250 to 2750

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, Section 6b and ASME N510-1989, Section 15 and Appendix B, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of s 30*C and greater than or equal to the relative humidity specified below.

ESF Ventilation System Penetration (%) RH(%)

SGT System 0.2 70 MCREC System 2.0 95

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and <

the charcoal adsorbers is less than the value specified '

below when tested in accordance with ASME N510-1989, Section 8.5.1, at the system flowrate specified below.

ESF Ventilation System 6P (inches wa) Flowrate (cfm)

SGT System <6 3000 to 4000 MCREC System <6 2250 to 2750

e. Demonstrate that the heaters for the ESF system dissipate the value specified below when tested in accordance with l ASME N510-1989, Section 14.5.1. l ESF Ventilation System Wattaae (kW)

SGT System 15 to 20 (continued)

HATCH UNIT 1 5.0-12 REVISION h

Programs and Manuals '

5.5 f 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testina Proaram (VFTP) (continued)

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

1 O

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(continued)

HATCH UNIT 1 5.0 ,14 / 2 A REVISION ,A' b

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1) Control Rod Block Instrumentation - Rod Block Monitor for Specification 3.3.2.1.

2) The Average Planar Linear Heat Generation Rate for Specification 3.2.1.

3) The Minimum Critical Power Ratio for Specifications 3.2.2 and 3.3.2.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1) NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel," (applicable amendment specified in the COLR).

k 2) " Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.157 to Facility Operating License DPR-57," dated September 12, 1988.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits and accident analysis limits) of the safety ,

analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

l l

(continued)

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HATCH UNIT 1 5.0-4lh REVIS10Nh

Reporting Requirements 5.6 l

5.6 Reporting Requirements (continued) 5.6.6 Post Accident Monitorina (PAM) Instrumentation Report l When a report is required by LC0 3.3.3.1, " Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the function to OPERABLE status.

O HATCH UNIT 1 REVISION 5.0-2)'O

UNIT 1 IMPROVED BASES i

I l

O P

O

l

)

PAM Instrumentation B 3.3.3.1 h

wJ BASES LC0 12. RHR Service Water Flow (continued) primary indication used by the operator during an accident.

Therefore, the PAM specification deals specifically with this portion of the instrument channel. ,

APPLICABILITY The PAM instrumentation LC0 is applicable in MODES 1 and 2.

These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2. In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES.

ACTIONS Note 1 has been added to the ACTIONS to exclude the MODE change restriction of LC0 3.0.4. This exception allows entry into the applicable MODE while relying on the ACTIONS

n even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the passive (v) function of the instruments, the operator's ability to diagnose an accident using alternative instruments and methods, and the low probability of an event requiring these instruments.

Note 2 has been provided to modify the ACTIONS related to PAM instrumentation channels. Section I.3, Completion Times, specifies that once a Condition has been entered, '

subsequent divisions, subsystems, components, or variables ,

expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required ;

Actions of the Condition continue to apply for each '

additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable PAM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable PAM Function.

t (continued)

HATCH UNIT 1 B 3.3-67 I REVISION A

PAM Instrumentation B 3.3.3.1 BASES g ACTIONS A.1 (continued)

When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining 0PERABLE channels (or, in the case of a Function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), tne passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

8.1 If a channel has not been restored to OPERABLE status in 30 days, this Required Action specifies initiation of action in accordance with Specification 5.6.6, which requires a l written report to be submi_tted to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions.

, This action is appropriate in lieu of a shutdown

( requirement, since alternative actions are identified before loss of functional capability, and given the likelihood of plant conditions that would require information provided by this instrumentation.

L1 When one or more Functions have two or more required channels that are inoperable (i.e., two channels inoperable in the same Function), all but one channel in the Function should be restored to OPERABLE status within 7 days. The l Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information. Continuous cperation with two required channels inoperable in a Function is not acceptable i

because the alternate indications may not fully meet all performance qualification requirements applied to the PAM f instrumentation. Therefore, requiring restoration of one l

inoperable channel of the Function limits the risk that the (continued)

B 3.3-68 REVISION HATCH UNIT 1 1

PAM Instrumentation B 3.3.3.1 BASES ACTIONS L1 (continued)

PAM Function will be in a degraded condition shotid an accident occur.

L.1 This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. The applicable Condition referenced in the Table is function dependent.

Each time an inoperable channel has not met the Required Action of Condition C, and the associated Completion Time has expired, Condition D is entered for that channel and provides for transfer to the appropriate subsequent Condition.

L.1 For the majority of Functions in Table 3.3.3.1-1, if any Required Action and associated Completion Time of Condition C is not met, the plant must be brought to a MODE O- in which the LC0 not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

f.d Since alternate means of monitoring drywell area radiation have been developed and tested, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6. These alternate means may be l temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the

-areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.

O-. (continued)

HATCH UNIT I B 3.3-69 REVISION

i PAM Instrumentation B 3.3.3.1 BASES (continued) h SURVEILLANCE As noted at the beginning of the SRs, the following SRs REQUIREMENTS apply to each PAM instrumentation Function in Table 3.3.3.1-1. ;

The Surveillances are modified by a second Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the other required channel (s) in the associated Function are OPERABLE. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.

The Note is based upon a NRC Safety Evaluation Report (Reference 1) which concluded that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability of properly monitoring post accident parameters, when necessary.

SR 3.3.3.1.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross enannel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

(continued)

HATCH UNIT 1 B 3.3-70 REVISION A

ECCS Instrumentation B 3.3.5.1 BASES ACTIONS F.1 and F.2 (continued) ,

it is not desired to place the channel in trip (e.g., as in '

the case where placing the inoperable channel in trip would result in an initiation), Condition H must be entered and its Required Action taken.

G.1 and G.2 Required Action G.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within similar ADS trip system Functions result in automatic initiation capability being lost for the ADS. In this situation (loss of automatic initiation capability), the 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months or 8 day allowance, as applicable, of Required Action G.2 is not appro)riate, and all ADS valves must be declared inoperable wit 11n I hour after discovery of loss of ADS initiation capability. l The Completion Time is intended to allow the operator time -

to evaluate and repair any discovered inoperabilities. This t Completion Time also allows for an exception to the normal '

C " time zero" for beginning the allowed outage time " clock."

For Required Action G.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable channels within similar ADS trip system Functions as described in the paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be ,

acceptable (Ref. 5) to permit restoration of any inoperable '

channel to OPERABLE status if both HPCI and RCIC are OPERABLE (Required Action G.2). If either HPCI or RCIC is inoperable, the time shortens to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . If the status of :

HPCI or RCIC changes such that the Completion Time changes ,

from 8 days to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , the 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months begins upon discovery of HPCI or RCIC inoperability. However, the total time for.

an inoperable channel cannot exceed 8 days. If the status of HPCI or RCIC changes such that the Completion Time changes from 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months to 8 days, the " time zero" for beginning the 8 day " clock" begins upon discovery of the (continued)

HATCH UNIT I B 3.3-129 REVISION A

l ECCS Instrumentation B 3.3.5.1 BASES ACTIONS G.1 and G.2 (continued) inoperable channel . If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events, lid With any Required Action and associated Completion Time not met, the associated feature (s) may be incapable of performing the intended function, and the supported feature (s) associated with inoperable untripped channels must be declared inoperable immediately.

SURVEILLANCE As noted in the beginning of the SRs, the SRs for each ECCS REQUIREMENTS instrumentation Function are found in the SRs column of Table 3.3.5.1-1.

The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months as follows: (a) for Functions 3.c and 3.f; and (b) for Functions other than 3.c and 3.f provided the associated Function or the redundant Function maintains l initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 5) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary.

(continued)

HATCH UNIT 1 B 3.3-130 REVISION,Af

Primary Containment Isolation Instrumentation .

B 3.3.6.1 ;

l BASES l

1. I BACKGROUND Main Steam Line Isolation (continued) i MSL Isolation Functions isolate the Group 1 valves. ;

2. Primary Containment Isolation l Most Primary Containment Isolation Functions receive inputs l from four channels. The outputs from these channels are j arranged into two two-out-of-two logic trip systems. One j trip system initiates isolation of all inboard primary i containment isolation valves, while the other trip system '

initiates isolation of all outboard primary containment isolation valves. Each logic closes one of the.two valves i on each penetration, so that operation of either logic j isolates the penetration. The TIP ball valves isolation :

does not occur until the TIPS have been fully retracted (The i logic also sends a TIP retraction signal). !

The exception to this arrangement is the Drywell Radiation - High Function. This Function has two channels, whose outputs are arranged in two one-out-of-one logic trip i systems. Each trip system isolates one valve per associated '

penetration, similar to the two-out-of-two logic described ,

above. :

r Primary Containment Isolation Drywell Pressure - High and !

Reactor Vessel Water Level - Low, Level 3 Functions isolate ;

the Group 2, 6, 7, 10, and 12 valves. Reactor Building and i Refueling Floor Exhaust Radiation - High Functions isolate j the Group 6, 10, and 12 valves. Primary Containment :

Isolation Drywell Radiation - High Function isolates the 18 >

inch containment purge and vent valves.

t

3. 4. Hiah Pressure Coolant Iniection System Isolation and Reactor Core Isolation Coolina System Isolation Most Functions that isolate HPCI and RCIC receive input from !

two channels, with each channel in one trio system using a !

one-out-of-one logic. Each of the two trip systems in each j isolation group is connected to one of the two valves on ;

each associated penetration.

t I

(continued)

HATCH UNIT 1 B 3.3-147 REVISION A i

Primary Containment Isolation Instrumentction B 3.3.6.1 BASES BACKGROUND 3. 4. Hiah pressure Coolant In.iection System Isolation and Reactor Core Isolation Coolina System Isolation (continued)

The exceptions are the HPCI and RCIC Turbine Exhaust Diaphragm Pressure - High and Steam Supply Line Pressure - Low Functions. These Functions receive inputs from four turbine exhaust diaphragm pressure and four steam supply pressure channels for each system. The outputs from the turbine exhaust diaphragm pressure and steam supply pressure channels are each connected to two two-out-of-two trip systems. Additionally, each trip system of the Steam _

Line Flow - High Functions receives input from a low differential pressure channel. The low differential pressure channels are not required for OPERABILITY. Each trip system isolates one valve per associated penetration.

HPCI and RCIC Functions isolate the Group 3, 4, 8, and 9 valves.

5. Reactor Water Cleanuo System Isolation The Reactor Vessel Water Level - Low Low, level 2 Isolation Function receives input from four reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected into two two-out-of-two trip systems.

The Area Temperature - High Function receives input from six temperature monitors, three to each trip system. The Area Ventilation Differential Temperature - High function receives input from six differential temperature monitors, three in each trip system. These are configured so that any one input will trip the associated trip system. Each of the two trip sy<.tems is connected to one of the two valves on the RWCU penetration. However, the SLC System Initiation Function only provides an input to one trip system, thus closes only one valve.

RWCU Functions isolate the Group 5 valves.

6. RHR Shutdown Coolina System Isolation The Reactor Vessel Water Level - Low, Level 3 Function receives input from four reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected to two two-out-of-two trip systems.

(continued)

B 3.3-148 HATCH UNIT 1 REVISIONf{

l l

Primary Containment Isolation Instrumentation B 3.3.6.1 b

V BASES APPLICABLE 2.d. 2.e. Reactor Buildino and Refuelino Floor Exhaust SAFETY ANALYSES, Radiation - Hiah LCO, and APPLICABILITY High secondary containment exhaust radiation is an (continued) indication of possible gross failure of the fuel cladding.

The release may have originated from the primary containment due to a break in the RCPB. When Exhaust Radiation - High is detected, valves whose penetrations communicate with the primary containment atmosphere are isolated to limit the release of fission products.

The Exhaust Radiation - High signals are initiated from radiation detectors that are located near the ventilation exhaust ductwork coming from the reactor building and the refueling floor zones, respectively. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Four channels of Reactor Building Exhaust - High Function and four channels of Refueling Floor Exhaust - High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are chosen to ensure radioactive releases do not exceed offsite dose limits.

These Functions isolate the Group 6, 10, and 12 valves.

Hiah Pressure Coolant In.iection and Reactor Core Isolation Coolino Systems Isole'i_on 3.a. 4.a. HPCI and RCIC Steam Line Flow - Hiah -

Steam Line Flow - High Functions are provided to detect a break of the RCIC or HPCI steam lines and initiate closure of the steam line isolation valves of the appropriate system. If the steam is allowed to continue flowing out of the break, the reactor will depressurize and the core can uncover. Therefore, the isolations are initiated on high flow to prevent or minimize core damage. The isolation action, along with the scram function of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Specific credit for these Functions is not assumed in any FSAR accident analyses since the bounding analysis is performed for large breaks such as (continued)

HATCH UNIT 1 B 3.3-157 REVISION A

Primary Containment Isolation Instrumentation I B 3.3.6.1 BASES h

APPLICABLE 3.a. 4.a. HPCI and RCIC Steam Line Flow - Hiah SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding.

The HPCI and RCIC Steam Line Flow - High signals are initiated from transmitters (two for HPCI and two for RCIC) that are connected to the system steam lines. Two channels of both HPCI and RCIC Steam Line Flow - High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation functicn.

1 The Allowable Values are chosen to be low enough to ensure I that the trip occurs to prevent fuel damage and maintains I the MSLB event as the bounding event. The Allowable Values correspond to s 215 inches water column for HPCI and s 190 inches water column for RCIC, which are the parameters monitored on control room instruments.

These Functions isolate the Group 3 and 4 valves, as appropriate.

3.b. 4.b. HPCI and RCIC Steam Supply line Pressure - Low Low MSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue operation of the associated system's turbine. These isolations are for equipment protection and are not assumed in any transient or accident analysis in the FSAR. However, -

they also provide a diverse signal to indicate a possible system break. These instruments are included in Technical Specifications (TS) because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations. Therefore, they meet Criterion 4 of the NRC Policy Statement (Ref. 6).

The HPCI and RCIC Steam Supply Line Pressure - Low signals are initiated from transmitters (four for HPCI and four for RCIC) that are connected to the system steam line. Four channels of both HPCI and RCIC Steam Supply Line Pressure - Low Functions are available and are required to (continued)

HATCH UNIT I B 3.3-158 REVISIONA[

/

LOP Instrumentation B 3.3.8.1 IO B 3.3 INSTRUMENTATION LJ B 3.3.8.1 Loss of Power (LOP) Instrumentation BASES BACKGROUND Successful operation of the required safety functions of the Emergency Core Cooling Systems (ECCS) is dependent upon the availability of adequate power sources for energizing the various components such as pump motors, motor operated valves, and the associated control components. The LOP instrumentation monitors the 4.16 kV emergency buses.

Offsite power is the preferred source of power for the 4.16 kV emergency buses. If the monitors determine that insufficient power is available, the buses are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources.

Each 4.16 kV emergency bus has its own independent LOP instrumentation and associated trip logic. The voltage for each bus is monitored at two levels: 4.16 kV Emergency Bus Undervoltage Loss of Voltage and Degraded Voltage, however, only the Loss of Voltage Function is part of this LCO. The Loss of Voltage Function causes various bus transfers and

/, _ disconnects and is monitored by two undervoltage relays for (3/ each emergency bus, whose outputs are arranged in a two-out-of-two logic configuration for all affected components except the DGs. The DG start logic configuration is one-out-of-two (Ref. 1). The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a LOP trip signal to the trip logic. j l

Each 4.16 kV emergency bus has its own independent LOP alarm I instrumentation to provide an anticipatory alarm and the l initiation of corrective measures to restore emergency bus j voltages. The alarms are set higher than the LOP relays. <

The alarm setpoints are approximately midway between the i calculated minimum expected voltage and the calculated ,

minimum required voltage, based on the maximum expected I operating; i.e., non-LOCA, load conditions. The alarm setpoints signify that adequate voltage is available for normal operations. The LOP anticipatory alarms provide a j total time deity of 60 seconds to reduce the possibility of j nuisance alarms., while permitting prompt detection of !

potential low voltage conditions.

I

(continued)

HATCH UNIT 1 B 3.3-201 REVISION C

LOP Instrumentation B 3.3.8.1 BASES h

BACKGROUND Each 4.16 kV emergency buc has a dedicated low voltage (continued) annunciator fed by two relays and their associated time delays. The logic for the annunciation function is arranged in a two-out-of-two configuration. l APPLICABLE The LOP instrumentation is required for Engineered Safety SAFETY ANALYSES, Features to function in any accident with a loss of offsite LCO, and power. The required channels of LOP instrumentation ensure APPLICABILITY that the ECCS and other assumed systems powered from the DGs, provide plant protection in the event of any of the Reference 2, 3, and 4 analyzed accidents in which a loss of offsite power is assumed. The initiation of the DGs on loss of offsite power, and subsequent initiation of the ECCS, ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

O 4

(continued)

HATCH UNIT 1 B 3.3-201A REVISION h

j LOP Instrumentation I B 3.3.8.1 !

BASES I

APPLICABLE The Bus Undervoltage Allowable Values are low enough to SAFETY ANALYSES, prevent inadvertent power supply transfer, but high enough LCO, and to ensure that power is available to the required equipment.

APPLICABILITY The Time Delay Allowable Values are long enough to provide (continued) time for the offsite power supply to recover to normal voltages, but short enough to ensure that power is available r to the required equipment. ,

Two channels of 4.16 kV Emergency Bus Undervoltage (Loss of i Voltage) Function per associated emergency bus are only ;

required to be OPERABLE when the associated DG is required :

to be OPERABLE to ensure that no single instrument failure i can 3reclude the DG function. (Two channels input to each !'

of tie three DGs.) Refer to LC0 3.8.1, "AC Sources - Operating," and 3.8.2, "AC Sources - Shutdown," .

for Applicability Bases for the DGs. . !

2. 4.16 kV Emeroency Bus Undervoltaae (Deoraded Voltaae)

A reduced voltage condition on a 4.16 kV emergency bus indicates that, while offsite power may not be completely :

lost to the respective emergency bus, available power may be i O insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS Function.

Therefore, power supply to the but is transferred from i

i i

offsite power to onsite DG power when the voltage on the bus '

drops below the Degraded Voltage Function Allowable Values (degraded voltage with a time delay). This ensures that adequate power will be available to the required equipment.

The Bus Undervoltage Allowable Values are low enough to i prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the large I ECCS motors. The Time Delay Allowable Values are long enough for the offsite power supply to usually recover. '

This minimizes the potential that short duration disturbances will adversely impact the availability of the i offsite power supply. Manual actions are credited in the i range of 78.8 to 92% of 4.16 kV to restore bus voltages or to initiate a plant shutdown. The range specified for manual actions indicates that sufficient power is available to the large ECCS motors; however, sufficient voltsge for j equipment at lower voltages required for LOCA conditions may ;

not be available. '

(continued)

HATCH UNIT 1 B 3.3-203 REVISION C

LOP Instrumentation B 3.3.8.1 BASES APPLICABLE Two channels of 4.16 kV Emergency Bus Undervoltage (Degraded SAFETY ANALYSES, Voltage) Function per associated bus are only required to be LCO, and OPERABLE when the associated DG is required to be OPERABLE APPLICABILITY to ensure that no single instrument failure can preclude (continued) the DG function. (Two channels input to each of the three emergency buses and DGs.) Refer to LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs.

3. 4.16 kV Emeroency Bus Undervoltaae (Anticipatory Alarm)

A reduced voltage condition on a 4.16 kV emergency bus indicates that, while offsite power is adequate for normal operating conditions, available power may be marginal for some equipment required for LOCA conditions. Therefore, the anticipatory alarms actuate when the 4.16 kV bus voltages approach the minimum required voltage for normal; i.e., non-LOCA conditions. This ensures that manual actions will be initiated to restore the bus voltages or to initiate a plant shutdown.

Two channels of 4.16 kV Emergency Bus Undervoltage l (Anticipatory Alarm) Function per associated bus are only required to be OPERABLE when the associated DG is required to be OPERABLE. (Two channels input to each of the three emergency buses.)

ACTIONS A Note has been provided to modify the ACTIONS related to LOP instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables i expressed in the Condition, discovered to be inoperable or -

not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable LOP instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable LOP instrumentation channel.

(continued)

HATCH UNIT 1 B 3.3-203A REVISION

l i

LOP Instrumentation ,

B 3.3.8.1 I

,~ ,

Q, , BASES ACTIONS A.1 (continued)

With one or more channels of Function 1 or 2 inoperable, the function does not maintain initiation capability for the associated emergency bus. Therefore, only I hour is allowed !

to restore the inoperable channel to OPERABLE status. Ihe Required Action does not allow placing a channel in trip since this action will result in a DG initiation. .

t P

G V !

l O <<e 14 ee)

HATCH UNIT 1 B 3.3-203B REVISION l

LOP Instrumentation B 3.3.8.1 BASES ACTIONS M (continued)

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channel s.

M Each 4.16 kV bus has a dedicated annunciator fed by two relays and associated time delays in a two-out-of-two logic configuration. Both relays and their associated time delays are required to be OPERABLE. Therefore, the loss of either i required relay or time delay renders Function 3 incapable of l performing the intended function. Since the intended function is to alert personnel to a lowering voltage condition and the voltage reading is available for each bus on the control room front panels, the Required Action is verification of the voltage to be above the annunciator setpoint (nominal) hourly.

C.1 l

l If any Required Action and associated Completion Time are not met, the associated Function does not maintain I

initiation capability for the associated emergency bus.

Therefore, the associated DG(s) is declared inoperable immediately. This requires entry into applicable Conditions and Required Actions of LCO 3.8.1 and LCO 3.8.2, which

provide appropriate actions for the inoperable DG(s).

SURVEILLANCE As noted at the beginning of the SRs, the SRs for each LOP REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.8.1-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains initiation capability (for Functions 1 and 2) and annunciation capability (for Function 3). Functions 1 and 2 maintain initiation capability provided that, for 2 of the 3 emergency buses, the following can be initiated by the Function: DG start, disconnect from the offsite power (continued)

HATCH UNIT 1 B 3.3-204 REVISION [h

LOP Instrumentation B 3.3.8.1 BASES SURVEILLANCE source, DG output breaker closure, load shed, and REQUIREMENTS activation of the ECCS pump power permissive. Upon (continued) completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions ,

taken.

i SR 3.3.8.1.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation or a failure of annunciation has not occurred. A CHANNEL CHECK is defined for Function 3 to be a comparison of the annunciator status to the bus voltage and an annunciator test confirming the annunciator is capable of lighting and sounding. A CHANNEL CHECK will detect gross channel failure or an annunciator failure; thus, it is key to verifying the instrumentation centinues to operate properly between each CHANNEL CALIBRATION.

If a channel is outside the match criteria, it may be an indication that the instrument has drifted outside its

/ limit.

The frequency is based upon operating experience that demonstrates channel failure is rare. Thus, performance of the CHANNEL CHECK ensures that undetected outright channel or annunciator failure is limited to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The CHANNEL CHECK supplements less formal, but more frequent, checks of :

channels during normal operational use of the displays '

associated with channels required by the LCO.

SR 3.3.8.1.2 l A CHANNEL FUNCTIONAL TEST is performed on each required <

channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be .

consistent with the assumptions of the current plant !

specific setpoint methodology. r The Frequency of 31 days is based on operating experience with regard to channel 0PERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event. l (Continued)

HATCH UNIT 1 B 3.3-204A REVISION

RHR Shutdown Cooling System - Cold Shutdown B 3.4.8 BASES ACTIONS B.1 and 8.2 (continued)

During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem or recirculation '

pump), the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.

SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.

O :

REFERENCES 1. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

k O

HATCH UNIT 1 B 3.4-43 REVISION A

RCS P/T Limits B 3.4.9 8 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LC0 limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

This Specification contains P/T limit curves for heatup, cooldown, and inservice leakage and hydrostatic testing, and also limits the maximum rate of change of reactor coolant temperature. The criticality curve provides limits for both heatup and criticality.

Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LC0 establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure. Therefore, the LC0 limits apply mainly to the vessel.

10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials. Reference I requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section III, Appendix G (Ref. 2).

The actual shift in the RT,, of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be adjusted, (continued)

HATCH UNIT 1 B 3.4-44 REVISION [

RCS P/T Limits f B 3.4.9 BASES BACKGROUND as necessary, based on the evaluation findings and the j (continued) recommendations of Reference 5.

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most ;

restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor t vassel will dictate the most restrictive limit. Across the i span of the P/T limit curves, different locations are more -

restrictive, and, thus, the curves are composites of the most restrictive regions. ;

i The heatup curve represents a different set of restrictions i than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The i thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

?

The criticality limits include the Reference I requirement !

that they be at least 40 F above the heatup curve or the :

cooldown curve and not lower than the minimum permissible !

temperature for the inservice leakage and hydrostatic j testing. :

The consequence of violating the LC0 limits is that the RCS j has been operated under conditions that can result in l brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event !

these limits are exceeded, an evaluation must be performed :

to determine the effect on the structural integrity of the ,

RCPB components. ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating i en operating event that causes an excursion outside the ;

limits. ;

APPLICABLE The P/T limits are not derived from Design Basis Accident i SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature ,

rate of change conditions that might cause undetected flaws -

to propagate and cause nonductile failure of the RCPB, a l condition that is unanalyzed. Reference 8 approved the ;

curves and limits specified in this section. Since the ,

(Continued)

HATCH UNIT 1 B 3.4-45 REVISION,D'{

RCS P/T Limits B 3.4.9 BASES h

APPLICABLE P/T limits are not derived from any DBA, there are no SAFETY ANALYSES acceptance limits related to the P/T limits. Rather, the (continued) P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 9). l LC0 The elements of this LC0 are:

a. RCS pressure and temperature are within the limits specified in Figures 3.4.9-1 and 3.4.9-2, and heatup or cooldown rates are s 100 F during RCS heatup, cooldown, and inservice leak and hydrostatic testing; ,

b. The temperature difference between the reactor sessel bottom head coolant and the reactor pressure vessel (RPV) coolant is 5 145 F during recirculation pump startup;

c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is s 50 F during recirculation pump startup;

d. RCS pressure and temperature are within the criticality limits specified in Figure 3.4.9-3, prior l to achieving criticality; and

e. The reactor vessel flange and the head flange temperatures are 2 76 F when tensioning the reactor '

vessel head bolting studs. l These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.

The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and inservice (continued)

HATCH UNIT 1 B 3.4-46 REVISIONf{

i RCS P/T Limits !

B 3.4.9 .

BASES l

ACTIONS C.1 and C.2 (continued)

Operation outside the P/T limits in other than MODES 1, 2, i and 3 (including defueled conditions) must be corrected so l that the RCPB is returned to a condition that has been :

verified by stress analyses. The Required Action must be j initiated without delay and continued until the limits are !

restored.

Besides restoring the P/T limit parameters to within limits, ,

an evaluation is required to determine if RCS operation is !

allowed. This evaluation must verify that the RCPB j

integrity is acceptable and must be completed before approaching criticality or heating up to > 212 F. Several i methods may be used, including comparison with pre-analyzed ;

transients, new analyses, or inspection of the components. i ASME Code, Section XI, Appendix E (Ref. 6), may be used to :

support the evaluation; however, its use is restricted to !

evaluation of the beltline. ;

i Condition C is modified by a Note requiring Required Action i C.2 be completed whenever the Condition is entered. The i Note emphasizes the need to perform the evaluation of the O effects of the excursion outside the allowable limits.

Restoration alone per Required Action C.1 is insufficient t

i because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

}

SURVEILLANCE SR 3.4.9.1 {

REQUIREMENTS i Verification.that operation is within limits is required l :

every 30 minutes when RCS pressure and temperature !

conditions are undergoing planned changes. This Frequency l is considered reasonable in view of the control room ,

indication available to monitor RCS status. Also, since ,

temperature rate of change limits are specified in hourly i increments, 30 minutes permits a reasonable time for l assessment and correction of minor deviations.

Surveillance for.heatup, cooldown, or inservice leakage and ;

hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the ;

activity are satisfied. !

(

(continued) i i

HATCH UNIT i B 3.4-49

-REVISION [

RCS P/T Limits B 3.4.9 BASES h SURVEILLANCE SR 3.4.9.1 (continued)

REQUIREMENTS This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and RCS inservice leakage and hydrostatic testing.

SR 3.4.9.2 A separate limit is used when the reactor is approaching -

criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.

Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.

SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable limits l ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idit' recirculation loop (Ref. 7) are satisfied.

Performing the Surveillance within 15 minutes before -

starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.

An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop.

(continued)

HATCH UNIT 1 B 3.4-50 REVISIONdh

/

l

RCS P/T Limits l B 3.4.9 i BASES l

SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4 (continued)

REQUIREMENTS SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, !

2, 3, and 4. In MODE 5, the overall stress on limiting ;

components is lower. Therefore, AT limits are not required. !

SR 3.4.9.5. SR 3.4.9.6. and SR 3.4.9.7 l Limits on the reactor vessel flange and head flange !

temperatures are generally bounded by the other P/T limits !

during system heatup and cooldown. However, operations l approaching MODE 4 from MODE 5 and in MODE 4 with RCS l temperature less than or equal to certain specified values f require assurance that these temperatures meet the LC0 i limits.

The flange temperatures must be verified to be above the :

limits 30 minutes before and while tensioning the vessel !

head bolting studs to ensure that once the head is tensioned l the limits are satisfied. When in MODE 4 with RCS :

temperature s 86 F, 30 minute checks of the flange l l temperatures are required because of the reduced margin to i the limits. When in MODE 4 with RCS temperature s 106'F, l j monitoring of the flange temperature is required every l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to ensure the temperature is within the limits ;

specified. l l t

The 30 minute Frequency reflects the urgency of maintaining l the temperatures within limits, and also limits the time !

that the temperature limits could be exceeded. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months !

Frequency is reasonable based on the rate of temperature change possible at these temperatures.

-l i

SR 3.4.9.5 is modified by a Note that requires the !

Surveillance to be performed only when tensioning the reactor vessel head bolting studs. SR 3.4.9.6 is modified by a Note that requires the Surveillance to be initiated

30 minutes after RCS temperature s 86 F in Mode 4. l !

SR 3.4.9.7 is modified by a Note that requires the Surveillance to be initiated 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature ;

5 106 F in Mode 4. The Notes contained in these SRs are l necessary to specify when the reactor vessel flange and head flange temperatures are required to be verified to be within l the limits specified. l ],

(continued)

HATCH UNIT 1 B 3.4-51 REVISIONf(E i

I RCS P/T Limits B 3.4.9 i BASES REFERENCES 1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.

3. ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.

4. 10 CFR 50, Appendix H.

5. Regulatory Guide 1.99, Revision 2, May 1988.

6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.

7. FSAR, Section 14.3.6.2.

8. George W. Rivenbark (NRC) letter to J. T. Beckham, Jr.

(GPC), Amendment 126 to the Operating License; dated June 20, 1986.

9. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

O HATCH UNIT I B 3.4-52 REVISION k

SCIVs B 3.6.4.2 l i

BASES SURVEILLANCE SR 3 6.4.2.2 REQUIREMENTS ;

(continued) Verifying that the isolation time of each power operated and ,

each automatic Unit 1 SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures i that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. The Frequency of this SR was developed based upon engineering judgment and i the similarity to PCIVs. ;

SR 3.6.4.2.3 Verifying that each automatic Unit 1 SCIV closes on a secondary containment isolation signal is required to prevent leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCIV will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 -

overlaps this SR to provide complete testing of the safety ;

function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply O during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. ,

REFERENCES 1. FSAR, Section 14.3.3. -

2. FSAR, Section 14.3.4.

l

3. Technical Requirements Manual. ,

4. NRC No.93-102, " Final Policy Statement on Technical i Specification Improvements," July 23, 1993. ;

i F

O !

HATCH UNIT 1 B 3.6-87 REVISION A l l

SGT System B 3.6.4.3 8 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 Standby Gas Treatment (SGT) System BASES BACKGROUND The SGT System is required by 10 CFR 50, Appendix A, GDC 41,

" Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment.

The Unit I and Unit 2 SGT Systems each consists of two fully redundant subsystems, each with its own set of dampers, charcoal filter train, and controls. The Unit 1 SGT subsystems' ductwork is separate from the inlet to the filter train to the discharge of the fan. The rest of the ductwork is common. The Unit 2 SGT subsystems' ductwork is separate except for the suction from the drywell and torus, which is common (However, this suction path is not required for subsystem OPERABILITY).

Each charcoal filter train consists of (components listed in order of the direction of the air flow):

a. A demister or moisture separator;

b. An electric heater;

c. A prefilter;

d. A high efficiency particulate air (HEPA) filter; -

e. Two charcoal adsorbers for Unit I subsystems and one charcoal adsorber for Unit 2 subsystems;

f. A second HEPA filter; and

g. An axial vane fan.

The sizing of the SGT Systems equipment and components is based on the results of an infiltration analysis, as well as an exfiltration analysis of the secondary containment. The internal pressure of the SGT Systems boundary region is (continued) 1 HATCH UNIT 1 B 3.6-88 REVISION [

f i

SGT System l B 3.6.4.3 j i

BASES I

BACKGROUND maintained at a negative pressure of 0.25 inches water gauge !

(continued) when the system is in operation, which represents the !

internal pressure required to ensure zero exfiltration of i air from the building when exposed to a 10 mph wind.

l The demister is provided to remove entrained water in the '

air, while the electric heater reduces the relative humidity I of the airstream to < 70% (Refs. 2 and 3). The prefilter i removes large particulate matter, while the HEPA filter .

removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorbers remove gaseous -

elemental iodine and organic iodides, and the final HEPA ,

filter collects any carbon fines exhausted from the charcoal i adsorber.

The Unit I and Unit 2 SGT Systems automatically start and operate in response to actuation signals indicative of t conditions or an accident that could require operation of !

the system. Following initiation, all required charcoal !

filter train fans start. Upon verification that the i required subsystems are operating, the redundant required i subsystem is normally shut down.

l O

APPLICABLE The design basis for the Unit I and Unit 2 SGT Systems is to l SAFETY ANALYSES mitigate the consequences of a loss of coolant accident and i fuel handling accidents (Refs. 2 and 3). For all events i analyzed, the SGT Systems are shown to be automatically i initiated to reduce, via filtration and adsorption, the !

radioactive material released to the environment.

The SGT System satisfies Criterion 3 of the NRC Policy ,

Statement (Ref. 4). :

LCO Following a DBA, a minimum of two SGT subsystems are required to maintain the Unit I secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LC0 requirements for three OPERABLE subsystems (two Unit 1 SGT subsystems and one Unit 2 subsystem) ensures operation of at least two SGT t subsystems in the event of a single active failure.

(continued) 4 HATCH UNIT 1 B 3.6-89 REVISION g

SGT System B 3.6.4.3 BASES LCO In addition, with Unit I secondary containment in the (continued) modified configuration, the Unit 1 SGT System valves to the Unit I reactor building zone are not included as part of Unit 1 SGT System OPERABILITY (i.e., the valves may be secured closed and are not required to open on an actuation signal).

APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, Unit I and Unit 2 SGT Systems OPERABILITY are required during these MODES.

In MODES 4 and 5, the probability and consequences of a LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT Systems in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment.

ACTIONS A.1 With one required Unit 1 or Unit 2 SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this condition, the remaining required OPERABLE SGT subsystems are adequate to perform the required radioactivity release control function. However, the -

overall system reliability is reduced because a single failure in one of the remaining required OPERABLE subsystems could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time

! is based on consideration of such factors as the l availability of the OPERABLE redundant SGT subsystems and the low probability of a DBA occurring during this period.

! B.1 and B.2 l

If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the (continued)

HATCH UNIT 1 B 3.6-90 REVISION A l

1

SGT System ,

B 3.6.4.3 I

BASES ACTIONS B.1 and B.2 (continued) plant must be brought to a MODE in which the LCO does not ,

apply. To achieve this status, the plant must be brought to :

at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within '

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant !

conditions from full power conditions in an orderly manner and without challenging plant systems.

C.I. C.2.1. C.2.2. and C.2.3 During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, when Required Action A.1 cannot be completed within the required Completion Time, two remaining required .

OPERABLE SGT subsystems should immediately be placed in ;

operation. This action ensures that the remaining subsystems are OPERABLE, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected. ;

An alternative to Required Action C.1 is to immediately '

suspend activities that represent a potential. for releasing radioactive material to the secondary containment, thus '

placing the plant in a condition that minimizes risk. If :

applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies must immediately be suspended. Suspension of ;

these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in .

order to minimize the probability of a vessel draindown and .!

subsequent potential for fission product release. Actions !

must continue until OPDRVs are suspended.

The Required Actions of Condition C have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving ,

irradiated fuel assemblies while in H0DE 4 or 5, LCO 3.0.3 !'

would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is ,

independent of reactor operations. Therefore, in either i case, inability to suspend movement of irradiated fuel ,

assemblies would not be a sufficient reason to require a !

reactor shutdown. l .

(continued)

HATCH UNIT 1 B 3.6-91 REVISION A l l

i

SGT System B 3.6.4.3 BASES h

ACTIONS RJ (continued)

If two or three required SGT subsystems are inoperable in MODE 1, 2 or 3, the Unit I and Unit 2 SGT Systems may not be capable of supporting the required radioactivity release control function. Therefore, LC0 3.0.3 must be entered immediately.

E.1. E.2. and E.3 When two or three required SGT subsystems are inoperable, if applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPORVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action E.1 has been modified by a Note stating that &

LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LC0 3.0.3 would not specify W

any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each required Unit I and Unit 2 SGT subsystem for 2 10 continuous hours ensures that they are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action.

Operation with the heaters on for 2 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and-controls and the redundancy available in the system.

(continued)

HATCH UNIT 1 , B 3.6-92 REVISION (

SGT System B 3.6.4.3 BASES i

SURVEILLANCE SR 3.6.4.3.2 ;

REQUIREMENTS i (continued) This SR verifies that the required Unit I and Unit 2 SGT '

filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP t includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and i following specific operations). Specific test frequencies !

and additional information are discussed in detail in the l VFTP.

f I

SR 3.6./.3.3 !

This SR verifies that each required Unit I and Unit 2 SGT subsystem starts on receipt of an actual or simulated ;

initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in i SR 3.3.6.2.5 overlaps this SR to provide complete testing of :

the safety function. While this Surveillance can be !

performed with the reactor at power, operating experience j has shown that these components usually pass the !

s Surveillance when performed at the 18 month Frequency. ;

, Therefore, the Frequency was found to be acceptable from a ;

reliability standpoint. j REFERENCES 1. 10 CFR 50, Appendix A, GDC 41. j

2. FSAR, Section 5.3. }

3. Unit 2 FSAR, Section 6.2.3.

4. NRC No.93-102, " Final Policy Statement on Technical :

Specification Improvements," July 23, 1993.

l 4

4 Q (continued) f j HATCH UNIT 1 B 3.6-93 REVISION i

i

o UNIT I MARKUP OF CURRENT TECIINICAL k SPECIFICATIONS AND DISCUSSION OF CHANGES i

I l

l O l l

l i

O

ed i. c S. DoerTt%c cycle - An nneratin _ evele is the intervet~ tie ~tIIeen ,

, (, the end of -hedtri re ue ., antage and the end of the

_new u sequent scheduled refueling outage for the same W t.

' Primary containment integrity - Primary containment integrity T

' ns that the drywell and suppression chamber are intact )

Y all . following conditions are satisfied: )

g\

\-

f All non- tomatic containment isolation valv on lines 1.

connected t he reactor coolant system o ontainment which are not r gired to be open du g accident conditions are closed. These va ves may be ened to perform '

operational activities. y

2. At least one door in e personne lock is closed and sealed.

NO 3. All aut ic containment isolation valves are able or ctivated in the isolated position. ,

L .

All blind flanges and manways are closed u

.~ ective Action - A protective action is an ac+4nn initieted by the pro 6cutiv: :"stam whefmer-TTiiift Ts reached. A protective actio c annel or system levei and is :::ential to n . e accomplishment of a safety action.

}

(]

N .ctiva lunction - A protective functi monitoring ~7 of one or more plant va.' hl : e tt tions and the associated Q

\, em actions which eventually resuit in _.)

R.TP) gTP_ An be e4Af rwArcore M%4Q

^-

g Rated Thermal Power - Ratsd 6nem.:::

-- r ~~ ni me r eac 6ar is T S. M. . 2..t i... M.. 9 2..ti.F. ,.R N. ,,.of 243 To. . . . . .

W . M.L....

E.,t " -

%roofwf

@ FTe'ittwJode - The reactor switch p mode is established by~tW f'tTEC SHUTDOWN, Switch pos .

, unsa ...e (nur possible reactor g START & HOT STANDBY modes e Mode, Shutdown Mode, Start & roi. Ste-

,p e and Run Mode. - }

^

k y Y'. Reactw hwer Operation - Reactor power o a+4ca 1: r operation with' " E' t L i e TART & HOT STANDBY or n the reactor h q % of rated h7 RUN ermal power. _

J I

HATCH - UNIT 1 1.0-4 7,f 2 2.

Insert i m PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Section 13.6, Startup and Power Test Program, of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

Insert 1.1F (MOT tASED)

A

'O Insert 1.1G MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

i I.0 Hatch Unit 1 Insert i e ,f n_

.SpeaLM 3 A 9 __

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

'v/ PRIMARY SYSTEM BOUNDARY 6 PRIMARY SYSTEM BOUE ARY elicability licability The iting Conditions for The urveillance Requirements apply Operat apply to the oper- to t riodic examination and stir.g sta s of the reactor testing irements for the coolant sys . reactor c lant system.

Obiective Objective The objective of the initing The objective of Surveillance Conditions for Operat n is to Requirements is to e mine the assure the integrity an safe condition of the resc r coolant operation of the reactor olant system and the operati f the system. safety devices related to t.

}eecifications Soecifications A. Reactor Coolant Heat-Un ag[ A. Reactor Coolant Heat-Un and Cooldown gutleggt The average rate of reactor The reactor coolant system coolant temperature change temperature and pressure SR).Q. 9,l bD4 9 during normal heatup or cool- shall be detemined to be within the limits of down shall not exceed 100*F/hr when averaged over a one-hour Specifications 3.6.A. and 3.6.B.

545.494.b period. at least once every 30 minutes during reactor coolant \ q g heatup and cooldown. )g g O B. Reacter Vessel Temperature and B. Egaetor Vessel Tennerature P.nd Pressure Pressure l 1. The reactor vessel shell temper- Reactor vessel metal toeperature ko 3,q. y atures during inservice hydro- at the outside surface of the gg3, q* 9 *!

static or leak testing shall be bottom head in the vicinity of at or above the temperatures the control rod drive housing shown on the curve of Figure 3.6-1, and reactor vessel shell adjacent M3A 9 I A to shell flange shall be re-corded at least every o \

(minutes during in-servit.e gv d hydrostatic nr leak testina The vessel pressure is 2 312 4

p1 4 <P_sig . -

i l

t fJ HATCH - tlNIT 1 3.6-1 Amendment No. H , 69, 126 Ch

4$(s.koa) 3.Y 9 .

l l

i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.6.B. Reactor Vessel Temperature and 4.6.B. Reactor Vessel Temperature and j Pressure (Continued) Pressure (Continued)

2. During heatup by non-nuclear Test specimens represe ing the 3 means, cooldown following nuclear actor vessel, base wel and weld LlolN'i shutdown or low level physics tests, the reactor vessel shell and fluid he affected zone metal inst led in the reactor ve el e

temperatures of Soecificatic.n 4.6.A. adjac to the vessel wall a I shall be at or above the tesperatures the co midplane level before M 5 4 i.I .4 shown on the curve of Figure 3.5-2. the start f operation. The number and of specimens are in accortlanc with GE report NEDD-10115. specimens meet the intent of E185-70.

3. During all operation with a critical core, other than for low level physics e next surveillan capsule

@ ~3.4 0 tests, the reactor vessel shell and sh I be removed f the ves- 4 , 2, fluid temperatures of Specification sel kpproximately 1 EFPY i 4.6.A. shall be at or above the temper- of ope tion, as reconne ed in atures snown on the curves of ASTM El -82, but not to coed N b Figure 3.6-3. 6 EFPY.

Profbse d s R f .4 9.L 3.6.C. Reactor vessel Head Stud Tensionino The reactor vessel head bolting

[ b 3,q.1 studs shall not be under tension V unless the temperature of the

< vessel head flance and the head St.~5.4 9.5,4f7 is greater than 76'F. ,

C. Reactor Vessel Head Stud ,

Tensionina l D. Idle Recirculation tooo Startuo When the reactor vessel head g g 3 4,3. c;-

studs are under tension and the The pump in an idle recirculation reactor is in the Cold Shutdown MM 34 loop shall not be started unless Condition, the reactor vessel g,( ) . 4 3 ,-}

d g,g the temperatures of the coolant shell temperature tunediately within the idle and operatino re- below the head flange shall be A *l 54349.i circulation loops are within 50*F of each other.

permanently reco % r e ,,3 s & , ,,m,;

D. Idle Recirculation Loon Startuo .po,. 5 l If.T.5~

SO9.9.6,a D -

or 1.0 and duri,un tiarfD of an 8,/t% 93 l

b idle recirculation loop, the tem-perature of the reactor coolant in the operating and idle loops shall be compared and permanently W 4 1,q recorded. '

I i

p HATCH - UNIT 1 3.5-2 Amendment No. 69, 126

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1

Si br abda 5. 4. T tlMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREENTS

( 3.6.E. Recirculation Puno Start 4.6.E. Racirtulation Pee Start The reactor recirculation pumps (14TFTiirsraHing ia a recirculation g 'N'g shall not be started unless the pop, the reactor Coolant temper-coolant temperatures between the atures in the done and in the 3

done and the bcttom head drain bottom head drain shall be com *g 3.q,$.3 sk5.4.9.3 are within 145'F. pared and permanently recorded.

i (OfoX b Y%

A ,6,~a e O

( HATCH - UNIT 3.6-3 344

. ${ET FAG ( *34_{

O EPPY 410121416 i

1200

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ADJUSTED CORE DELTLiset, f 1M T PLAW 1m i .

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[ h VERTICAL LIRAIT LINE POR PRESOURE ADOVE SWE MVDROTEST G12 pelgl, 400 #

mASED oN 1oCPRoo AprEseDix o REQUIREMENT OF (RTggyy + Sp*F),

PLANGE RES40N RTNOT*18'E DOLT PRELOAD TEteERATURE OF M*F DASED ON RECObe8 ENDED 200 , (RTegyy + 00*F) POft 8.364N. PLAW ON CLOSURE PLAf8GE REGION, RTggyy = 198F 0

0 100 200 300 RPV METAL TEMPERATURE (#F)

Figure .34- - *=cea= --~ "= Mini- = T--;:x t;n Or a-e"= Tr.ts ,- ,

B==^d aa (" - " ? = = Teat. M ;1t:

Te e pe r,+see - Pr sss ure Lui4: fe Jnser vice Hy dra+ k. l/ p,. (>

HATCH - UNIT 1 " d # ""' "* h ' 7" Y3 Amendment No. 59. 225. 137 l l

~. _ ._. _

frecJ hcqG 3 4.4 M M7 O

~1600 i

VALID TO 16 EFFECTIVE FULL FOWER YEARS OF OPERATION 1400 l l

- 1200

.t.

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W ADJUSTED CORE BELTLINE 1/4 T FLAW, RTNDT

YF E BARADIATION SHIFT = 123*F l

W s00 / " '

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E 400 l FEEDWATER NOZZLE TEMPERATURE I LIMT FOR 1/4 T' FLAW (SWR 4 l RESULTS ADJUSTED TO 40*F RT NOTI 200

/

/ g MINIMUM OPERATING TEMPERATURE OF 7E*F BASED ON RECOMMENDED IRTNOT + SD*F) FOR O.24-IN. FLAN IN CLOSURE FLANGE REGION, RTNOT*18'F 0 f 1 0 100 200 300 400 500 800 MINIMUM VESSEL METAL TEMPERATURE ('F) 4,4 Figure . . Dmen~ ;;r:;; Mir -- T 7er: tem 1:7 ;;;n_p;gc33 7 e

M?!?UT/CCli Sad tr':.' P; ;- 7l,;iu 7,;is

.' {.9-1, " Teep ree+vre - Pre ss e e Lim i+s -Te v Kon uve loae H re+sp, r

Lo w f6 we r PhyssS Tesis cad CoolLown NllM'*y g*

HATCH - UNIT 1 4 shW/esn Amendments No. 75,57,126

. . . _ . . _ . . - . . . ._ . . _ . . _ _ . . _ . . _ . . _ . ~ . . _ ._ _ _ _ _ _

1 - fTET fy I P843hC46 3.4 9 -

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l 1000 VAUD TO 16 EFFECTIVE FULL POWER YEARS OF OPERATION ;

t 1400 !

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1 I E ADJUSTED CORE BELTLINE, I W 1M T FLAW, RTl e0T " 18*F-g ' 4RRADIATION SMOFT = 135'F f

, B =

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a ,

000

a W

E l 400 t

! i I

FEEDWATER NOZZLE TEtrERATURE I LIMIT FOR 1M T FLAW 19WIUS I

' l RESULTS ADJUSTED TO 40*F RT NOTI I 200 '

I i MINIMUM OPERATING TEMPERATURE !

LM8tf OF 79*F FROM 10CFR00 APPE00 DIX 0 i REQUIREMENT THAT (Taggg = RTaggy

WI, {

FLAleOE R7 !

o - 0007*18'F- -

0 100 200 300 400 500 soo l

~

MINIMUM VESSEL METAL TEMPERATURE (#F) l 7 H R~5 *Te mper. Lie- Preuvre L;nsh for c,;4te.IHy l l Figure #.' . Pr::r"~ varent "'-i r T1;^ ret;re fer on Critical !

0;;retir Othe- t%= tr ";nr Phy;i;;; T::'a (!;.;lJee i

^ f . = T r . .. i . W L.7 ! ^"."!O b... L Ei l hdb i HATCH - UNIT 1 Amenenent No. 75,59,126 !

i

. - . _ - ,~ -. .. ._ _ _ - _ _ _ .

[

)

i g DISCUSSION OF CHANGES

( ITS: SECTION 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS ,

ADMINISTRATIVE A.1 Reformatting and renumbering requirements are in accordance with the BWR ,

Standard Technical Specifications, NUREG 1433. As a result, the Technical

~

Specifications should be readily readable, and therefore, understandable by plant operators as well as other users. During this reformatting and i renumbering process, no technical changes are to the Technical i Specifications have been made unless there are identified and justified.

In the specific case of the Primary System Boundary Section the new '

section number is 3.4, which has been titled " Reactor Coolant System (RCS)."

A.2 These surveillances are a duplication of the regulations found in 10 CFR !

50 Appendix H. These regulations require licensee compliance and can not ,

be revised by the licensee. Therefore, these details of the regulations ,

within the Technical Specifications are repetitious. Furthermore, ;

approved exemptions to the regulations, and exceptions presented within ,

the regulations themselves, are also details which are adequately presented without repeating the details within the Technical Specifications. Therefore, retaining the requirement to meet the requirements of 10 CFR 50 Appendix H, as modified by approved exemptions, and eliminating the Technical Specification details that are also found in Appendix H, is considered a presentation preference which is administrative.

A.3 For clarity, the terms " prior to and during startup" and " prior to" have been replaced with "15 minutes." This Frequency is effectively the same '

since the proposed Surveillance now must be performed no more than 15 minutes prior to startup of the idle recirculation loop. This is essentially equivalent to the current requirements.

A.4 Title changes to the P/T curves have been made for consistency with the :

ITS SRs.

TECHNICAL CHANGE - MORE RESTRICTIVE M.1 A Surveillance Requirement has been added. SR 3.4.9.2 ensures the RCS l pressure and temperature are within the criticality limits once within 15 minutes prior to control rod withdrawal for the purpose of achieving ;

criticality. This is an additional restriction on plant operation. !

M.2 Three Surveillance Requirement frequences have been added. SR 3.4.9.5 ensures the vessel head is not tensioned at too low a temperature every 30 <

minutes. SRs 3.4.9.6 and 3.4.9.7 ensure the vessel and head flange i temperatures do not decrease below the minimum allowed temperature every 30 minutes or every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , depending upon the RCS temperature. These ;

are additional restrictions on plant operation since the current '

requirements have no times specified.

HATCH UNIT 1 1 REVISION

i DISCUSSION OF CHANGES . l ITS: SECTION 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS j i

M.3 ACTIONS have been added (proposed ACTIONS A, B, and C) to provide l direction when the LC0 is not met. Currently, no ACTIONS are provided. !

Proposed ACTIONS are consistent with the BWR Standard Technical Specifications, NUREG 1433, and are additional restrictions on plant operation. !

i I

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I O

f 1

F I

l l

I O !

HATCH UNIT 1 1k REVISION hl i

i

DISCUSSION OF CHANGES p)s

( ITS: SECTION 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS j TECHNICAL CHANGE - LESS RESTRICTIVE I

i

" Specific" L.1 The Frequency of this Surveillance has been changed from 15 minutes to 30 minutes. In addition, the Surveillance must be performed at all pressures, not just at 2: 312 psig. The metal temperature is not expected to change rapidly due to its large mass. Thus a 30 minute Frequency is adequate. In addition, this new Frequency is consistent with current Unit 2 requirements as well as the BWR Standard Technical Specifications, NUREG ;

1433.

i b

d

[

O HATCH UNIT I 2 REVISION /([{~ 3

Spe;&% 34,q 3 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS

( 3.7.B.2. Perfomance Reauirements 2. Filter Testina

i. The results of the in-place DOP as fa. The tests and analysis shall be halogenated hydrocarbon tests at performed at least once per design flows on HEPA filters and operating cycle, not to exceed 18 b.I charcoal absorber banks shall show months, or after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of A./

995 00P removal and 995 halogenate<l system operation, or following Aov'I hydrocarbon removal when tested in painting, fire or chemical release MoutJ lo accordance with ANSI N510-1975. in any ventilation zone Spegg'4*d 4 '" Odd ,

communicating with the system.

). The results of laboratory carbon fI/

f.5 ~7 sample analysis shall show 90% of b. DOP testing shall be performed radioactive methyl iodine removal after each complete or partial i when tested in accordance with replacement of the HEPA filter bank RDT-M16-IT (80*C, 95% R.H.). or after any structural maintenance

c. Fsns'shall be shown to operate within +105 -0% design flow when c. Hal enated hydrocarten testing tested in accordance with ANSI shal be perfomed after each (N510-1975. complete or partial replacement of the charcoal absorber bank or after

\ anystructuralmaintenanceonthej

( system housing.

d. Each circuit shall be operated with the heaters Q at least 10 nour g SR3.4.4.'5.1 O

O --1 2.1-Ila A - - t o. II.

us

L g DISCUSSION OF CHANGES ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM ADMINISTRATIVE A.1 The technical content of this requirement is being moved to Section 5 of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specifications, NUREG 1433. Any technical changes to this requirement are addressed in the Discussion of Changes associated with proposed Specification 5.5.7. A surveillance requirement (proposed SR 3.6.4.3.2) is added to clarify that the tests of the Ventilation Filter Testing Program must also be completed and passed for determining OPERABILITY of the SGT System. Since this is a presentation preference -

that maintains current requirements, this change is considered administrative. ,

A.2 The description of the signal used to automatically initiate the SGT System " actual or simulated initiation signal" has been added for clarity.

This is consistent with the BWR Standard Technical Specifications, NUREG 1433, and no change is intended.

A.3 This Surveillance has been deleted since there is no bypass valve in the system. The system has internal orifices for filter cooling.

A.4 A new ACTION is proposed (ACTION D) which directs entry into LC0 3.0.3 if Q two or more required standby gas treatment subsystems are inoperable in V MODES 1, 2, or 3. This avoids confusion as to the proper action if in MODES 1, 2, or 3 and simultaneously handling irradiated fuel or conducting operations with a potential for draining the vessel. Since this proposed ACTION effectively results in the same action as the current specification, this change is considered administrative.

TECHNICAL CHANGE - MORE RESTRICTIVE M.1 An Applicability has been added. The SGT System is now required to be OPERABLE during operations with a potential for draining the reactor vessel to provide mitigation if an inadvertent vessel draindown event occurs. Appropriate Required Actions have also been added (Required Actions C.2.3 and E.3. In addition, Required Actions have been added (proposed Required Actions C.2.2 and E.2) to suspend CORE ALTERATIONS, consistent with the Applicability of Secondary Containment (and SGT System). These are additional restrictions on plant operation.

M.2 An additional shutdown action has been added (Required Action B.1) to not only be in Cold Shutdown (MODE 4) within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , but to also be in Hot Shutdown (MODE 3) within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This is an additional restriction on plant operation.

O HATCH UNIT 1 1 REVISION

DISCUSSION OF CHANGES

[J L

T ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM TECHNICAL CHANGE - MORE RESTRICTIVE (continued)

M.3 The time to suspend fuel handling has been changed from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to immediately. This is an additional restriction on plant operation.

M.4 This allowance has been deleted since it is not needed in the proposed Specifications. The new Specifications will allow Unit I reactor operations for 7 days with both Unit 2 SGT subsystems inoperable (see comment L.2). Thus, an additional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , as provided by this allowance, is not needed. The deletion of this allowance is more restrictive on plant operation.

M.5 SR 3.6.4.3.1 requires the SGT System to be run 10 continuous hours each 31 days, while the CTS state a total of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . This is an additional restriction on plant operations.

TECHNICAL CHANGE - LESS RESTRICTIVE

" Specific" m

L.1 The proposed change will delete the requirement to test the other SGT (d' subsystems when one subsystem is inoperable. The requirement for demonstrating operability of the redundant subsystems was originally prescribed because there was a lack of plant operating history and a lack of sufficient equipment failure data. Since that time, plant operating experience has demonstrated that testing of the redundant subsystems when one subsystem is inoperable is not necessary to provide adequate assurance of system operability.

This change will allow credit to be taken for normal periodic Surveillances as a demonstration of operability and availability of the remaining components. The periodic frequencies specified to demonstrate operability of the remaining components have been shown to be adequate to ensure equipment operability. As stated in NRC Generic Letter 87-09, "It is overly conservative to assume that systems or components are inoperable l when a surveillance requirement has not been performed. The opposite is in fact the case; the vast majority of surveillances demonstrate the -

systems or components in fact are operable." Therefore, reliance on the '

specified surveillance intervals does not result in a reduced level of confidence concerning the equipment availability. Also, the original General Electric Standard Technical Specifications (STS), NUREG 123, and, more specifically, all the Technical Specifications approved for recently licensed BWR's accept the philosophy of system operability based on satisfactory performance of monthly, quarterly, refueling interval, post i maintenance or other specified performance tests without requiring 1

- additional testing when another system is inoperable (except for diesel I generator testing, which is not being changed). j l

HATCH UNIT 1 2 REVISIONfh

1 l

1

)

DISCUSSION OF CHANGES O ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM i

TECHNICAL CHANGE - LESS RESTRICTIVE l (continued)

L.2 An alternative is proposed to suspending operations if a SGT subsystem cannot be returned to OPERABLE status within seven days, and movement of irradiated fuel assemblies, CORE ALTERATIONS, or operations with the !

potential for draining the reactor vessel are being conducted.

~

The alternative is to initiate two OPERABLE subsystems of SGT and continue to conduct the operations. Since two subsystems are sufficient for any accident, the risk of failure of the subsystems to perform their intended function is significantly reduced if they are running. ,

L.3 The proposed ACTIONS will allow the one required Unit 2 SGT subsystem to be inoperable (thus both Unit 2 SGT subsystems are inoperable) for up to l 7 days without requiring a Unit I shutdown or suspension of operations. l' This is consistent with the 7 days allowed for an inoperable Unit 1 SGT subsystem. With two OPERABLE subsystems (two Unit 1, or one Unit I and one Unit 2), the safety analysis assumptions are met, provided no single -

active failure occurs. Thus, since 7 days has been found to be acceptable for one of the two cases (one Unit 1 SGT subsystem inoperable), it is considered justifiable for the other case (one required Unit 2 SGT .

subsystem inoperable).

f

%./

HATCH UNIT 1 3 REVISION [h

[gehs'cde'on I.I,7 tlMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS

8. Standbv Gas Treatment System B. Standby Gas Treatment System ,_

1. Operability Reovirements 1. Surveillance When System Operable 1.a. A minimum of three (2 of 2 in Unit l 1 and 1 of 2 in Unit 2) of the four At least once per operating cycle, not independent standby gas treatment f*g*7 to exceed 18 months, the following system trains shall be operable at conditions shall be demonstrated: -

all times when Unit 1 secondary containment integrity is required, a. Pressure drop across the combined y,g,7,J HEPA filters and charcoal absorber With one of the Unit 1 standby gas bank is less than 6 inches of treatment systems inoperable, for waterfat the system design flow any reason, Unit 1 reactor ,3 rate (+10% -0%). _

operation and fuel handling and/or

< handling of casks in the vicinity D. Operability of inlet heater at ef the spent fuel pools is 5, g7, e, rated power wheir tested ir. A permissible for a period of seven accortlance with(#tSI N510-197 g g (7) days provided that all active components in the remaining c. Air distr ution is unifo within operable ~ standby gas treatment 0% across the filter trai when systems in each unit (minimum of 1 t sted in a ordance with in Unit 1 and 1 in Unit 2) shall be O'\

NS 0-1975.

demonstrated to be operable within ,

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and daily thereaf ter.

O See .O r3%s,,,, g _

= b .TT.5 : 3. L. M' Le 2-

. NorE I to r TS L C.7 T %54 , ; ,, 3 e, y y

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sore 2 r. xrs f.5 7 A,\

O _

HATCH - UNIT 1 3.7-10b Amendment No. 118 M4

Spc$r'ca,% CS. ]

_ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

-. 7.B.2. Perfonnance Recuirements 2. Filter Testino _

a. The results of the.in-place DOP and 'a. The tests and analysis shall be performed at least once per halogenated hydrocarbon tests at g, p,7, a design flows on HEPA filters and operating cycle, not to exceed 18 charcoal absorber banks shall show months, or after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of f,f 7. /r 99% DOP removal and 99% halogenated S *I '

system operation, or following hydrocarbon removal when tested in painting, fire or chemical release '

in any ventilation zone accordance with(ANSI N510-mb. '

connunicating with the system.

b. The results of laboratory carbon #

sample analysis shall show of b. DOP testing shall be performed f,f.7. C radioactive methyl iodine val af ter each complete or partial when tested in accordance wtth replacement of the HEPA filter bani g ,g RDT-M16-lT (80*C, 95% R.H.).] [S7 or af ter any structural maintenance' on the system housing.

c. Fans shall be shown to ooerate -

CWithin Huz -05 desian floWIwnen c. Halogenated hydrocarbon testing g,g,7,d tested in accordance withfJiSI] shall be performed after each QS10-1975. j [.f *7 complete or partial replacement of the charcoal absorber bank or after any structural maintenance on the OM system housing.

6 Each circuit shall be operated with]

I the heaters on at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month. ,

Nc Mussregaf O

IT'5: 3 6.Y.[

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HATCH - UNIT 1 3.7-11 a Amendment No. 118 Sv'f 4

l 1

SpdNeaTlm C.S7 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.12.A.2. Performance Recuirements 4.12.A.2. Filter Testina

a. The results of the in-place a. The tests and analysis DOP and halogenated hydro- shall be performed at S 7* 4 carbon tests at design flows least once per operating on HEPA filters and charcoal f,8. 7 cycle, not to exceed 18 f,f,7,[ absorber banks shall show months, or af ter every 199-percent DOP removal and 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system opera-199-percent halogenated '

tion or following painting, hydrocarbon removal, respec- fire or chemical release in tively when tested in accordance any ventilation zone communi -

with(ANSI M510-1975.K cating with the system.

~

b. The results of laboratory p,t b. DOP testing shall be per-carbon sampJs, analysis f formed af ter each complete shall show[>90-percenfradie-l or partial replacement of f.f,7. C active methya iocice f,8. 7 the HEPA filter bank or removal when test _ed in after any structural acenrdancewith>RDI-Filb-IT)_ maintenance on the system 25'C, 95-percent R.H.) d l housing.

c. Fans shall be shown to c. Halogenated hydrocarbon operate within +10-percent l testing shall be performed 6.5.7. d desian flow wheii tested in accordance wit f f,7 af ter each complete or G510-1973. ,

partial replacement of the charcoal adsorber bank of f af ter any structural b* maintenance on the system housing.

O gore 2.

3 f. r. , 3

8. Isolation Valve Operability and B. Isolation Valve Testina Closino Time t (Deleted) :

(Deleted) - >

i O ,

HATCH - UNIT 1 3.12-2 Amendment.No. 22 U , 156 Ny

f DISCUSSION OF CHANGES (3j ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP)

TECHNICAL CHANGES - MORE RESTRICTIVE M.1 The current Technical Specifications use laboratory test standard RDT-M16-IT (80 C and 95% relative humidity for SGT System, and 25 C and 95%

relative humidity for MCREC System) for testing the charcoal in the SGT and MCREC Systems. Proposed ITS 5.5.7.c requires laboratory testing in accordance with ASTM D3803-1989 at a terrperature 5 30 C and 2 70% relative humidity for SGT and 95% relative humidity for MCREC. The ASTM D3803-1989 testing standard is more conservative than the current RDT-M16-IT standard and is endorsed by the NRC for use throughout the industry.

M.2 CTS for SGT require laboratory carbon sample analysis to show 90% methyl iodide removal, which is equivalent to 10% penetration. The proposed penetration acceptance criterion is 0.2% penetration. This penetration acceptance criterion will ensure the adsorber efficiency assumed in the accident analyses is maintained.

CTS for MCREC require laboratory carbon sample analysis to show 90% methyl iodide removal, which is equivalent to 10% penetration. The proposed penetration acceptance criterion is 2.0% penetration. This penetration acceptance criterion will ensure the adsorber efficiency assumed in the accident analyses is maintained.

I

(/ The laboratory testing acceptance criteria for both SGT and MCREC are more restrictive than CTS.

,n G/

HATCH UNIT 1 2 REVISION

_ - . - . . . . . . . - . - - -. - _ . . - .- . . . - - - - . . = . - _ -

i DISCUSSION OF CHANGES ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM.(VFTP) ,

i TECHNICAL CHANGE - LESS RESTRICTIVE ;

" Generic" -

LA.1 Details of the methods for implementing this specification are relocated to the FSAR and procedures. Additionally, changes to the procedures and !

the FSAR are controlled in accordance with 10 CFR 50.59. j i

" Specific"

.l L.1 Comment number not used. l t

i L.2 The current Technical Specifications require testing of the SGT System 1) l after any structural maintenance on the HEPA filter or charcoal adsorber :

housings, or 2) following painting, a fire or chemical release in any !

ventilation zone communicating with the system. Plant Hatch has performed >

tests and evaluations and has determined that the use of water based i paints and the performance of metal grinding, buffing or welding are not '

detrimental to the charcoal filters of the SGT System, either prior to or O during operation. These activities should not require surveillance of the SGT system upon their conclusion. This applies to all types of welding conducted at Plant Hatch and tracking of the quantity of weld material i used is not necessary. !

b f

1 i

O l 1

HATCH UNIT 1 3 REVISIONl[ -l

?

i ADMINISTRATIVE CONTROLS M8 8 N'b ^ ad I

(Vl .C 6 8+ MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown eroerience shall be submitted on a monthly batitIto tne uirector, Ottice or rianagement ano rro ram analysis, U. 5. Nuclear Regulatory Commission, Washington, D. C. 2 555, with a copy to the Recional Office of insnection and A*g , Enforromants no later than the 15th of each month following the calendar month covered by the report.

gJ, y CORE OPERATING LIMITS REPORT f 6,,,[,4 6.9.1.ll.a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

[.E. 8 e **b (1) Operation with a Limiting Control Rod Pattern (for Rod Withdrawal Error. RWE) for Specification 3.3.F.

f,g,, ,f, a ,2.) (2) The Average Planar Linear Heat Generation Rate (APLHGR) for Specification 3.II.A, Linea \HeatGenera on Rate (LH% for Specifft4 tion t 3.1 f,4'f*A.p(4) The Minimum Critical Power Ratio (MCPR) for Specifications 3.3.F and 3.11.C and Surveillance Requirement 4.ll.C.

f" 6. f, /r b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the following documents.

O f ,f, J . /y, h (1) NEDE-240ll-P-A, " General Electric Standard Application for Reactor l'

Fuel," (applicable amendment specified in the CORE OPERATING LIMITS REPORT).

f*f f , / ,2.) (2) " Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 157 to Facility Operating License DPR-57,"

dated September 12, 1988.

f,6, j', d* c. The core, operatino limits chall be determined <n that all annlicable 1%its (t:.g., f uel thirmal-mecnantchi limits', ~cornnairmal-rtyttraultc D limits, LECS limits, nuclear limits swch at shut 4 % margin. hnd transiant acd accident acalysis limits) Df the setecy unnysts are met.

i f,4, ,f, J d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be erovided upon issuanea for each reload cycle, to the NorrDocument Control Desk with copies to the Regi dministrator and Resident Inspector.

Y A

m HATCH - UNIT 1 6-15d Amendment No. H 4, 49, 48, 190 hof Io

(This page intentionally left blank.)

O O HATCH UNIT ITS 6.G.G of IO

G o

% J y NOTES FOR T ABLE 3.2-11 (Continuedi M

x g.1. With the plant in the power operation, etertup. or hot shutdown condition and with the number of h operable channels less then the required operable channele. Initiete the peoplanned ofternete rnethod of enonitoring the appropriate peremeter within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months end:

Q

e. either ts:te= the Inoperable chenpolie) to operable statue within 7 days of the event or l

'# propero end subrret e special toport to the NRC pursuant to Specification 6.9.2. within 14 deye t Sr mgf a ,, ,f

b. l b* following the event outfiedng the action taken the cause of the inoperability. and the plane c

~

end schedule for toetoring the eyetem to operable statue, b D 3, y Ig.2. One instrument ebennel mey be inoperable for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to perform required survolliances prior to entering other applicable setione.

INwwy#> 84 .

$h boo 3.]

h. A channel contains two detectors: one for mid-range noble gos. and one for high tenge noble gee.

Both detectors must be operable to consider the chonnel operable.

I. Instrumentetion ohes be operable with continuove earnpling capabluty within 30 rninutes of en ECCS actuotion during a LOCA. See Section 3.7.A.S.o for the LIMITING CONDITION FOR OPERATION.

'?

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e O.

. 4 e+

g O @

C

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.$ t 2

~

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- e rw E

DISCUSSION OF CHANGES .

ITS: SECTION 5.6 - REPORTING REQUIREMENTS TECHNICAL CHANGES - MORE RESTRICTIVE !

M.1 The current TS requirement in 6.9.1.5 b to submit an annual report for all l challenges to safety / relief valves has been moved to proposed ITS 5.6.1.4 i for monthly reports. Since the report is required on a monthly basis :

instead of the current annual basis, this change is more restrictive in !

nature.

M.2 This change details the information to be included in the report. These details are necessary to assure the reports are provided with similar content and format for comparison with other plants and with prior reports.

TECHNICAL CHANGE - LESS RESTRICTIVE I

" Generic" LA.1 The details associated with CTS 6.9.1.1, 6.9.1.2, and 6.9.1.3, " Start-Up Report," are proposed to be relocated to the FSAR. The Start-Up Report !

provides the NRC a mechanism to review the appropriateness of licensee !

activities after-the-fact, but provides no regulatory authority once the report is submitted (i.e., no requirement for NRC approval). The Quality Assurance requirements of 10 CFR 50, Appendix B and .the Startup Test i Program provisions contained in the FSAR provide assurance the listed ;

activities will be adequately performed and that appropriate corrective j actions, if required, are taken. The placement of these CTS requirements in the FSAR also ensures that change control is performed in accordance .

with 10 CFR 50.59. !

l I

O HATCH UNIT 1 4 REVISION [b

UNIT 1 NO SIGNIFICANT HAZARDS DETERMINATION l

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NO SIGNIFICANT HAZARDS DETERMINATION .

l ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM L.4 CHANGE Not used.

O 1

HATCH UNIT I 5 REVISION [h

a _ _ - ---

NO SIGNIFICANT HAZARDS DETERMINATION O ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP) j I

-L.1 CHANGE l

Not used, i i

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HATCH UNIT I 1 REVISION (( l l

. _ . _ _ . . . _ . _ __ _ . . - _ - _ . . _ . . . . . _ . _ . . . _ J

O UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS O

O

Definitions 1.1 1.1 Definitions (continued)

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested. ,

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation (s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in O Table 1.1-1 with fuel in the reactor vessel.

OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function (s) and when all necessary attendant instrumentation, controls, normal or energency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, -

component, or device to perform its specified safety function (s) are also capable of performing their related support function (s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Chapter 14, Initial Tests and Operation, of the FSAR; (continued)

HATCH UNIT 2 1.1-5 REVISION A

Definitions 1.1 1.1 Definitions PHYSICS TESTS b. Authorized under the provisions of (continued) 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2436 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:

a. The reactor is xenon free;

b. The moderator temperature is 68 F; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

(continued)

HATCH UNIT 2 1.1-6 REVISION 1

. - - . = - -

PAM Instrumentation 3.3.3.1 t

() 3.3 INSTRUMENTATION 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1-1 .

shall be OPERABLE.

i i

APPLICABILITY: MODES 1 and 2.

ACTIONS

NOTES------------------------------------ .

1. LC0 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each Function.

i CONDITION REQUIRED ACTION COMPLETION TIME f

\-

A. One or more Functions A.1 Restore required 30 days i with one required channel to OPERABLE channel inoperable. status.

l B. Required Action and B.1 Initiate action in Immediately l associated Completion accordance with i Time of Condition A Specification 5.6.6. l ;

not met. ,;

i t

C. One or more Functions C.1 Restore all but one 7 days t with two or more ccquired thannel to l required channels OPERABLE status. .

inoperable. '

(continued) ;

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O V

HATCH UNIT 2 3.3-23 REVISION [3 I

PAM Instrumentation 3.3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C Table 3.3.3.1-1 for not met. the channel.

E. As required by E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action D 1 and referenced in Table 3.3.3.1-1.

F. As required by F.1 Initiate action in Immediately Required Action D.1 accordance with and referenced in Specification 5.6.6. l Table 3.3.3.1-1.

O SURVEILLANCE REQUIREMENTS

_------------------NOTES------------------------------------

1. These SRs apply to each Function in Table 3.3.3.1-1.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required '

Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the other required channel (s) in the associated Function is OPERABLE.

SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK. 31 days SR 3.3.3.1.2 Perform CHANNEL CALIBRATION. 18 months O

HATCH UNIT 2 3.3-24 REVISION

I ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS ;

NOTES------------------------------------ ,

1. Refer to Taole 3.3.5.1-1 to determine which SRs apply for each ECCS !

Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions 3.c i and 3.f; and (b) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for Functions other than 3.c and 3.f i provided the associated Function or the redundant Function maintains l initiation capability. .

4 t

SURVEILLANCE FREQUENCY (

r SR 3.3.5.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months t

SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days l i

O SR 3.3.5.1.3 Perform CHANNEL CALIBRATION. 92 days t

i SR 3.3.5.1.4 Perform CHANNEL CALIBRATION. 18 months SR 3.3.5.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months !

i SR 3.3.5.1.6 Verify the ECCS RESPONSE TIME is within 18 months on a limits. STAGGERED TEST i

BASIS '

HATCH UNIT 2 3.3-39 REVISION l

l

ECCS Instrumentation 3.3.5.1 Table 3.3.5.1 1 (pope 1 of 6)

Emergency Core Cooling System Instrunentation APPLICABLE CONDITIONS MCOES REQUIRED REFERENCED OR OTHER CHANNELS FRM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

1. Core Spray System

a. Reactor Vessel Water 1,2,3, 4(b) B SR 3.3.5.1.1 2 113 incnes Level - Low Low Low, SR 3.3.5.1.2 Level 1 4(8), 5(*) SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

b. Drywett 1,2,3 4(b) B SR 3.3.5.1.1 s 1.92 psig Pressure - High SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.5 SR 3.3.5.1.6

c. Reactor Steam Dome 1,2,3 4 C SR 3.3.5.1.1 t 390 psig Pressure - Low SR 3.3.5.1.2 and (Injection Permissive) SR 3.3.5.1.4 s 476 psig SR 3.3.5.1.5 SR 3.3.5.1.6 4(a), 5(a) 4 B SR 3.3.5.1.1 2 390 psig SR 3.3.5.1.2 and SR 3.3.5.1.4 s 476 psig SR 3.3.5.1.5 SR 3.3.5.1.6

d. Core Spray Purp 1,2,3, 1 per E SR 3.3.5.1.1 2 570 ppm Discharge Flow - Low subsystem SR 3.3.5.1.2 and (Bypass) 4(a), $(a) SR 3.3.5.1.4 5 745 spm SR 3.3.5.1.5

2. Low Pressure Coolant Injection (LPCI) System

a. Reactor vesset Water '1,2,3, 4(b) B SR 3.3.5.1.1 2 113 inches Level - Low Low Low, SR 3.3.5.1.2 Level 1 4(a), $(a) SR 3.3.5.1.4 SR 3.3.5.1.5 .

SR 3.3.5.1.6 (continued)

(a) When associated subsystem (s) are required to be OPERABLE.

(b) Also required to initiate the associated diesel generator (DC) and isolate the associated plant service water (PSW) turbine building (T/B) isolation valves.

O' HATCH UNIT 2 3.3-40 REVISION A

LOP Instrumentation 3.3.8.1 3.3 INSTRUMENTATION 3.3.8.1 Loss of Power (LOP) Instrumentation LCO 3.3.8.1 The LOP instrumentation for each function in Table 3.3.8.1-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, When the associated diesel generator (DG) is required to be OPERABLE by LC0 3.8.2, "AC Sources -- Shutdown."

ACTIONS

NOTE-------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Restore channel to I hour (x inoperable for OPERABLE status.

Functions 1 and 2.

B. One or more channels B.1 Verify voltage on Once per hour inoperable for associated 4.16 kV Function 3. bus is 2 3825 V.

1 l

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C. Required Action and C.1 Declare associated DG Immediately l l associated Completion inoperable.

Time not met.

1 HATCH UNIT 2 3.3-69 REVISION C

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LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS hl

NOTE-------------------------------------

1. Refer to Table 3.3.8.1-1 to determine which SRs apply for each LOP Function.

2. When a 4.16 kV Emergency Bus Undervoltage channel is placed in an inoperable status solely for performance of required Surve111ances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains initiation capability l (for Functions 1 and 2) and annunciation capability (for function 3).

SURVEILLANCE FREQUENCY SR 3.3.8.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.8.1.2 Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.8.1.3 Perform CHANNEL CAllBRATION. 18 months SR 3.3.8.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months i

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HATCH UNIT 2 3.3-70 REVISION p {

1 l

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RCS P/T Limits 3.4.9 m

() 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (P/T) Limits LC0 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within limits.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1 Restore parameter (s) 30 minutes Required Action A.2 to within limits.

shall be completed if this Condition is AND entered.

A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for

(' Requirements of the continued operation.

LC0 not met in e MODES 1, 2, and 3.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion ,

Time of Condition A AND '

i not met. <

B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

O~~-

HATCH UNIT 2 3.4-21 REVISION ((fI

RCS P/T Limits 3.4.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE--------- C.I Initiate action to Immediately Required Action C.2 restore parameter (s) shall be completed if to within limits.

this Condition is entered. AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LCO not met in other cperation. or 3 l than MODES 1, 2, and 3.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 -------------------NOTE--------------------

O Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify:

30 minutes

a. RCS pressure and RCS temperature are within the limits specified in Figures 3.4.9-1 and 3.4.9-2; and

b. RCS heatcp and cooldown rates are s 100 F in any I hour period.

(continued) l 9

HATCH UNIT 2 3.4-22 l REVISION [

l

l RCS P/T Limits 3.4.9

,r8 l t ) SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY l Il SR 3.4.9.2 Verify RCS pressure and RCS temperature are Once within !

within the criticality limits specified in 15 minutes '

Figure 3.4.9-3. prior to l control rod withdrawal for the purpose of achieving criticality (continued) I

)

U 1

I (J \

l HATCH UNIT 2 REVISION 3.4-2%k

i RCS P/T Limits l 3.4.9

( SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.3 --------------------NOTE-------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the bottom 15 minutes head coolant temperature and the reactor pressure vessel (RPV) coolant temperature is s 145 F. l SR 3.4.9.4 -------------------NOTE--------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the reactor 15 minutes

(')

(s coolant temperature in the recirculation loop to be started and the RPV coolant temperature is s 50*F.

SR 3.4.9.5 -------------------NOTE--------------------

Only required to be performed when tensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head 30 minutes flange temperatures are 2 90 F. l (continued)

(vD HATCH UNIT 2 3.4-243 j REVISION 1

.j

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.6 -------------------NOTE--------------------

Not required to be performed until 30 minutes after RCS temperature s 100 F in l MODE 4.

Verify reactor vessel flange and head 30 minutes flange temperatures are 2 90 F. l SR 3.4.9.7 -------------------NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature s 120 F in MODE 4. l Verify reactor vessel flange and head 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months f1ange temperatures are 2 90 F. l 9

O HATCH UNIT 2 3.4-2f h REVISIONf h

RCS P/T Limits -

3.4.9

(

1600 NA I 1400 .

8 i

8 , i

^ .

.?

1 r

ef E o

'f l

1200~ i ,

o 8f b

z ,

c 1 .

4 g i r 8 1000 ll d

w .

U A' - CORE BELTUNE N AFTER ASSUMED 32 EFPY g 800 shirr rROM AN INITIAL O WELD RTcOF -S0*F b

6 E /

O' z 600 A - SYSTEM HYDROTEST UMIT wiTH rUEL IN VESSEL ;

t- ;

E

- VE5SEL DISCONTINUITY ,

E 400 uuns '

D

= - CORE BELTLINE WITH i in 312 Psc -- 32 EFPY SHIFT l a.

80LTUP CURvT A* 15 NOT UMITING ; ;

e 93 r FOR INFORMATION ONLY !

I CURVE A 15 VAUD FOR i 32 EFPY OF OPERATION '

O , , i 0 100 200 300 400 500 600 :

MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F) l i

Figure 3.4.9-1 (page 1 of I) !

Temperature / Pressure Limits for Inservice Hydrostatic and Inservice Leakage Tests l

HATCH UNIT 2 3.4-)4 $ 4 A REVISION [6

RCS P/T Limits 3.4.9 9

1600 l

B' 8 1400 -

I

l w

2 'I 1200 O

m ,f Z s L

S 1000 [

d w

l M B' - CORE BELTLINE N AFTER ASSUuCD 32 EFPY e 800 shirt rROu AN INITIAL O WELD RTecnOF -50'F U

b z 600 B - NON-NUCLEAR HEATUP/

CooloOwN uurr W 6 )

a

- VESSEL DISCONTINUITY

' uunS 400 D I - - CORE BELTLINE WITH w

m2 esc J 32 EFPY SHIFT

/

B3LTUP CURVE B' IS NOT UMITING 90*r ,

FOR INFORMATION ONLY CUWE B 15 VAUD FOR 32 EFPY OF OPERATION O , , ,

0 100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE ff)

Figure 3.4.9-2 (page 1 of 1)

Temperature / Pressure Limits for Non-Nuclear Heatup, Low Power Physics Tests, and Cooldown Following a Shutdown HATCH UNIT 2 3.4-N. 54 b REVISION [E l

l

i RCS P/T Limits l 3.4.9 l O l 1600 '

l I,

C' C 1400 !

i 9 8

} 1200 -

l e

o

b z s' i

n. 1000 ,

? '

s --

W C' - CORE BELTLINE gn AFTER ASSUMED 32 EFPY p 800 shirt rROM AN INITIAL WELD RTunOF -50*F M

O t b i O' 6 x 600 C - NUCLEAR (CORE CRITICAL) uurr t

3 i '

t- l l 2 f

- VESSEL DISCONTINUITY 400 y / - = CORE BELTLINE WITH l ,

D tn 312 esc ) 32 ErPY SHIFT i 1

a / .

gggp ( CURVE C' 15 NOT LIMITING gre / FOR INFORMATION ONLY

/ CURVE C 15 VAUD FOR 32 EFPY OF OPERATION O , , ,

0 100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F) i l i figure 3.4.9-3 (page 1 of 1) '

Temperature / Pressure Limits for Criticality O

HATCH UNIT 2 3.4-M 2 U REVISION /h

i SGT System--Operating l 3.6.4.7 e f

()

i

.3.6 CONTAINMENT SYSTEMS 3.6.4.7 Standby Gas Treatment (SGT) System -- Operating j i

LCO 3.6.4.7 Two Unit 1 and two Unit 2 SGT subsystems shall be OPERABLE. ;

APPLICABILITY: MODES 1, 2, and 3. f l

ACTIONS

[

___________________________----------NOTE-------------------------------------

When two Unit 1 SGT subsystems are placed in an inoperable status solely for l inspection of the Unit I hardened vent rupture disk, entry into associated Conditions and Required Actions may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , provided .

both Unit 2 SGT subsystems are OPERABLE. !

CONDITION REQUIRED ACTION COMPLETION TIME O A. One Unit 1 or Unit 2 SGT subsystem inoperable.

A.1 Restore SGT subsystem to OPERABLE status, 7 days B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months :

associated Completion

{

Time not met. AND r B.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months j i

1 i

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HATCH UNIT 2 3.6-53 REVISION A

SGT System-Operating 3.6.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.7.1 Operate each required Unit I and Unit 2 SGT 31 days subsystem for 210 continuous hours with heaters operating.

SR 3.6.4.7.2 Perform required Unit I and Unit 2 SGT In accordance filter testing in accordance with the with the VFTP Ventilation Filter Testing Program (VFTP).

l SR 3.6.4.7.3 Verify each Unit I and Unit 2 SGT subsystem 18 months actuates on an actual or simulated initiation signal.

O

}

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HATCH UNIT 2 3.6-54 REVISION,A[

SGT System-0PORVs 3.6.4.8 ;

3.6 CONTAINMENT SYSTEMS 3.6.4.8 Standby Gas Treatment (SGT) System - OPDRVs ;

i LCO 3.6.4.8 Two Unit 2 SGT subsystems shall be OPERABLE. i APPLICABILITY: During operations with a potential for draining the reactor vessel (OPDRVs). .

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i

A. One Unit 2 SGT A.1 Restore SGT subsystem 7 days :

subsystem inoperable. to OPERABLE status. !

B. Required Action and B.1 Place OPERABLE Unit 2 Immediately associated Completion SGT subsystem in i Time not met. ~ operation. }

=

B.2 Initiate action to Immediately suspend OPDRVs.

l C. Two Unit 2 SGT C.1 Initiate action to Immediately -

subsystems inoperable. suspend OPDRVs.

l l

4 f

HATCH UNIT 2 3.6-55 REVISION A

SGT System-0PDRVs 3.6.4.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.8.1 Operate each Unit 2 SGT subsystem for a 10 31 days continuous hours with heaters operating.

SR 3.6.4.8.2 Perform required Unit 2 SGT filter testing In accordance in accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).

SR 3.6.4.8.3 Verify each Unit 2 SGT subsystem actuates 18 months on an actual or simulated initiation signal.

O i

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l e HATCH UNIT 2 3.6-56 REVISION

I i

SGT System-Refueling '

3.6.4.9 SURVEllLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY i

SR 3.6.4.9.1 Operate each required Unit 1 and Unit 2 SGT 31 days i subsystem for 2: 10 continuous hours with heaters operating.

f SR 3.6.4.9.2 Perform required Unit I and Unit 2 SGT In accordance filter testing in accordance with the with the VFTP Ventilation Filter Testing Program (VFTP).

SR 3.6.4.9.3 Verify each required Unit 1 and Unit 2 SGT 18 months !

subsystem actuates on an actual or !

simulated initiation signal.

O !

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i HATCH UNIT 2 3.6-59 REVISION 1

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Ventilation Filter Testino Prooram (VFTP)

The VFTP will establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies ,

specified in Regulatory Guide 1.52, Revision 2, Section 5a and at least once per 18 months or 1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, 2) following painting, fire or chemical release in any ventilation zone communicating with the system, or 3) after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of

charcoal adsorber operation.

NOTES-----------------------------

1. Tests and evaluations have determined the impact on the Standby Gas Treatment (SGT) System filters of certain types of painting, buffing and grinding, and welding. The use of water based paints and the performance of metal grinding, buffing, or welding are not detrimental to the charcoal '

filters of the SGT System, either prior to or during operation. These activities will not require surveillance of the system upon their conclusion. This applies to all ,

types of welding conducted at Plant Hatch, and tracking of the quantity of weld material used is not necessary.

2. For testing purposes, the use of refrigerants equivalent to those specified in ASME N510-1989 is acceptable.

a. Demonstrate for each of the ESF systems that an inplace test of the HEPA filters shows a penetration and system bypass

< 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, Section 5c and ASME N510-1989, Section 10, at the system flowrate specified below.

ESF Ventilation System Flowrate (cfm)

SGT System 3000 to 4000 i Main Control Room Environmental 2250 to 2750 Control (MCREC) System (continued)

O HATCH UNIT 2 5.0-11 REVISION A

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testina Proaram (VFTP) (continued)

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, Section 5d and ASME N510-1989, Section 11, at the system flowrate specified below.

[SF Ventilation System Flowrate (cfm)

SGT System 3000 to 4000 MCREC System 2250 to 2750 1

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.S2, Revision 2, Section 6b and ASME N510-1989, Section 15 and Appendix B, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of s 30 C and greater than or equal to the relative humidity specified below.

ESF Ventilation System Penetration (%) RH(%)

SGT System 0.2 70 MCREC System 2.0 95

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with ASME N510-1989, Section 8.5.1, at the system flowrate specified below. .

ESF Ventilation System 6P (inches wa) Flowrate (cfm)

SGT System <6 3000 to 4000 MCREC System <6 2250 to 2750

e. Demonstrate that the heaters for the ESF system dissipate the value specified below when testing in accordance with ASME N510-1989, Section 14.5.1.

ESF Ventilation System Wattaae (kW1 SGT System 17 to 20 (continued)

HATCH UNIT 2 5.0-12 REVISION

\

Programs and Manuals l 5.5 )

i

[~5 s 5.5 Programs and Manuals

} ,

l 5.5.7 Ventilation Filter Testina Proaram (VFTP) (continued) i The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

O O

o

(continued)

HATCH UNIT 2 5.0-)4 I2A REVISION [{

i Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1) Control Rod Block Instrumentation - Rod Block Monitor for Specification 3.3.2.1.

2) The Average Planar Linear Heat Generation Rate for !

Specification 3.2.1. ;

3) The Minimum Critical Power Ratio for Specifications ,

3.2.2 and 3.3.2.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by ;

the NRC, specifically those described in the following ,

documents: ,

1) NEDE-24011-P-A, " General Electric Standard Application l for Reactor Fuel," (applicable amendment specified in the COLR). ,

2) " Safety Evaluation by the Office of Nuclear Reactor .

Regulation Supporting Amendment Nos.151 and 89 to ,

Facility Operating Licenses DPR-57 and NPF-5," dated !

January 22, 1988. :

c. The core operating limits shall be determined such that all i applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits and accident analysis limits) of the safety '!

analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, e shall be provided upon issuance for each reload cycle to the NRC. j (continued)

~

HATCH UNIT 2 5.0 26 ! j REVISIONp[

7

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) l 5.6.6 Post Accident Monitorino (PAM) Instrumentation Report l When a report is required by LCO 3.3.3.1, " Post Accident Monitoring (PAM) instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

i l

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n m a usi n 5. % Rms10sje l

UNIT 2 IMPROVED BASES

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O l

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PAM Instrumentation B 3.3.3.1 l l

() BASES LCO 12. RHR Service Whter Flow (continued) primary indication used by the operator during an accident.

Therefore, the PAM specification deals specifically with !

this portion of the instrument channel. l l

APPLICABILITY The PAM instrumentation LCO is applicable in MODES 1 and 2. .

These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES I and 2. In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES. l t

ACTIONS Note I has been added to the ACTIONS to exclude the MODE change restriction of LCO 3.0.4. This exception allows entry into the applicable MODE while relying on the ACTIONS O

U even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the passive ,

function of the instruments, the operator's ability to '

diagnose an accident using alternative instruments and methods, and the low probability of an event requiring these instruments.

1 Note 2 has been provided to modify the ACTIONS related to PAM instrumentation channels. Section 1.3, Completion '

Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables '

expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable PAM instrumentation channels provide appropriate '

compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable PAM Function.

1 (continued)

HATCH UNIT 2 B 3.3-67 REVISION A

PAM Instrumentation B 3.3.3.1 BASES ACTIONS M (continued)

When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channels (or, in the case of a Function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

M If a channel has not been restored to OPERABLE status in 30 days, this Required Action specifies initiation of action in accordance with Specification 5.6.6, which requires a l written report to be submitted to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions.

This action is appropriate in lieu of a shutdown W

requirement, since alternative actions are identified before loss of functional capability, and given the likelihood of plant conditions that would require information provided by this instrumentation.

C.1 When one or more Functions have two or more required -

channels that are inoperable (i.e., two channels inoperable in the same Function), all but one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation ,

and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function is not acceptable i because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoratian of one j inoperable channel of the Function limits the risk that the I (continued) i HATCH UNIT 2 B 3.3-68 REVISIONf{

l 1

PAM Instrumentation l B 3.3.3.1 BASES ACTIONS L1 (continued)

PAM Function will be in a degraded condition should an '

accident occur.

D.i.1 !

This Required Action directs entry into the appropriate :

Condition referenced in Table 3.3.3.1-1. The applicable Condition referenced in the Table is Function dependent. '

Each time an inoperable channel has not met the Required Action of Condition C, and the associated Completion Time has expired, Condition D is entered for that channel and provides for transfer to the appropriate subsequent Condition.

L.1 For the majority of Functions in Table 3.3.3.1-1, if any Required Action and associated Completion Time of Condition C is not met, the plant must be brought to a MODE m in which the LCO not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

L.1 .

Since alternate means of monitoring drywell area radiation -

have been developed and tested, the Required Action is not -

to shut down the plant, but rather to follow the directions of Specification 5.6.6. These alternate means may be l temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels. i O (continued)

I HATCH UNIT 2 B 3.3-69 REVISION k I

l l

PAM Instrumentation l B 3.3.3.1 l BASES (continued) h SURVEILLANCE As noted at the beginning of the SRs, the following SRs l l REQUIREMENTS apply to each PAM instrumentation Function in j

Table 3.3.3.1-1. !

l The Surveillances are modified by a second Note to indicate ,

that when a channel is placed in an inoperable status solely !

for performance of required Surveillances, entry into '

l associated Conditions and Required Actions may be delayed l for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the other required channel (s) in the associated Function are OPERABLE. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.

The Note is based upon a NRC Safety Evaluation Report (Reference 1) which concluded that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability of properly monitoring post accident parameters, when necessary.

SR 3.3.3.1.1 l

Perfcrmance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter l indicate 6 on one channel against a similar parameter on other chaenels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the -

instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indicaticn l

that the sensor or the signal processing equipment has I drifted outside its limit.

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1 (continued) l l

l HATCH UNIT 2 B 3.3-70 REVISION A !

_ _ _ _ _ _ _ _ _ _ _ _-___--.__-.-\

ECCS Instrumentatien B 3.3.5.1 i

BASES ACTIONS F.1 and F.2 (continued) it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition H must be entered and its Required Action taken.

G.1 and G.2 Required Action G.] is intended to ensure that appropriate actions are taken if multiple, inoperable channels within similar ADS trip system functions result in automatic initiation capability being lost for the ADS. In this situation (loss of automatic initiation capability), the !

96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months or 8 day allowance, as applicable, of Required Action G.2 is not appropriate, and all ADS valves must be l declared inoperable within I hour after discovery of loss of ADS initiation capability.

The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This O

V Completion Time also allows for an exception to the normal

" time zero" for beginning the allowed outage time " clock."

i For Required Action G.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable channels within similar ADS trip system Functions as described in the paragraph above. The I hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

Because of the diversity of sensors available to provide -

initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE (Required Action G.2). If either HPCI or RCIC is inoperable, the time shortens to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . If the status of i HPCI or RCIC changes such that the Completion Time changes I from 8 days to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , the 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable channel cannot exceed 8 days. If the status ,

of HPCI or RCIC changes such that the Completion Time changes from 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months to 8 days, the " time zero" for i beginning the 8 day " clock" begins upon discovery of the l (continued)

HATCH UNIT 2 B 3.3-129 REVISION A

l ECCS Instrumentation l B 3.3.5.1 BASES h'

ACTIONS G.1 and G.2 (continued) inoperable channel. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events.

lid With any Required Action and associated Completion Time not met, the associated feature (s) may be incapable of performing the intended function, and the supported feature (s) associated with inoperable untripped channels must be declared inoperable immediately.

SURVEILLANCE As noted in the beginning of the SRs, the SRs for each ECCS REQUIREMENTS instrumentation Function are found in the SRs column of Table 3.3.5.1-1.

The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months as follows: (a) for Functions 3.c and 3.f; and (b) for Functions other than 3.c and 3.f provided the associated Function or the redundant Function maintains l initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be -

returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 5) assumption of the average time required to perform channel surveillance. That l analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary.

(continued)

HATCH UNIT 2 B 3.3-130 REVISION [

Pricary Centainment Isolation Instrumentaticn B 3.3.6.1 BASES BACKGROUND 1. Main Steam Line Isolation (continued)

MSL Isolation Functions isolate the Group 1 valves. l

2. Primary Containment Isolation i

Most Primary Containment Isolation Functions receive inputs :

from four channels. The outputs from these channels are !

arranged into two two-out-of-two logic trip systems. One i trip system initiates isolation of all inboard primary l' containment isolation valves, while the other trip system initiates isolation of all outboard primary containment l isolation valves. Each logic closes one of the two valves on each penetration, so that operation of either logic ,

isolates the penetration. The TIP ball valves isolation ,

does not occur until the TIPS have been fully retracted (The ,

logic also sends a TIP retraction signal). l The exception to this arrangement is the Drywell l Radiation-High function. This Function has two channels, !

whose outputs are arranged in two one-out-of-one logic trip I systems. Each trip system isolates one valve per associated O penetration, similar to the two-out-of-two logic described above.

l Primary Containment Isolation Drywell Pressure-High and Reactor Vessel Water Level-Low, Level 3 Functions isolate the Group 2, 6, 7,10, and 12 valves. Reactor Building and Refueling Floor Exhaust Radiation-High Functions isolate the Group 6, 10, and 12 valves. Primary Containment Isolation Drywell Radiation-High Function isolates the 18 inch containment purge and vent valves. -

3. 4. Hioh Pressure Coolant In.iection System Isolation and Reactor Core Isolation Coolina System Isolation Most Functions that isolate HPCI and RCIC receive input from two channels, with each channel in one trip system using a one-out-of-one logic. Each of the two trip systems in each isolation group is connected to one of the two valves on each associated penetration.

(continued)

HATCH UNIT 2 B 3.3-147 REVISION A

Primary Containment Isolation Instrumentation B 3.3.6.I BASES h

BACKGROUND 3. 4. Hiah Pressure Coolant Iniection System Isolation and Reactor Core Isolation Coolina System Isolation (continued)

The exceptions are the HPCI and RCIC Turbine Exhaust Diaphragm Pressure - High and Steam Supply Line Pressure - Low Functions. These Functions receive inputs from four turbine exhaust diaphragm pressure and four steam supply pressure channels for each system. The outputs from the turbine exhaust diaphragm pressure and steam supply pressure channels are each connected to two two-out-of-two trip systems. Additionally, each trip system of the Steam Line Flow - High Functions receives input from a low differential pressure channel. The low differential pressure channels are not required for OPERABILITY. Each trip system isolates'one valve per associated penetration.

HPCI and RCIC Functions isolate the Group 3, 4, 8, and 9 valves.

5. Reactor Water Cleanuo System Isolation The Reactor Vessel Water Level - Low Low, Level 2 Isolation Function receives input from four reactor vessel water level ,

channels. The outputs from the reactor vessel water level channels are connected into two two-out-of-two trip systems.

The Area Temperature - High Function receives input from six temperature monitors, three to each trip system. The Area Ventilation Differential Temperature - High Function receives input from six differential temperature monitors, three in each trip system. These are configured so that any one input will trip the associated trip system. Each of the two trip systems is connected to one of the two valves on -

the RWCU penetration. However, the SLC System Initiation function only provides an input to one trip system, thus closes only one valve.

RWCU Functions isolate the Group 5 valves.

6. RHR Shutdown Coolina System Isolation The Reactor Vessel Water Level - Low, Level 3 function receives input from four reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected to two two-out-of-two trip systems.

(continued)

HATCH UNIT 2 B 3.3-148 REVISION

Prinary Containment Isolatien Instrumentaticn B 3.3.6.1 ,

l 1

BASES APPLICABLE 2.d. 2.e. Reactor Buildina and Refuelino Floor Exhaust SAFETY ANALYSES, Radiation-Hich LCO, and APPLICABILITY High secondary containment exhaust radiation is an (continued) indication of possible gross failure of the fuel cladding.

The release may have originated from the primary containment due to a break in the RCPB. When Exhaust Radiation-High is detected, valves whose penetrations communicate with the primary containment atmosphere are isolated to limit the release of fission products.

The Exhaust Radiation-High signals are initiated from radiation detectors that are located near the ventilation exhaust ductwork coming from the reactor building and the refueling floor zones, respectively. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Four channels of Reactor Building Exhaust-High Function and four channels of Refueling Floor Exhaust-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are chosen to ensure radioactive O releases do not exceed offsite dose limits.

These Functions isolate the Group 6,10, and 12 valves.

Hich Pressure Coolant Iniection and Reactor Core Isolation Coolina Systems Isolation 3.a. 4.a. HPCI and RCIC Steam Line Flow-Hiah Steam Line Flow-High Functions are provided to detect a break of the RCIC or HPCI steam lines and initiate closure of the steam line isolation valves of the appropriate system. If the steam is allowed to continue flowing out of the break, the reactor will depressurize and the core can uncover. Therefore, the isolations are initiated on high flow to prevent or minimize core damage. The isolation action, along with the scram function of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Specific credit for these Functions is not assumed in any FSAR accident analyses since the bounding analysis is performed for large breaks such as

( (continued) !

HATCH UNIT 2 B 3.3-157 REVISION A

l Primary Containment Isolation Instrumentation l B 3.3.6.I BASES h 1 l

APPLICABLE 3.a. 4.a. HPCI and RCIC Steam Line Flow -- Hioh i SAFETY ANALYSES, (continued)

LCO, and )

APPLICABILITY recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding.

The HPCI and RCIC Steam Line Flow - High signals are initiated from transmitters (two for HPCI and two for RCIC) that are connected to the system steam lines. Two channels of both HPCI and RCIC Steam Line Flow - High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are chosen to be low enough to ensure that the trip occurs to prevent fuel damage and maintains the MSLB event as the bounding event. The Allowable Values correspond to s 200 inches water column for HPCI and s 139 inches water column for RCIC, which are the parameters monitored on control room instruments.

These Functions isolate the Group 3 and 4 valves, as appropriate.

3.b. 4.b. HPCI and RCIC Steam Supply line Pressure - Low Low MSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue operation of the associated system's turbine. These isolations are for equipment protection and are not assumed in any transient or accident analysis in the FSAR. However, -

they also provide a diverse signal to indicate a possible system break. These instruments are included in Technical Specifications (TS) because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations. Therefore, they meet Criterion 4 of the NRC Policy Statement (Ref. 7).

The HPCI and RCIC Steam Supply Line Pressure - Low signals are initiated from transmitters (four for HPCI and four for RCIC) that are connected to the system steam line. Four channels of both HPCI and RCIC Steam Supply Line Pressure - Low Functions are available and are required to (continued)

HATCH UNIT 2 B 3.3-158 REVISION

i LOP Instrumentation B 3.3.8.1

() BASES ACTIONS a1 With one or more channels of Function 1 or 2 inoperable, the Function does not maintain initiation capability for the associated emergency bus. Therefore, only I hour is allowed ,

to restore the inoperable. channel to OPERABLE status. The Required Action does not allow placing a channel in trip ,

since this action will result in a DG initiation. ,

f 4

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4

() (continued) HATCH UNIT 2 B 3.3-203B l REVISIONf[3

LOP Instrumentation B 3.3.8.1 BASES h ACTIONS A.1 (continued) The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channel s . IL1 Each 4.16 kV bus has a dedicated annunciator fed by two relays and their associated time delays in a one-out-of-two logic configuration. Only one relay and its associated time delay is required to be OPERABLE. Therefore, the loss of the required relay or time delay renders Function 3 incapable of performing the intended function. Since the intended function is to alert personnel to a lowering voltage condition and the voltage reading is available for each bus on the control room front panels, the Required Action is verification of the voltage to be above the annunciator setpoint (nominal) hourly. u O If any Required Action and associated Completion Time are not met, the associated Function does not maintain initiation capability for the associated emergency bus. Therefore, the associated DG(s) is declared inoperable immediately. This requires entry into applicable Conditions and Required Actions of LC0 3.8.1 and LC0 3.8.2, which provide appropriate actions for the inoperable DG(s). - SURVEILLANCE As noted at the beginning of the SRs, the SRs for each LOP REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.8.1-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains initiation capability (for Functions 1 and 2) and annunciation capability (for function 3). Functions 1 and 2 maintain initiation capability provided that, for 2 of the 3 l (continued) HATCH UNIT 2 B 3.3-204 REVISION

LOP Instrumentation B 3.3.8.1 g] . BASES l SURVEILLANCE emergency buses, the following can be initiated by the j REQUIREMENTS Function: DG start, disconnect from the offsite power ! (continued) source, DG output breaker closure, load shed, and activation ! of the ECCS pump power permissive. l l Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. SR 3.3.8.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation or a failure of annunciation has not occurred. A CHANNEL CHECK is defined for Function 3 to be a comparison of the annunciator status to the bus voltage and an annunciator test confirming the annunciator is capable of lighting and sounding. A CHANNEL . CHECK will detect gross channel failure or an annunciator failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL D CALIBRATION. If a channel is outside the match criteria, it may be an , indication that the instrument has drifted outside its limit. The frequency is based upon operating experience that i demonstrates channel failure is rare. Thus, performance of the CHANNEL CHECK ensures that undetected outright channel or annunciator failure is limited to 12 hours. The CHANNEL CHECK supplements less formal, but more frequent, checks of ! channels during normal operational use of the displays i associated with channels required by the LCO. ! SR 3.3.8.1.2 l A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. (continued) (O .) l HATCH UNIT 2 83.3-204h REVISIONfh l

LOP Instrumentation B 3.3.8.1 BASES SURVEILLANCE SR 3.3.8.1.2 (continued) REQUIREMENTS The Frequency of 31 days is based on operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event. O (continued) HATCH UNIT 2 93.3-204h REVISIONfh

1 l RHR Shutd:wn Cooling Syst_m - Cold Shutdown B 3.4.8 , b) V BASES ACTIONS B.1 and B.2 (continued) During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem or recirculation pump), the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate. SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.

 /7 U

REFERENCES 1. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993. 4 HATCH UNIT 2 B 3.4-43 REVISION A

RCS P/T Limits B 3.4.9 8 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.9 RCS Pressure and Temperature (P/7) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature ch' ..'. These loads are introduced by startup (heatup) and st < (cooldown) operations, power transients, and reat trips. This LC0 limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation. This Specification contains P/T limit curves for heatup, cooldown, and * ' vice leakage and hydrostatic testing, and also limits th v num rate of change of reactor coolant temperature. .iticality curve provides limits for both heatup and criticality. Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within th lowable region. The LC0 establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure. Therefore, the LC0 limits apply mainly to the vessel. 10 CFR 50, Appendix G (Ref. 1), requires the establishment - of P/T limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section III, Appendix G (Ref. 2). The actual shift in the RT., of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be adjusted, (continued) HATCH UNIT 2 B 3.4-44 REVISION [

                    --         . = _ .

RCS P/T Limits j B 3.4.9 BASES l BACKGROUND as necessary, based on the evaluation findings and the ! (continued) recommendations of Reference 5. The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor . vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the , most restrictive regions. The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls. The criticality limits include the Reference 1 requirement that they be at least 40 F above the heatup curve or the cooldown curve and not lower than the minimum permissible O V temperature for the inservice leakage and hydrostatic testing. The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in 1 brittle failure of the RCPB, possibly leading to a ) nonisolable leak or loss of coolant accident. In the event ; these limits are exceeded, an evaluation must be performed i to determine the effect on the structural integrity of the : RCPB components. ASME Code, Section XI, Appendix E l (Ref. 6), provides a recommended methodology for evaluating i an operating event that causes an excursion outside the l limits. l t APPLICABLE The P/T limits are not derived from Design Basis Accident . SAFETY ANALYSES (DBA) analyses. They are prescribed during. normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a i condition that is unanalyzed. Reference 8 approved the curves and limits specified in this section. Since the (continued) l HATCH UNIT 2 B 3.4-45 REVISION h

RCS P/T Limits B 3.4.9 BASES APPLICABLE P/T limits are not derived from any DBA, there are no SAFETY ANALYSES acceptance limits related to the P/T limits. Rather, the (continued) P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition. RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 8). LCO The elements of this LC0 are:

a. RCS pressure and temperature are within the limits specified in Figures 3.4.9-1 and 3.4.9-2, and heatup or cooldown rates are s 100 F during RCS heatup, cooldown, and inservice leak and hydrostatic testing;

b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is s 145 F during recirculation pump startup;

c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vusel is 5 50 F during recirculation pump startup;

d. RCS pressure and temperature are within the criticality limits specified in Figure 3.4.9-3, prior l to achieving criticality; and

e. The reactor vessel flange and the head flange temperatures are 2 90 F when tensioning the reactor l vessel head bolting studs.

These limits define allowable operating regions and permit a large number of operating cycles while also providir.g a wide margin to nonductile failure. The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and inservice (continued) HATCH UNIT 2 B 3.4-46 REVISION

RCS P/T Limits ; B 3.4.9 ; BASES ACTIONS C.1 and C.2 (continued) Operation outside the P/T limits in other than MODES 1, 2, , and 3 (including defueled conditions) must be corrected so ' that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored. Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 212 F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Code, Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline. , Condition C is modified by a Note requiring Required Action f C.2 be completed whenever the Condition is entered. The , O Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and i may have affected the RCPB integrity. SURVEILLANCE SR 3.4.9.1 REQUIREMENTS Verification that operation is within limits is required l every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations. Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied. (continued) j HATCH UNIT 2 B 3.4-49 REVISION %

i

RCS P/T Limits B 3.4.9 BASES h SURVEILLANCE SR 3.4.9.1 (continued) REQUIREMENTS This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and RCS inservice leakage and hydrostatic testing. SR 3.4.9.2 A separate limit is used when the reactor is approaching _ criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical . Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal. SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable limits l ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied. Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start. An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop. (continued) HATCH UNIT 2 B 3.4-50 REVISIONfh 1

RCS P/T Limits ! B 3.4.9 ( BASES SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4 (continued) REQUIREMENTS SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4. In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required. SR 3.4.9.5. SR 3.4.9.6. and SR 3.4.9.7 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LC0 limits. The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS O temperature s 100 F, 30 minute checks of the flange l C/ temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature s 120 F, l monitoring of the flange temperature is required every 12 hours to ensure the temperature is within the limits specified. l The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour Frequency is reasonable based on the rate of temperature - change possible at these temperatures. SR 3.4.9.5 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs. SR 3.4.9.6 is modified by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature 5 100 F in Mode 4. l SR 3.4.9.7 is modified by a Note that requires the Surveillance to be initiated 12 hours after RCS temperature 5 120 F in Mode 4. The Notes contained in these SRs are l q (continued) HATCH UNIT 2 B 3.4-51 REVISION

RCS P/T Limits B 3.4.9 BASES h SURVEILLANCE SR 3.4.9.5. SR 3.4.9.6. and SR 3.4.9.7 (continued) REQUIREMENTS necessary to specify when the reactor vessel flange and head flange temperatures are required to be verified to be within the limits specified. l REFERENCES 1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.

3. ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.

4. 10 CFR 50, Appendix H.

5. Regulatory Guide 1.99, Revision 2, May 1988.

6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E. ,

7. FSAR, Section 15.1.26.

8. Kahtan N. Jabbour (NRC) letter to W. G. Hairston, III (GPC), Amendment 118 to the Operating License, dated January 10, 1992.

9. NRC No. 93-102, " Final Policy Statement on Technical l Specification Improvements," July 23, 1993.

O HATCH UNIT 2 8 3.4-52 REVISION

I SGT System-Operating i B 3.6.4.7 - B 3.6 CONTAINMENT SYSTEMS B 3.6.4.7 Standby Gas Treatment (SGT) System-Operating I BASES BACKGROUND The SGT System is required by 10 CFR 50, Appendix A, GDC 41,

                   " Containment Atmosphere Cleanup" (Ref. 1). The function of         ;

the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary ' containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment. The Unit I and Unit 2 SGT Systems each consists of two fully redundant subsystems, each with its own set of dampers, charcoal filter train, and controls. The Unit 1 SGT subsystems' ductwork is separate from the inlet to the filter train to the discharge of the fan. The rest of the ductwork is common. The Unit 2 SGT subsystems' ductwork is i separate except for the suction from the drywell and torus, ; which is common (However, this suction path is not required for subsystem OPERABILITY). l t Each charcoal filter train consists of (components listed in : order of the direction of the air flow): '

a. A demister or moisture separator; i

b. An electric heater; l

c. A prefilter; j

d. A high efficiency particulate air (HEPA) filter; -

E

e. Two charcoal adsorbers for Unit I subsystems and one charcoal adsorber for Unit 2 subsystems; i

f. A second HEPA filter; and

g. A centrifugal fan.

The sizing of the SGT Systems equipment and components is based on the results of an infiltration analysis, as well as an exfiltration analysis of the Unit I and Unit 2 secondary containments. The internal pressure of the SGT Systems l i (continued) HATCH UNIT 2 B 3.6-111 REVISIONfb l

SGT System-Operating B 3.6.4.7 BASES h , BACKGROUND boundary region is maintained at a negative pressure of (continued) 0.25 inches water gauge when the system is in operation, which represents the internal pressure required to ensure zero exfiltration of air from the building when exposed to a 10 mph wind. The demister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity l of the airstream to < 70% (Refs. 2 and 3). The prefilter removes large particulate matter, while the HEPA filter removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorbers remove gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber. The Unit I and Unit 2 SGT Systems automatically start and operate in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following initiation, all required charcoal filter train fans start. Upon verification that the required subsystems are operating, the redundant required subsystem is normally shut down. g APPLICABLE The design basis for the Unit I and Unit 2 SGT Systems SAFETY ANALYSES during MODES 1, 2, and 3 is to mitigate the consequences of a loss of coolant accident (Refs. 2, 3, and 4). For this event, the SGT Systems are shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment. One SGT subsystem is required to draw-down the Unit 2 secondary - containment and two SGT subsystems are required to draw-down the Unit I secondary containment. The need for Unit I secondary containment during a Unit 2 LOCA arises because of potential leakage past the Unit 2 drywell head onto the refueling floor (i.e., into the Unit I secondary containment). The SGT System satisfies Criterion 3 of the NRC Policy Statement (Ref. 5). (continued) HATCH UNIT 2 B 3.6-112 REVISIONf[

l l SGT Systca-Operating j B 3.6.4.7 l BASES LC0 In addition, with Unit I secondary containment in the (continued) modified configuration, the Unit 1 SGT System valves to the Unit I reactor building zone are not included as part of Unit 1 SGT System OPERABILITY (i.e., the valves may be secured closed and are not required to open on an actuation signal). , APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product , release to primary containment that leaks to Unit 1 and Unit t 2 secondary containments. Therefore, Unit I and Unit 2 SGT Systems OPERABILITY are required during these MODES. SGT Systo requirements for MODES 4 and 5 are covered by LCOs 3.6.4.8 and 3.6.4.9, "SGT System-0PDRVs" and

                           -Refueling," respectively.

ACTIONS The Actions are modified by a Note to indicate that when i both Unit 1 SGT subsystems are placed in an inoperable status for inspection of the Unit I hardened vent rupture O, disk, entry into associated Conditions and Required Actions may be delayed for up to 24 hours, provided both Unit 2 SGT subsystems are OPERABLE. Upon completion of the inspection or expiration of the 24 hour allowance, the Unit 1 SGT subsystems must be returned to OPERABLE status or the applicable Conditions entered and Required Actions taken. i The 24 hour allowance is based upon precluding a dual unit shutdown to perform the inspection, yet minimizing the time both Unit 1 SGT subsystems are inoperable. N ; With one Unit 1 or Unit 2 SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in , 7 days. In this condition, the remaining OPERABLE SGT subsystems are adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single i failure in one of the OPERABLE subsystems could result in j the radioactivity release control function not being - adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the (continued) HATCH UNIT 2 B 3.6-113 REVISION A

4 l SGT System-Operating I B 3.6.4.7 I l BASES h ACTIONS A.] (continued) l l OPERABLE redundant SGT subsystems and the low probability of a DBA occurring during this period. B.1 and B.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.4.7.1 REQUIREMENTS Operating each required Unit I and Unit 2 SGT subsystem for 2 10 continuous hours ensures that they are OPERABLE and W that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for 2 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliabilit.y of fan motors and controls and the redundancy available in the system. SR 3.6.4.7.2 This SR verifies that the required Unit I and Unit 2 SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies (continued) HATCH UNIT 2 B 3.6-114 REVISIONh-

SGT System-0PDRVs B 3.6.4.8 BASES (continued) APPLICABILITY In MODES 4 and 5, the probability and consequences of a LOCA event is reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Unit 2 SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during OPDRVs, since this condition could lead to an inadvertent vessel draindown event. SGT System requirements for MODES 1, 2 and 3, and during other conditions for which significant releases of radioactive material can be postulated, are covered by LCOs 3.6.4.7 and 3.6.4.9, " Standby Gas Treatment (SGT) System-Operating" and

                     -Refueling," respectively.

ACTIONS A_d With one Unit 2 SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this condition, the remaining OPERABLE Unit 2 SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced.because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as , the availability of the OPERABLE redundant Unit 2 SGT subsystem and the low probability of a DBA occurring during this period. B.1 and B.2 - During OPDRVs, when Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE Unit 2 SGT subsystem should immediately be placed in operation. This action ensures that the remaining subsystem is OPERABLE, ' that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected. An alternative to Required Action B.1 is to immediately suspend activities that represent a potential for releasing 4 radioactive material to the Unit 2 secondary containment, l thus placing the plant in a condition that minimizes risk. ) l (continued) l HATCH UNIT 2 B 3.6-117 REVISION A 1

SGT System-0PDRVs B 3.6.4.8 BASES h ACTIONS B.1 and B.2 (continued) Therefore, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. C.1 and C.2 When two Unit 2 SGT subsystems are inoperable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. SURVEILLANCE SR 3.6.4.8.1 REQUIREMENTS Operating each Unit 2 SGT subsystem for 1 10 continuous hours ensures that they are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for 2 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system. SR 3.6.4.8.2 - This SR verifies that the required Unit 2 SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP. (continued) HATCH UNIT 2 B 3.6-118 REVISIONfh

SGT System-Refueling B 3.6.4.9

~T (b  BASES LC0           valves may be secured closed and are not required to open on (continued) an actuation signal).

APPLICABILITY During CORE ALTERATIONS or movement of irradiated fuel assemblies in the Unit I secondary containment, a fuel handling accident could lead to a fission product release to the Unit 1 secondary containment. Therefore, Unit I and Unit 2 SGT System OPERABILITY is required during these conditions. SGT System requirements in MODES 1, 2 and 3, and during other conditions for which significant releases of radioactive material can be postulated, are covered by LCOs 3.6.4.7 and 3.6.4.8, "SGT System-Operating" and "-0PDRVs," respectively. ACTIONS Ad O With one required Unit 1 or Unit 2 SGT subsystem inoperable, 'Q the inoperable subsystem must be restored to OPERABLE status in 7 days. In this condition, the remaining required OPERABLE SGT subsystems are adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in one of the remaining required OPERABLE subsystems could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystem and the - low probability of a DBA occurring during this period. B.l. B.2.1. and B.2.2 During movement of irradiated fuel assemblies, in the Unit I secondary containment or during CORE ALTERATIONS, when Required Action A.1 cannot be completed within the required Completion Time, the two remaining required OPERABLE SGT subsystems should immediately be placed in operation. This action ensures that the remaining subsystems are OPERABLE, that no failures that could prevent automatic actuation have O (continued) /) Q HATCH UNIT 2 B 3.6-121 REVISION A l i

1 I SGT System-Refueling l B 3.6.4.9 j BASES i ACTIONS B.1. B.2.1. and B.2.2 (continued) j An alternative to Required Action B.1 is to immediately suspend activities that represent a potential for releasing radioactive material to the Unit I secondary containment, thus placing the plant in a condition that minimizes risk. If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies must immediately be suspended. Suspension of these activities must not preclude completion of movement of a component to a safe position. The Required Actions of Condition B have been modified by a Note stating that LC0 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3,- the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. C.1 and C.2 When two or three required SGT subsystems are inoperable, if applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in Unit I secondary containment must immediately be suspended. Suspeasion of these activities shall not preclude completion of movement of a component to a safe position. Required Action C.1 has been modified by a Note stating that LC0 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LC0 3.0.3 would not specify - any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.6.4.9.1 REQUIREMENTS Operating each required Unit I and Unit 2 SGT subsystem for 2 10 continuous hours ensures that they are OPERABLE and (continued) HATCH UNIT 2 B 3.6-122 REVISION [{

SGT System-Refueling B 3.6.4.9 BASES SURVEILLANCE SR 3.6.4.9.1 (continued) REQUIREMENTS that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for 2 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA ; filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system. SR 3.6.4.9.2 This SR verifies that the required Unit 1 and Unit 2 SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical ' properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP. + 0 SR 3.6.4.9.3 This SR verifies that each required Unit I and Unit 2 SGT subsystem starts on receipt of an actual or simulated initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in i SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. While this Surveillance can be performed with the reactor at power, operating experience ; has shown that these components usually pass the ; Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was found to be acceptable from a j reliability standpoint. l l 1 REFERENCES 1. Unit 1 FSAR, Section 5.3. ]

2. FSAR, Section 6.2.3.

(continued) HATCH UNIT 2 B 3.6-123 REVISION

SGT System-Refueling B 3.6.4.9 BASES REFERENCES : (continued) 3. FSAR, Section 15.1.41. l

4. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

O O HATCH UNIT 2 B 3.6-124 REVISION A

a ,, A-uA.. a,4a A .4eed - 2A- . - .J-J S J 44-a. k-4 --.---- E ,..a. s a t i t Q UNIT 2 MARKUP OF CURRENT TECHNICAL SPECIFICATIONS AND DISCUSSION OF CHANGES l l 1

i I i i O . I f i f l f i i f i O - I i l

55ctoo 10 DEFINITIONS edd per ( (MINIMUM CRITICAL POWER RATIOlmcPO JThe MINIMUM CRITICAL POWER RATIGlMCPR[shall be the smallest (CPR3wh4ch/ W4 ! exists in the cor gg ) gt s d,'I PERABLE - OPERABILITY A- , g,9 A system, subsystem, tN, component, or device shall be OPERABLE "% l I have OPERABILITY when it is capable of performing its specified tion (s)/46 I.g' kit 'r. thi: & fiatt!:r sh:P 5: th: :::: )r b*3 attendant instrumentatio , controls, normal f4-tt:- power G G Kb, dooling emer th:t all necessary y} Af seal water, lubrication Ogency electricaiother auxiliary ' equipment that are required for the system, subsy, stem, tr4 g, component , k their or device to perform its. function (s) are also capable of performing g related support function (s) eciped 59fd 3 b ODATf6HAL rnNDITION

An OPERATIONAL CONDITION shal ene inciertve--cambination of mode l switch pos e reactor coolant temperature as indTtTted in (Tabi a .z. PHYSICS TESTS g=I g3 0 .6,} ) PHYSICS TESTS shall be those tests performed to measure the fundamental j ' g,3 nuclear characteristics of the eactor core and related instrumentation.7(erM %,. and 6/escribed in Chapter 1 oftheFSAR;4)/uthorizedunderthe provisions of 10 CFR 50.59, or /therwiseapprovedbytheCommission. _ h,FfESSUREJEhuNDARY/ LEAKAGE b.0 p PRESSUP4 COUNOARY LEAKACC shell be 1 e through a no olable fault j ! in a peactor coolant gystem component dy,pipewall)o[rvesselwall. (Ets ) G5 Y $ W hYA(rf Q s b i i O HATCH - UNIT 2 1-4 Amendment No. 14 74 2o

Insert 1.1I for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation (s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. , s Insert 1.1J MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. Insert 1.1E (notusED) L o i 1 L I l Hatch Unit 2 Insert 1.1-4 1o 4 10

DISCUSSION OF CHANGES ! ITS: 1.0 - USE AND APPLICATION l ADMINISTRATIVE , A.19 (continued) In LC0 3.8.1, new times have been provided to perform the determination of ; redundant feature OPERABILITY. These changes are discussed in the Discussion of Changes for LC0 3.8.1. r A.20 Comment number not used. A.21 The definitions of Primary Containment Integrity and Secondary Containment Integrity have been deleted from the proposed Hatch Unit 2 Technical Specifications. This was done because of the confusion associated with these definitions compared to their use in their respective LCOs. The change is editorial in that all the requirements are specifically 1 addressed in the LCOs for the Primary Containment and Secondary ! ( Containment, along with the remainder of the LCOs in the Containment ! Systems section. Therefore, the change is an administrative presentation ; preference adopted by the BWR Standard Technical Specifications, NUREG i 1433. ; A.22 The definition of SHUTDOWN MARGIN has been modified to address stuck control rods. This is consistent with the existing requirement found in i Surveillance 4.1.1.b to account for the worth of a stuck control rod. The relocation of this requirement is considered to be editorial. . A.23 The definition of STAGGERED TEST BASIS has been modified to be consistent with its usage throughout the proposed Hatch Unit 2 Technical Specifications. The intent of the frequency of testing components on a Staggered Test Basis is not changed. The revised definition allows the , minimum Surveillance interval to be specified in the Surveillance l Requirements' Frequency column of the applicable LCOs independent of the ) number of subsystems. This represents a human factored improvement to the current approach, which requires a determination of the Surveillance sub-interval from the test schedule based on the number of subsystems. A.24 The definitions of 0FFSITE DOSE CALCULATION MANUAL and PROCESS CONTROL' i PROGRAM have been moved to the Administrative Controls section. Any technical changes to these definitions are addressed in the Discussion of l Changes associated with Section 5.0. J A.25 These footnotes are addressed by the exceptions allowed to LC0 requirements in the proposed Special Operations section (currently titled ( "Special Test Exceptions"). Any technical changes to this requirement will be addressed with the content of the proposed chapter location. Refer to proposed LCO 3.10.1, LC0 3.10.2, LC0 3.10.3, and LC0 3.10.4. l HATCH UNIT 2 4 REVISION l

REACTOR COOLANT SYSTEM Ikb 3.4.9 O 3/4.4.6 PRESSURE / TEMPERATURE LIMITS y f REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ! U*3.d.i 3.4.6.1 The reactor coolant system temperature and reactor vessel pres-sure shall be limited in accordance with the limit lines shown on (1) : 93.n 2 ' Figure 3.4.6.1-1 for heetup by non-nuclear means, cooldown following a ' M 5*M noclear shutdown and low power PHYSICS TESTS; (2) Figure 3.4.6.1-2 for

        " ' #^ *'d operations with a critical core other than low power PHYSICS TESTS; and A.'t 9(3) Figure 3.4.6.1-3 for inservice hydrostatic or leak testing, with:                                                  g!

A l s e s.u.t.Lb a. A maximum heatup of 100*F in any one hour period, and f

b. A maximum cooldown of 100*F in any one hour period.

APPLICABILITY: At all times. ACTION: 6MPoM ,1, Y l pa*3 ith any of the above limits exceeded, restore the temperature and/or W\ pressure to within the limits &itnin .50 minutes; perform an engineering y 0 630 y evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the reactor coolant system; determine @* Wh* P"M , that the reactor coolant system remains acceptable for continued opera- Wk c. M tions or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN Pr G M (withinthenext24 hours. 6 ( hk , i SURVEILLANCE REQUIREMENTS g.9 A 4.4.6.1.1 The reactor coolant system temperature and reactor vessel . pressure shall be determined to be within the limits at least once per 30 minutes 5during system heatup, cooldown and inservice leak and hydrostatic U Q esting operations. Mg3M .4. 6.1.2 The pressure reactor shall coolant system be determined to betemperature to the rightand reactor of the vessellimit criticality line of Figure 3.4.6.1-2 within 15 minutes prior to the withdrawal of h. control rods to bring the reactor to criticality. E.4 6.1.3 The react material irradiat n surveillance sp cimens shall be r oved and examine to determine chan s in material pro rties, as requir by 10 CFR 50, A endix H. The res ts of these exam ations ! shall be ed to update Fi res 3.4.6.1-1, 3. 6.1-2 and 3.4.6. 3. ' m . t. m O A . 3, p yopee SE ~3.4.9# ut. > .9. u j Q M.J.4 9. y HATCH - UNIT.2 3/4 4-13 Ame,ndment No. 118 MC

                                                                                                                            .                     L

  ..-   -          . - - - .                 . . . . . = _ _ - .           -                  .               -               - =_.
                                - SitT PME                                                                             6 P'd li'a h*' 349 O                           1600 9

4 B' B 1400

                          -                                                         lll 2                                                         ej                                                 \

E.

1200 -

l 8

                           =                                                      .
                                                                                   .V                                                     \
                                                                                  *,                                                      1 g                                                     ,' l                                                     !

e 1000 Y s sw . - conc .EtTtut 7 r 800 l AFTER A55Uh4ED 32 EFPY 5Hrr N AN N g =ctO rr.or -so r l 9 \ O g 600 I 8 - Sh NUCLEA yMcATUP/ e

                           ,                                                   /

a ) - VE5SEL OtSCONTINUffY w f L864rr5 g 400 7

                                                                                                -- co=c ctTt Nc wits m                                             j y                312 Psc                                                   32 EFPY SHIFT 200 CURVE B' 15 NOT LIMmNG soLTUP                  [                           FOR INFORMATION ONLY sov                   #

CURVE B 5 VAUD FOR 2 m W mm D . i s 0 100 200 300 400 500 600 MNIMUM REACTOR VE35EL METAL TEMPERATURE fr) Kup E 7.4 1 PERATURE PRESSURE LIMITS FOR NON NUCLEAR HEATUP, LOW POWER PRYSICS TESTS ANO

,                                                                     COOLDOWN FOLLOWING A SHUTOOWN O                                                                              -

2 oW HATCH - UNIT 2 3/4 4-14 Amendment No. 118

                     - $$sh pge -                                                                                SP eciGlcahm S.4.f O                      1600 C' C 1400 a                                                      .

1200 .

8 2 e ll e 1000 [ ci in la l\ ' c' - CORE RELTUNE I h AFTER ASSUMED 32 EM i g 500 SHIFT FROM AN WITEL 1 o WELD RTegnOF -50*F b - O C - NUCLEAR (CORE CRmCAL) z 600

uurr l

t- l 2

                                                                                     - VE55EL DiscoNnNUITY W
                        @D f                      ""

f -- CORE RELTLwE wrrH

                   $                  sir esc                 j                              32 Erry surT E                                    /

2oe

                                         .,        j                                 o r-,oN .Lv
                                                 /                                   CURVE C 5 VAUD FOR D                    f                                     32 EFPY OF OPERATION D            100           200                  300          400        500         600 44NMUM REACTOR VE5SEL METAL TEMPERATURE PF) 4                     p,auas 5 4.9- 3
                                      'T'EMPERATURE#RESSURE LIMITS FOR CRITICALITY
                                                 "'?? ' "a" * ^ ^^'*:! ' '_ '"*.* " ^ ".0 ;N A5
                                                        .""^** Y 40CNISO-1tPP4H
                                                                        "~'~~

O 3d5 HATCH - UNIT 2 3/4 4-15 Amendment No. 118 j

p_suye s@ n~+~ m . O , t 1600 l l A' A , 1400 . r e

      $g,                                   e s                                                                 ,

1200- l 8

2 e 1000 ) ci m ! j M N f f A' - CORE BELTUNE , AFTER ASSUMED 32 EFPY ' g 800 SHFT FROM AN MfTML o WELD RT,,nor -Soar b O i 5 E g 600 [ A - SY5 TEM NYDROTEST UMfT wTH FUEL w vEs5EL ; I:

2, -

                                                            - VE5SEL DISCONTWUfTY 400                                                 "
                                                            = = CORE BELTLME wfTH                            l Lo                                                           32 EFPY 5HFT 312 Psc  .-

u Q. I 200 ' SDLTuP CURVE A' 15 NOT LMfilNG so*r FOR INFORMATION ONLY l CURVE A 15 VAUD FOR ! 32 EFPY OF OPERATION l 0 s i 0 100 200 300 400 500 600 MN1WUM REACTOR VE5SEL METAL TEMPERATURE fr) scua en.9hTEMPERATURE #RESSURE LIMITS FOR ! INSERVICE HYDROSTATIC 7007 on d Lsovice. i N8tNW4'4694 4 4ko,c Tc.sfs .6 ; 4 HATCH - UNIT 2 3/4 4-16 Amendment No.118

ffec&b>3.99_ REACTOR COOLANT SYSTEM k IDLE RECIRCULATION LOOP STARTUP LIMITING CONDITION FOR OPERATION j .4.i 3.4.1.3 An idle recirculation loop shall not be started unless the temperature differential between the reactor coolant within the dome and he bottom head drain is 1 I45 F, and ! b.

a. The temperature differential between the reactor coolant g,tk within the idle loop to be ' started up and the coolant in the reactor pressure vessel is s 50 F when both loops have been idle, or [

b. The temperature differential between the reactor coolant u, ,

within the idle and operating recirculation loops is s 50'F _ when only ont loop has been idle at operati loop i trate q%SCAof Patea toqp t 1o% APPLICABILITY: CONDITIONS 1, 2, 3 and 4. goy b 3Rs3.413d M T 4 ACTION: N

       )     Yith temperature differences and/or flow rate exceeding the above limits, '

suspend startup of any idle recirculation loop, a f6fese k h.h WIDO l A,6, %h6 1 SURVEILLANCE REQUIREMENTS l g,4.1.3.3 50 4A4 4.4.1.3 The temperature differential and flow rate shall be determined to be within the limit within minutes prior to startup of an idle , recirculation loop.

                                               \     n.q l

l a HATCH - UNIT 2 3/4 4-3 suf

m DISCUSSION OF CHANGES ITS: SECTION 3.4.9 - RCS PRESSURE AND TEMPERATURE (P/T) LIMITS ADMINISTRATIVE (continued) , A.4 These requirements are presented as Surveillances in the P/T limits Specification. The requirements remain unchanged. As such, this change is administrative. A.5 Title changes to the P/T curves have been made for consistency with the i ITS SRs. TECHNICAL CHANGE - MORE RESTRICTIVE 1 M.] A specific Completion Time for the engineering evaluation and determination is proposed. The proposed time of 72 hours is considered reasonable for operation in MODES I, 2, and 3 because the limits represent , controls on long term vessel fatigue and usage factors. In MODES 4 and 5, the proposed time (prior to entering MODE 2 or 3) would prevent entry in the operating MODES which is consistent with the current LCO 3.0.4. M.2 Three Surveillance Requirements have been added. SR 3.4.9.5 ensures the , vessel head is not tensioned at too low a temperature. SRs 3.4.9.6 and ' 3.4.9.7 ensure the vessel and head flange temperatures do not decrease

 ,         below the minimum allowed temperature every 30 minutes, or every 12 hours,  !

depending upon the RCS temperature. These are additional restrictions on plant operation. M.3 The ACTIONS required to be taken when a recirculation pump is started without having met the temperature requirements have been changed. Currently, the ACTION only states to suspend the startup of a recirculation loop. This however, does not provide an action if the loop is already operating. Proposed ACTIONS A, B, and C now require an engineering evaluation to be performed to ensure continued operation is acceptable. This is an additional restriction on plant operation. M.4 The Surveillance Frequency has been changed to require the temperature checks to be performed within 15 minutes prior to startup of the idle recirculation pump, instead of the current 30 minutes. This is an additional restriction on plant operation. O . HATCH UNIT 2 2 REVISION

i i i

  .e                                  DISCUSSION OF CHANGES                                                        !

g4 ITS: SECTION 3.4.9 - RCS PRESSURE AND TEMPERAT,URE (P/T) LIMITS , I i IECHNICAL CHANGE - LESS RESTRICTIVE l ,. "Goneric" LA.1 The details relating to system design and operational limits have been ! relocated to plant controlled documents (e.g., updated FSAR and l procedures). The single operating loop limit on flow rate is considered ' an operational limit since it is not directly related to the ability of ! the system to perform its safety analysis function. The flow rate is ! limited only to minimize reactor vessel internals vibration. Changes to I plant controlled documents will be in accordance with the provisions of 10 CFR 50.59. I i i 1

O ,

I l 1 O HATCH UNIT 2 3 REVISIONf{

I j CONTAINMENT SYSTEMS l 3/4.6.6 CONTAINMENT ATMOSPHERE CONTROL STANDBY GAS TREATMENT SYSTEM 4' 4' 7 j LIMITING CONDITION FOR OPERATION O, ('/ ll o 3.6.6.1 Two Hatch-Unit 2 independent standby gas treatment subsystems ! 364'l and two Hatch-Unit 1 independent standby gas treatment subsystems shall ! be OPERABLE. , APPLICABILITY: CONDITIONS 1, 2, 3,(1r F Q - ACTION. r A c.7, l a. WithoneoftheabovereutredstandbfsystemtoOPERABLEgas treatment tatus subs stems pb6 d inoperable, restore the noperable su ithin 7 days or be in at least HOT SHUTDOWN within the next 12 hours pg6 kand in COLD SHUTDOWN within the following 24 hours,

b. With two or more of the above required standby gas treatment subsystems inoperable be in at least HOT SHUTDOWN within 12 hours and in COLD LA O 3 o.3 SHUTDOWN within the next 24 hours,[engt as sii - 6 , M un c.;

c. With both of the Hatch-Uni 1indekendentstandbtallaionoftheUnftItorushar gas treatm t 3 ned vent, subsystems inoperable for i Unit 2 operation may continu for a cumulative total of up to 7 ays provided all of the following equirements are met:

1. Prior to removing either U t I standby gas treatment subsyste from service demonstrate that a egative pressure can be maintained n theUnil2secondarycontain nt and the Unit I modified seconda containment under the followin conditions:

The Unit I secondary contai ent is in the modified mode.
Both Unit 2 standby gas treat .nt subsystems are aligned with suction from both of the subje areas and are operating with each filter train flow rate not are than 4000 cfm.
Calm wind conditions (< 5 mph) ex t.

A 2. Main in both Unit 2 standby gas treatme t subsystems OPERABLE.

3. Mainta n Unit 2 secondary containment inte rity, except for Unit 1

[q) standb gas treatment system OPERABILITY re utrements. V 4. Maintain nit I modified secondary containmen integrity, except for Unit I sta dby gas treatment system OPERABILIT requirements.

5. Allow no Uni ! CORE ALTERATIONS.

6. Allow no handl of irradiated fuel or spent fuel hipping casks in the modified Un I secondary containment, if both Unit I standby gas eatment subsystems are not restore to OPERABLE status wittin the allowable c ulative time period of 7 days, or f any of the above requirements cannot be , be in at least HOT SHUTDOWN wit 'n the next j

                  ,                g,in rni n WUTDOWN w ~ in the following 24 hours.                           _

SURVEILLANCE REOUIREMENTS l 4.6.6.1.1 Each Hatch-Unit 2 standby gas treatment subsystem shall be I demonstrated OPERAF.LE: l e control room. tinw throuah h a N P a. blirI a fro atinbar.co urbers anmeHf ving that the system hj lg . g.7. ) gerates for aton e heaters rrast am m a total of 10 houTb each 31 au c controi -

                                                                                      - -days
                                                                                          - - - with              2                 l
                        'L.      At least once per 18 months or (1) after any structural main-                                      I tenance on the HEPA filter or charcoal adsorber housings, or              I                        ;

S U .6 M 7.2. following painting fire or chemical release in any venti- i f2)ionzonecommunicatlngwiththesystemby: lat l Q .'l 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test proceoures 4%>00d b of Regulatory Positions C.S.a C.5.c and C.5.d of Regula-tory Liuide 1.52 Revision I duly 1976,andthesystem

     $fdu48g                           flow rate is 40 0 + 0, -100 cfm.

bbl 2. Verifyigwithin31dafsafterremovalthatalaboratory analyst $ofarepresenativecarbonsampleobtainedin I g f a accordance with Regulatory Position C.6.b of Regulatory ( ) Guide 1.52 Revision 1. July 1976, meets the laboratory

ib 1 v I uy b b isrW'M ,

                      ,                                                                                          c ko-ama O%

enperforminginservicehdrostaticorleaktestingwiththereactp 3.go.3, Ls rva e c olant temperature above 2 2*F. - g gn HATCH - UNIT 2 3/4 6-40 Amendment No. 94, 12 TctQ 8pv fw, y 3,

i 4 -fdr ') no )

DISCUSSION OF CHANGES ITS: SECTION 3.6.4.7 - STANDBY GAS TREATMENT SYSTEM - OPERATING ADMINISTRATIVE l l A.1 This allowance the hardened is beingeen vent has installed. deleted since it is a one-time allowance only and A.2 The technical content of this requirement is being moved to Section 5 of ! the proposed Technical Specifications in accordance with the format of the l BWR Standard Technical Specifications, NUREG 1433. Any technical changes to this requirement are addressed in the Discussion of Changes associated i with proposed Specification 5.5.7. A surveillance requirement is added i (proposed SR 3.6.4.7.2) to clarify that the tests of the Ventilation j Filter Testing Program must also be completed and passed for determining , OPERABILITY of the SGT System. Since this is a presentation preference ; that maintains current requirements, this change is considered i administrative. { A.3 The technical content of this requirement is being divided into two i Surveillances. The majority of the Surveillance will be performed in LCO ! 3.3.6.2 requirements. The actual system functional test portion will be performed as SR 3.6.4.7.3. This ensures the entire system is tested with l proper overlap. j i TECHNICAL CHANGE - MORE RESTRICTIVE i M.1 The Unit 1 SGT System Surveillances have been specifically written into this LCO instead of providing a cross reference. The current Hatch Unit . I surveillances are written as pro)osed SRs 3.6.4.7.1 and SR 3.6.4.7.2. i Also, proposed SR 3.6.4.7.3 now app' ies to the Unit 1 SGT System. This is not currently required by the Unit 1 Technical Specifications. These changes and additions, therefore, are considered an additional restriction on plant operation. ; i M.2 SR 3.6.4.7.1 requires the SGT System to be run 10 continuous hours each 31 days, while the CTS state a total of 10 hours. This is an additional l restriction on plant operations. TECHNICAL CHANGE - LESS RESTRICTIVE ,

   " Generic"                                                                                         f LA.1 Details of the methods for performing this surveillance are relocated to                      3 the Bases and procedures. Changes to the Bases will be controlled by the                     '

provisions of the proposed Bases Control Process described in Chapter 5 of ; the Technical Specifications. Changes to the procedures will be i controlled by the provisions of 10 CFR 50.59. 9 O l sma Um 2 1 ; RmS109s

DISCUSSION OF CHANGES o) ( ITS: SECTION 3.6.4.7 - STANDBY GAS TREATMENT SYSTEM - OPERATING TECHNICAL CHANGE - LESS RESTRICTIVE

  " Specific" L.1   The phrase " actual or," in reference to the automatic initiation signal, has been added to the surveillance requirement for verifying that each subsystem actuates on an automatic initiation signal.          This allows satisfactory automatic system initiations for other than surveillance purposes to be used to fulfill the surveillance requirements. Operability is adaquately demonstrated in either case since the subsystem itself cannot discriminate between " actual" or " simulated" signals.

L.2 Comment number not used. O L.3 An ACTION Note is proposed to allow inspection of the Unit I hardened vent rupture disk while Unit 2 is operating. This inspection will cause both the Unit 1 SGT subsystems to be inoperable and, thus the allowance to delay entry into associated Conditions and Required Actions is needed, provided both the Unit 2 SGT subsystems are operable. The 24 hour allowance allows Unit 2 to continue operation during the inspection and minimizes the time when the Unit 1 SGT subsystems are inoperable. /N U HATCH UNIT 2 2 REVISIONf6

I I J CONTAINMENT SYSTEMS

                                                                                                                 #cI khpe/ f.f, 7 3/4.6.6 CONTAINMENT ATHD5pHERE CONTROL STANDBY CAS TREATMENT SYSTEM V                            LIMITINE CONDITION FOR OPERATION
                           ~3.6.6.1 Two Hatch-Unit 2 tnoepencent standby gas treatment subsystems See                 and two Hatch-Unit 1 independent standby gas treatment subsystems shall be OPERABLE.

bwwed of APPLICARILITY: CONDITIONS 1, 2, 3, and *. Qedr y 1 In: 3 6.511 gilgg:

      %1 Yh"                a. With one of the above required standby gas treatment subsystems inoperable, restore the inoperable subsystem to OPERABLE status b'yi)                     within s or be in at least HOT SHUTDOWN within the next 12 hours 7 day $HUTDOWN within the following 24 hours.

i and in COLD j L t, , I b. With two or more of the above requtred standby yas treatment subsystems I inoperable be in at least HOT SHUTDOWN within L2 hours and in COLD SHUTDOWNwIthinthenext24 hours,exceptasal'owedbyActionc.

c. With both of the Hatch-Unit 1 independent standby gas treatment subsystems inoperable for installation of the Unit I torus hardened went, Unit 2 operation may continue for a cumulative total of up to 7 days proviced all of the following requirements are met: l

1. Prior to removing either Unit I standby gas treatment subsystem from service demonstrate that a negative pressure can be maintained in theUnil2secondarycontainmentandtheUnitImodiftedsecondary containment under the following conditions: .

The Unit I secondary containment is in the modified mode.
Both Unit 2 standby gas treatment subsystems are alioned with suction from both of the sub, ject areas and are operating with each filter train flow rate not more than 4000 cfm.
Calm wind conditions (< 5 mph) exist.

2. Maintain both Unit 2 standby gas treatment subsystems OPERABLE. !

s- / 3. Maintain Unit 2 secondary containment integrity, except for Unit I \ standby gas treatment system OPERABILITY requirements.

4. Maintain Unit 1 modified secondary containment integrity, except for Unit I standby gas treatment system CPERABILITY requirements.

5. Allow no Unit 1 CORE ALTERATIONS.

6. Allow no handling of irradiated fuel or spent fuel shipping casks in the modified Unit I secondary containment, if both Unit I standby gas treatment subsystems are not restored to OPERABLE '

status within the allowable cumulative time period of 7 days, cr if any of the above recuirements cannot be met, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. SURVEILLANCE RE00fREMENTS 4.6.6.1.1 Each Hatch-Unit 2 standby gas treatment subsystem shall be demonstrated OPERABLE:

a. from the control room, flow through the HEPA By initiating filters and c6arcoal adsorbers and verifying that the system operates for at least a total of 10 hours each 31 days with

                                         'ha heaters on automatic raneral

b. At least once per IB months or (1) after any structural main-D *b,l tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venti- "

lation Zone coinnunicating with the system by: l

1. Verifying that the cleanup system satisfies the in-place S $."T 't testing acceptante criteria and uses the test crecedures A.1 brbb# N ek'N flourrate is 4000 + 0, -1000 cfm ie7baI e y en

                                                                                                              - LA.

(.$,'7. c. 2. Verifying Eitetn 37 davs af ter removal that a laboratory analysts of a representattve carbon sample obtained in e "IcA;g accordance with17egulatory Position t.e.o or peculatnry uide I N. RMisinn 1 Alv 19 0 mee n m .aporatlFf g,3 [ A d4g f4.h g testing criteria of Regulatory Position C.6.4 of Regula-( gg g tory Gutde 1.52 Revtsien 1, Duly 1976.

                            *dhen performing inservice h drostatic or leak testing with the reactor colanttemperatureabove2f2'F.

HATCH - UNIT 2 3/4 6-40 Amenoment No. G4, 124 id b

[pelNesbu $*SE CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) y,5, y,J 3. Verifying a system flow rate of 4000 +0, -1000 cfm during system operation when tested in accordance wi

-                                                         N510-19/d.

c. After everv_720 hours of charcoal adsorber operation by vertrying[within 3J d hs]after removal that a laboratory OL.A.l analysis or a representative carbon samnle obtained in accord _-

f 7.7, C antewith]RegulatoryPositionC.6.bofRegulatoryGuide1.52,] evision 1, July 1976f meets Ine iaooratory testing u i m .. A, 3 p or Regulatory Position C.6.a of Regulatory Guide 1.52, Revision Ql, July 1976. g

d. At least once per 18 months by: g

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water

[. f. 7. ) Gauge while operatin' g the filter train at a flow rate of 4000 +0, -1000 cfm.

2. Verifying that the filter train starts and isolation fee pw dampers open on each of the following test signals:

              ,"                                           a.                         Drywell pressure-high, 3M                                                     b.                        High radiation on the; W 5eska
   .J b                                                                               1)   Refueling floor,

2) Reactor building.

c. Reactor Vessel Water Level-Low Low (Level 2).

3. Verifying that the heater $ dissipate 18.T+~l~.5 An A when testatt )A accordance Ath(ANSI N510-194 g qpp , , , - c.r.9 O-HATCH - UNIT 2 3/4 6 41 Amendment No. D , 109 M G.

fyge,'[o'esTio n fr Soi CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

e. After each complete or partial replacement of a HEPA filter - I f,,f,7,4 bank by verifying that the HEPA filter banks remove 2 99% of the nnp when they are tested in place in accordance with S 10-19757while operating the system at a flow rate of 40

                   +0, -1000 cfm.

f. After each complete or partial replacement of a charcoal f, f,7, 4 adsorber bank by verifying that the charcoal adsorbers remove 5 2 99% of a halogenated hydrocarbon regrigerant test gas when they are tested in place in accordance with QNSI Nb10-1975 A,q while operating the system at a flow rate of 4000 + 0, -1000 cfm. '

          . 6.1.2 Each Hatch-Unit 1 standby gas treatment subsystem shall be                        ;

emonstrated OPERABLE per Hatch-Unit 1 Technical Specifications. h

Disc a s w VT C L 9e5 Y ETS
3 L.4 T IN Sec k y f,g, f

k i O - HATCH - UNIT 2 3/4 6-42 3c4'(,

                                                                    &ge.$lcci!Ok       $* S*

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) y, y, 7, g 2. Verifying that the cleanup systein satisfies the in- ' place testing acceptanca criteria and uses the test procedures of> Regulatory Positions C.5.a. C.5.c an , A,2 L.b.d of Regulatory Guide 1.52. Revision 1, July 1976, ano tne system. flow rate is 2500 cfm + 10 percent. Verifying [wibin 3k after removal that a laboratory analysis of a re>resentative carbon sample obtained in accordance with11egulatory Position C.6.b of Hegulatory) ! hs3 Guide 1.52, Revision 1, July 1976fmeets the laboratory fg

                            / testing criteria of Hegulatory Position C 6.a of Regula- v[,4 l

( tory Guide 1.52, Revision 1, July 1976. _ Ml f" f,7, J

4. Verifying a system flow rate of 2500 cfm i 10 percent -

during system 03eration when tested in accordance with AN5I N510-1975)

d. After every 720 hours of harcoal adsorber operation by verifying twi%1n 3r daKD after removal that a laboratory O analysis of a re2resentative carbon sample obtained in accordance withflegulatory Position C.6.b of Reculatory Guide) 1.52, Revision 1. Julv 1976I meets the . laboratory testing I[

gcriteria of Regulatory Position C.6.a of Regulatory Guide L, , 4 [ 1.52, Revision 1, July 1976.

e. At least once per 18 months by: '

        )* J

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 in.

W. G. while operating the system at a flow rate of 2500 cfm 2 10 percent.

2. (Deleted)

J i O i l i HATCH - UNIT 2 3/4 7-7 Amendment No. 96 i

t 1 l I PLANT SYSTEMS . b('eei$ca haa 5.C. 7 __ SURVEILLANCE RE0VIREMENTS (Continued) k3. Verifying that on each of the below pressurization mode actuation test signals, the system automatically switches to the pressurization mode of operation and maintains the main control room at a positive pressure of = 0.1-in. W.G. relative to the adjacent turbine building during 9 paggoa ! system operation at a flow rate s 400 cfm. g g3 a) Reactor vessel water level - low low low [4 [a b) Drywell pressure - high 1 7, .,f c) Refueling floor area radiation - high d) (Deleted) l e) Main steam line flow - high t f) Control room intake monitors radiation - high

f. After each complete or partial replacement of a HEPA filter 56 l^ bank by verifying that the HEPA filter banks remove 2 99 percent of P when they are tested in-place in accordance wit A g,q 0-19 while operating the system at a flow rate of 2500 cfm 1 10 percent.

g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove f.6% a 99 percent of a halogenated hydrocarbon refrigerant test cas.

when they are tested in-place in accordance with LNSI N510 107 44 while operating the system at a flow rate of 2500 cfm 10 percent. T i A,5 SR 3 ** d '" ** *W O HATCH - UNIT 2 3/4 7-8 Amendment No. 4 , 96, 127 g

I I DISCUSSION OF CHANGES ( ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP) i l ADMINISTRATIVE A.1 Note 2 is added to the proposed Technical Specifications to provide an allowance, in the future, to use refrigerants equivalent to those specified in ASME N510-1989 for testing purposes. The use of R-11 as a test gas is expected to be changed due to environmental considerations. This change maintains equivalent test methods to those currently specified : in the standards and is, therefore, considered an administrative change. A.2 Current Technical Specifications for in-place testing of the SGT and MCREC Systems reference Regulatory Position C.5.a, C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 1, July 1976. Proposed Technical Specification 5.5.7.a references Regulatory Guide 1.52, Revision 2, Section 5c and ASME N510-1989, Section 10. The change to the current reference is an update to the later revision of Regulatory Guide 1.52 but , does not change the current testing requirements. Therefore, this change is considered administrative. A.3 Current Technical Specifications for the SGT and MCREC Systems reference Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, for obtaining a representative sample of charcoal for testing purposes. This reference is proposed to be changed to Regulatory Guide 1.52, O m Revision 2, Section 6b and ASME N510-1989, Section 15, and Appendix B. The proposed change updates the present reference without changing current testing requirements. Since present Technical Specification testing i methods are retained, this change is considered administrative. A.4 The current Technical Specification reference to ANSI N510-1975 is proposed to be changed to ASME N510-1989 for the SGT and MCREC Systems. The proposed change in testing standards will provide an update to the present standard without changing current testing requirements. Therefore, the proposed change is considered administrative. A.5 A statement of applicabilty of SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for surveillance frequency extensions do apply, since these SRs are not normally applied to frequencies identified in the Administrative Controls section of the Technical Specifications. Since this change is a clarification needed to maintain provisions that would be allowed in the LC0 sections of the Technical Specifications, it is considered administrative. A.6 CTS state testing criterion as C.6.a of Regulatory Guide 1.52, Rev.1. ' This is replaced with explicit acceptance criterion of 0.2% penetration, which is consistent with the value specified in Regulatory Guide 1.52, , Rev. 2, March 1978 (when rounded) and with the penetration value calculated using the formula stated in BWR Standard Technical . Specifications, NUREG 1433. This change is considered administrative. See Discussion of Change 5.5.7, Comment M.1, for details of laboratory - Os testing method change. , HATCH UNIT 2 1 REVISIONh 1

DISCUSSION OF CHANGES O ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP) TECHNICAL CHANGES - MORE RESTRICTIVE M.1 Current Technical Specifications for the SGT and MCREC Systems reference Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976, ' for the laboratory testing of the charcoal samples. The current laboratory test standard used for the charcoal is RDT-M16-IT. Proposed ITS 5.7.12.c requires laboratory testing in accordance with ASTM D3803-1989 at a temperature s 30cc and 2 95% relative humidity. The ASTM D3803- r 1989 testing standard is more conservative than the current RDT-M16-IT , standard and is endorsed by the NRC for use throughout the' industry. 1 M.2 Comment number not used. i O i i l i TECHNICAL CHANGES - LESS RESTRICTIVE

    " Generic" i

LA.1 Details of the methods for implementing this specification are relocated ! to the FSAR and procedures. Additionally, changes to the procedures and O, the FSAR are controlled in accordance with 10 CFR 50.59. ; i HATCH UNIT 2 2 REVISION ((

DISCUSSION OF CHANGES O ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP) TECHNICAL CHANGES - LESS RESTRICTIVE LA.2 The visual inspection of the MCREC System and all components before each leak test is not included in the proposed TS. This type of general maintenance inspection is included in procedures and not usually made a part of Technical Specification requirements. The placement of this type of requirement in plant procedures is considered a generic less restrictive change. i

    " Specific"                                                                        ,

L.1 Comment number not used. i L.2 The current Technical Specifications require testing of the SGT System 1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or 2) following painting, a fire or chemical release in any , ventilation zone communicating with the system. Plant Hatch has performed tests and evaluations and has determined that the use of water based paints and the performance of metal grinding, buffing, or welding are not A detrimental to the charcoal filters of the SGT System, either prior to or l (/ during operation. These activities should not require surveillance of the l SGT System upon their conclusion. This applies to all types of welding conducted at Plant Hatch and tracking of the quantity of weld material ' used is not necessary. ' L.3 Comment number not used. L.4 CTS require that charcoal carben samples meet the laboratory testing criterion of Regulatory Guide 1.52, Revision 1, Position C.6.a. This position in turn references you 'co Table 2 of the Regulatory Guide. ITS proposes an explicit acceptance criterion of 2.0% when tested at 95% RH. j Background The Hatch Main Control Room Environmental Control (MCREC) system contains two filtration units, each complete with upstream and downstream HEPA filters, a 2-in. bed charcoal adsorber section, and a fan. Note that the system does not contain heaters which would be equivalent to that described in RG 1.52. The MCREC system design considers the operation of I O only one filtration unit at a time. The air entering each filtration unit consists of 400 cfm (maximum) of outside air and a portion of main control room recirculated air (2I00 cfm) for a total of approximately 2500 cfm. HATCH UNIT 2 3 REVISION r

i p DISCUSSION OF CHANGES Q ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP) TECHNICAL CHANGES - LESS RESTRICTIVE The leaving air from each filtration unit mixes with the remainder of the ! recirculated air from the main control room, and then flows through air handling units with direct expansion cooling coils. The air handling units control the temperature and dehumidify the air in the main control ; room. Thus, heaters are not installed in the filtration units. However, ! as discussed in the Analysis section below, the MCREC system will maintain relative humidity of the air entering the filtration units less than 70% RH. The main control room dose calculations for plant accident conditions credit the charcoal in the filtration unit adsorber bed section with an - overall average efficiency of 95% for all forms of iodine. To provide I assurance of the quality of charcoal to meet its intended design function, the charcoal is periodically laboratory tested. Since the licensing of the plant, the charcoal in the MCREC system has been tested using the RDT-M16-1T standard. The Technical Specification Improvement Program (TSIP) will introduce the NRC recommended ASTM D3803-1989 standard. It is therefore necessary to establish the laboratory testing and acceptance criteria. Proposed Charcoal Testino Criterion: Testing Method: ASTM D3803-1989 Testing Parameters: 30 C 0 95% RH Acceptance Criterion: Maximum Methyl Iodide Penetration of 2.0% Analysis: A calculation has been performed which documents the relative humidity in the main control room during normal and pressurization modes of operation. The calculation documents the condition of the air entering the filtration unit during pressurization mode of operation as being less than 70% RH, assuming that the recirculated air from the control room mixes with outside air at 100% RH. The calculation is based on the design heat load of the main control room. The MCREC system will maintain relatively constant relative humidity in the control room for varying heat load conditions because the temperature of the air leaving the air handling units is held constant by the direct expansion cooling coils. Based on this discussion, it should be acceptable from a design and safety perspective to test the charcoal using the ASTM D3803-1989 standard for the 30 C and 70% RH condition. However, since the system does not have heaters which maintain the 70% RH, the testing relative humidity recommended by the NRC is 95%. To maintain a filter efficiency equiv'alent to the current credited FSAR value of 95% while testing the charcoal using the ASTM D3803-1989 test HATCH UNIT 2

                                            %                            REVISION

DISCUSSION OF CHANGES l (Jm) ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP) TECHNICAL CHANGES - LESS RESTRICTIVE criterion, it is necessary to establish a safety factor. The proposed safety factor is 2.5. Using the methodology of RG Guide 1.52: Removal Efficiency = 100% - 100%-efficiency credited in safety analysis = 100 - (100 - 95) (for test) Safety Factor 2.5 will yield a required test efficiency of 98%, which corresponds to a penetration of 2.0%. This required test removal efficiency is based on assumed removal efficiencies of elemental and organic iodide being the same at 95%. However, it is generally recognized that removal efficiency of elemental iodine is considerably higher than that of organic (methyl) iodide. The expected iodine species primarily exists in the form of elemental and organic iodide. Laboratory testing of carbon is performed by challenging carbon samples with methyl iodide which is an organic form of iodine. However, if the testing of the carbon samples were performed using elemental iodine, it is presumed that the efficiency established would have been much higher, because of the higher removal efficiency of elemental iodine. To demonstrate this effect, assume an iodine species partition as described in Regulatory Guide 1.3 (91% elemental, 5% particulate, 4% organic). Then assume relative efficiencies for elemental / particulate ' versus organic, as described in Regulatory Guide 1.52, for uncontrolled humidity, instead of assuming equal removal efficiencies of 95%, to demonstrate expected differences in removal efficiency (30% organic versus 90% elemental yields (1-0.30/1-0.90-7)). Then the following estimates are , made: 0.96E, + 0.04E, - 0.95 (FSAR credited value) (1-E,)7 E, (elemental vs organic penetration) Where: E, - elemental + oarticulate efficiencies E, - organic efficiency i Solving simultaneously yields E, - 0.96 and E, - 0.72 l i It is proposed that the laboratory tested methyl iodide penetration be 2.0% maximum. Then Allowable Penetration - (1-0.72)/(Safety Factor) .02 (or 2.0% maximum) Safety factor established by this methodology: 14 l HATCH UNIT 2 h REVISION h

t DISCUSSION OF CHANGES ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP) i TECHNICAL CHANGES - LESS RESTRICTIVE ! Comparing the two safety factors, adequate assurance exists that , sufficient safety margin will still be present to protect the credited ! charcoal efficiency in the dose calculations. Recommendation: The laboratory testing of the charcoal in the MCREC system filters will be ; conducted using the ASTM D3803-1989 standard to demonstrate s 2.0% methyl iodide penetration for 2-inch bed depths at 30 C and 95% RH. This will maintain or exceed the charcoal efficiency credited in the FSAR accident analyses and with an established safety margin. j l O 1 O HATCH UNIT 2 f( REVISION g

fpecI caTecn f,4 ADMINISTRATIVE CONTROLS - 1 l l f,LL{MONTHLYOPERATINGREPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basisito the Director, Office of Management (Fnd Program Analysis, U. 5. Nuclear Regulatory Commission, Washington,

           ,D. C. 20555, with a co)y to the Regional Office of Inspection and 88      Enforcementfno later tian tne 45cn vi eat.n montn tollowing the calendar ifRinth covered by the report.

f f,y CORE OPERATING LIMITS REPORT g,f5,a6.9.1.ll.a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following: f,f,f,g,l)(1) Control Rod Program Controls - Rod Block Monitor for Specification 3.1.4.3, f,f. f , g a.) (2) The Average Planar Linear Heat Generation Rate for Specification 3.2.1 and Surveillance Requirement 4.2.1, f,4,5, 4, 3) (3) The Minimum Critical Power Ratio for Specifications 3.1.4.3 and 3.2.3 and Surveillance Requirement 4.2.3, and g*y ( The Linear eat Generatio RateforSpecif\ cation 3.2.4 and Surveill ce Requiremen 4.2.4. \ f*& , 5. /, b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the following documents. f,6.5.f..d (1) NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," (applicable amendment specified in the CORE OPERATING LIMITS REPORT). f,6. f, 4.3.) (2) " Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment Nos. 151 and 89 to Facility Operating Licenses DPR-57 and NPF-5," dated January 22, 1988. f, f . 5, c. c . The core operating limite shall ba detacmined so that all appliccble limits (e.g., fuel therma hmechanical limits, core thermat-nydraulic limi u , ECCS limits, nuclear limits such as shutdown margin, and trattsient and accident analysis limitc) 9 of the safety analysis are met. 5,6, $, ) d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be )rovided upon 9 issuance,foreachreloadcycle,totheNR7DocumentControP esk with copies to the Kegional Aaministrator and Resident pector. A.l HATCH - UNIT 2 6-14d Amendment No. 48, 86, 406, 129

M t

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l l l DISCUSSION OF CHANGES ITS: SECTION 5.6 - REPORTING REQUIREMENTS i TECHNICAL CHANGES - MORE RESTRICTIVE > M.1 The current TS requirement in 6.9.1.5.b to submit an annual report for all challenges to safety / relief valves has been moved to proposed ITS 5.6.1.4 for monthly reports. Since the report is required on a monthly basis instead of the current annual basis, this change is more restrictive in nature. j M.2 This change details the information to be included in the report. These ! details are necessary to assure the reports are provided with similar . content and format for comparison with other plants and with prior j reports. ; TECHNICAL CHANGE - LESS RESTRICTIVE

 " Generic" LA.1 The details associated with CTS 6.9.1.1, 6.9.1.2, and 6.9.1.3, " Start-Up Report," are proposed to be relocated to the FSAR. The Start-Up Report provides the NRC a mechanism to review the appropriateness of licensee      ,

activities after-the-fact, but provides no regulatory authority once the i report is submitted (i.e., no requirement for NRC approval). The Quality Assurance requirements of 10 CFR 50, Appendix B, and the Startup Test Program provisions contained in the FSAR provide assurance the listed ! activities will be adequately performed and that appropriate corrective actions, if required, are taken. The placement of these CTS requirements in the FSAR also ensures that change control is performed in accordance with 10 CFR 50.59. i l l l l O i HATCH UNIT 2 3 REVISION 1

O UNIT 2 NO SIGNIFICANT HAZARDS DETERMINATION O O

 - e   a                                  _        -      -     ._m-N0 SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.7 - STANDBY GAS TREATMENT SYSTEM-0PERATING L.2 CHANGE Not used.

i O .

HATCH UNIT 2 2 REVISIONf[

NO SIGNIFICANT HAZARDS DETERMINATION O ITS: SECTION 5.5,7 - VENTILATION FILTER TESTING PROGRAM (VFTP) L.1 CHANGE Not used. O ! O HATCH UNIT 2 1 REVISIONfh

a 4 9_. 4_,m.=--4.e. ,,._.m....a&_ .a#. - , . . . . - . _ . . . - . . . . -a .-o..---_ ..x. . , .. ... ... i.~ -. s. --. m .. . . _ . ;-4.. _ .

1 l N0 SIGNIFICANT HAZARDS DETERMINATION ! Og ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP) [ L.3 CHANGE j l Not used. e i i i

4 l i i l

I

i i r 1 i i k 1

                                                                                                                                                                                   -t i

l l HATCH UNIT-2 3 REVISION

I (q v NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 5.5.7 - VENTILATION FILTER TESTING PROGRAM (VFTP) L.4 CHANGE ; In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Current Technical Specifications specify charcoal laboratory testing acceptance criterion as "in accordance with Regulatory Guide 1.52." This testing is to ensure that the charcoal adsorber efficiency assumed in the accident analyses is met. It has been demonstrated the 95% average filter efficiency assumed in the accident analyses will be ensured if the laboratory testing acceptance criterion is 2% methyl iodide penetration. The Main Control Room Environmental Control System is not an accident initiator in any previously evaluated accident. Therefore, the change in acceptance criterion will not increase the probability of an accident previously evaluated. Since the proposed laboratory testing acceptance criterion will still ensure that the filter efficiency assumed in the accident analyses is met, the proposed change does not involve a significant increase in the consequences of an accident previously O evaluated. d 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not introduce a new mode of plant operation and does not involve physical modifications to the plant. Therefore, it does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

This change does not involve a significant reduction in a margin of safety, since the proposed change will continue to ensure charcoal adsorber removal efficiencies assumed in the accident analyses. k HATCH UNIT 2 f% REVISIONfhi

I NUREG 1433 COMPARISON DOCUMENT - SPECIFICATIONS O O l

Definitions 1.1 1.1 Definitions LEAKAGE 2. LEAKAGE into the drywell atmosphere from (continued) sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; i

c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE;

d. Pressure Boundary LEAKAGE l LEAKAGE through a nonisolable fault in a b,> Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

e J LI R HEAT GENERATION Th HGR shall be the heat generat rate per ! f RATE GR) unit th of fuel rod. It is the i gral of : the heat over the heat transfer are associated wi e unit lenath. LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST or all+1ogic components (i.e., all& relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the

                      #g
 $a/t,%          MJ               sensor as practicable up to, but not including, v                                the actuated device, to verify OPERABILITY. The.            .

LOGIC SYSTEM FUNCTIONAL TEST may be perforined by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested. ; D shall be the largest valite of the [ OF IMUM FRACTION ITING The M fraction f limiting power density in he core. The fractio of limiting power density all be POWER ITY (MFLPD) the LHGR exis ' at a given location div (ed by the specified LHG t for that bundle type z , 4 O <ceetinuee) BWR/4 STS 1.1-5 Rev. O,09/28/92 Y

1 Definitions i 1.1 l 1.1 Definitions (continued) MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core %for each class of fuely. The CPR is that power in the f,( assembly that is calculated by application of the appropriate correlation (s) to cause some point in the assembly to experience boiling transition, ' divided by the actual assembly operating power. MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. OPERABLE-0PERABILITY A system. subsys m, o , or device t s all be OPERABL when it is capable of perfoming g kW its specified safety function (s) and when all p f+Mg necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and gf. seal water, lubrication, and other auxiliary equipment that are required for the system,

                       . .go        subsystem,,t* component, or device to perform its specified safety function (s) are also capable of performing their related support function (s).

O6M3 _ PHYSICS TESTS PHYSICS TESTS shall be those tests perfomed to measure the fundamental nuclear characteristics of 'b the reactor core and related instrumentation. ud 2 f,6 These tests are: ,m ,,

   ' f d '>

h a. r tA,) pusn

                                )        Described in Chapterd14_, Initial Tests ol oped,L M1 ProgramKoftheFSAR;                             ~

D N [ b. Authorized under the provisions of g ['dct T

                    ,,                   10 CFR 50.59; or

[1t # c. Otherwise approved by the Nuclear Regulatory Commission. PRES RE AND The PTLR is e unit specific do ment that TEM RATURE LIMIT provides th reactor vessel pre ure and ' RT (PTLR) temperatur limits, including atup and co down ; rates, f the current reacto vessel flu ce period. These pressure and emperature mits (continued) O BWR/4 STS 1.1-6 Rev. O, 09/28/92

Definitions 1.1 1.1 Definitions PRES RE AND shall e determined fo each fl nce d n TE ERATURE LIMITS accor ance with Speci cation .0.1.77 P1 t p,h 'R ORT (PTLR) oper tion within the opera ng limi i

       .(continued) -

ad essed in LCO 3. 38, "R Pressu ad TT perature (P/T) imits." RATED THERMAL POWER RTP shall be a total reactor core heat transf (RTP) rate to the reactor coolant of f2436KMWt. gf REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be suberitical assuming that: '

a. The reactor is xenon free;

b. The moderator temperature is 68'F; and 6f,7 c- All control rods are fully inserted excePt for the single control rod of highest reactivity .

worth, which is assumed to be fully withdrawny Gith control rods not capable of being fully inserted, the reactivity worth of these control , rods must be accounted for in-the determination of - SDM. STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the l testing of one of the systems, subsystems, channels, or other designated components during l the interval specified by the Surveillance l Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, l subsystems, channels, or other designated i components in the associated function. ' (continued) BWR/4 STS 1.1-7 Rev. O, 09/28/92

I I Definitions 1.1 1.1 Definitions (continued) THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

                                                                                         -          l TURBINE BYPASS SYSTEM        The TURBINE BYPASS SYSTEM RESPONSE TIME consists                  i RESPONSE TIME                of two components:                                                l

a. The time initial movement of the main turbine stop valve or control valve until 80%

of the turbine bypass capacity is established; g \[ ,

                /G 4 .3 and Av                                        /

( ; l / j b. The timeJef initial movement of the main p, h' V turbine stop valve or control valve until initial movement of the turbine bypass valve. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. O i

                                                                                                 *i 0

BWR/4 STS 1.1-8 Rev. O, 09/28/92

Feedwater and Main Turbine Trip Instrumentation ! 3.3.2.2 . I SURWILLANCE REQUIREMENTS

       ...................................--NOTE-------------------------------                   ..-..

When a chatnel is placed in an inoperable status solely for perfomance of required Surveillances, entry into associated Conditions and Required Actions may be is capability delayed for up to 6 hours provided feedwater maintained. , and main turbineptrip.

                                                                           .. 4.,.f,7..j eget.)3 SURVEILLANCE                                  FREQUENCY 3%     SR 3.3.2              fez imwCiiAnnEt CHECK.                            ?' hants X                                                                                            -

SR 3.3.2.2. Perfom CHANNEL FUNCTIONAL TEST. (92 ays l

              /i@-                                                                                                                       l SR 3.3.2.2.1             Perfom CHANNEL CALIBRATION. The               ,(18]< months Allowable Value shall be s       inches.

O k (55 5}utanty / cay - k*\ SR 3.3.2.2.( Perfom LOGIC SYSTEM FUNCTIONAL TEST (18Fmonths including (valve}(actuation. ,- (p's@ ; qp e-v O BWR/4 STS 3.3-21 Rev. O, 09/28/92

PAM Instrumentation 3.3.3.1 3.3 INSTRUMENTATION 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1-1 shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS

   -------------------------------------NOTES------------------------------------

1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required 30 days with one required channel to OPERABLE channel inoperable. status. B. Required Action and B.1 Initiate action in Immediately l associated Completion accordance with Time of Condition A Speci-fication G i not met. E l'gpg i A I . . . (a,ll bu0 C. - - - - - - -- NOT E- - - --- -Q C.1 Restore /onerequired 7 days NotDpp e to channel to OPERABLE [ hydr n sbnttor] status. N g\{ t . . . . . . . . . . . . . . . . . . . . . g,d chaffnels. .,q One or more Functions with twgrequired channels) inoperable. W mo^G

                           },\                                                          (continued)

BWR/4 STS 3.3-22 Rev. O, 09/28/92 l

PAM Instrumentation 3.3.3.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME

N -{rza ired hydrogen D.1 Restore one [ required '2 .vurs monitor] chan - h 2 ;; .. munitorj inoperable. channeT t %

           '                                    status.

g f.1 Enter the Condition Inunediately

    'g    d . Required    Action associated       and Completion  AP    referenced in                                   '

U(/ Time of Condition C L/ Table 3.3.3.1-1 for see not met. the channel.

1. As required by K.1 Be in MODE 3. 12 hours j '

Required Action f.1 y; g and referenced in E ' Table 3.3.3.1-1. ! O h4. As required by I p A.1 Initiate action in Immediately E/ Required Action .1 A accordance with I

             % and referenced in Table 3.3.3.1-1.       @      Specification 5"
                                                                                       >(       l

(\ l- A .%.2.c.. A;

                                                                                              .i
                                               .b                                                 l l

l O BWR/4 STS 3.3-23 Rev. O, 09/28/92 i

PAM Instrumentation 3.3.3.1 SURVEILLANCE REQUIREHENTS j, ........................-------------NOT ------------------------------------ These SRs apply to each function in Table 3.3.3.1-1. ~ SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK. 31 days SR 3.3.3.1.2 Perfom CHANNEL CALIBRATION. ;{18) months k _ I 4 O BWR/4 STS 3.3-24 Rev. O, 09/28/92 1

ECCS Instru'nentation 3.3.5.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME l

G. As required by G.1 -------- E-- ----- l Required Action A.1 On pplic le r l ncti s 4.c 4.e and referenced in Table 3.3.5.1-1. .e. fq 4. , 4.g, 5.c, 5.f, and g. t.....................; ; Declare ADS valves I hour from inoperable. discovery of : loss of ADS ! initiation capability in i both trip : systems ! AND G.2 Restore channel to 96 hours from OPERABLE status. discovery of inoperable channel concurrent with O HPCI or RCIC inoperable r r M . , 8daysf th Ji I i g.\ ere B l i H. Required Action and H.1 Declare associated Inmediately associated Completion supported feature (s) Time of Condition B, inoperable. C, D, E, F, or G not

met.

E o ! BWR/4 STS 3.3-39 Rev. O, 09/28/92 ! I l

ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREHENTS

         -------------------------------------NOTES------------------------------------

1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be dela ed as follows: (a) for up to 6 hours for Functions 3.c anJ; and (b) for up to 6 hours for Functions other p

j. li than 3.cp 3.f. provided the associated Function or the redundant Function maintains initiation capability.

         ..... __ ...-.......              ......... ........... ._....................____..                       g 9.n SURVEILLANCE                                                  FREQUENCY SR     3.3.5.1.1    Perform CHANNEL CHECK.                                          12 hours SR     3.3.5.1.2    Perform CHANNEL FUNCTIONAL TEST.                                (92Fdays 3.3.5.1 1    La iprite-the-trip-=it.                                        ] 92 [ 5 7i O

h~\. y --- Perform CHANNEL CALIBRATION. 92 days x j,SR3.3.5.1

SR 3.3.5.1. Perform CHANNEL CALIBRATION. W (18Pmonths g;

    & SR          3.3.5.1. Perform LOGIC SYSTEM FUNCTIONAL TEST.                           418      onths
    ) .\

t (*\

dio SR 3.3.5.1. Verify the ECCS RESPONSE TIME is within 4.1 months on limits. a STAGGERED h TEST BASIS e

BWR/4 STS 3.3-40 Rev. O, 09/28/92

LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS .,e~ N !

A .. .. C .'.f.... .........-NOT S- ---- .S ' OO- -- . . .. . ..

1. I Refer to Table 3.3.8.1 1 to detemin ich SRs apply for each LOP l Function. <

l M-chenheHs place'd-in-an'TnoperabMohikfor perfomance of ! Q required Surveillances, entry i .ci d Conditions and Requir,ed hour prov de6'theiassWeiatet! Fun'ction Acti_ons ma be delayed for up o maintain

           ...s..Q init
                                      .ptio(cyb,lityg g (Q
                                                                                 ........b...c-}    jog (cmf.2)?
            / arri cauunciecHon conlailiiv (fu Twcfio,13W r

V m__ / ' - -SUR9EILLAN'CE ' " " \"

                                                                                                  'FMI s ,-

3 v.- - x s 7 SR 3.3.8.1.1 Perfom CHANNEL CHECK. 12 hours - l -._ ( SR 3%bb Perform CHANNEL FUNCTIONAL TEST.

                                                                      ./ ~~           - Q       r w_.f ~

31 days wi j

SR 3.3.8.lg Perfom CHANNEL CALIBRATION. fmonths SR 3.3.8.1 k Perfom LOGIC SYSTEM FUNCTIONAL TEST. (18Fmonths u : l I l O BWR/4 STS 3.3 ~5 Rev. O, 09/28/92

LOP Instrumentation 3.3.8.1 Table 3.3.8.1 1 (page 1 of 1) { } Loss of Power Instrumentation REQUIRED CMANNELS SURVEILLANCE ALLOWABLE VALUE FUNCTION % s ( PER SUS. REQUIREMENTS 1. 4.16 kV Emergency Bus undervoltage V m \ (Loss of Voltage)

                                                                                         ;^4 :. .;.i.;i                                                   2800 N = ; - l

a. Bus Undervoltate MZ/ l SR 3.3.8.1.2 t

sa 3.3.8.1.3 st 3.3.8.1.4

b. Time Detsy p2A tst 3.3.8.1.23 SR 3.3.8.1.3 2: : - - - _ .:

sp.5geconds SR 3.3.8.1.4 ( 2.' 4.16 kV Emergency sus Undervoltage h (Degraced Voltage) \

a. Bus Undervoltage W23.L/ _ _ . _ . . . . . .

                                                                                                                                                        , [32801 V w ' ' 'I st    3.3.8.1.2
                                                                                           $4 3.3.8.1.3 st 3.3.8.1.4                 /

b. Time Delay WZW tst - .~

   '- -                                                                                          3.3.8.1.21 sa 3.3.8.1.3            $                        [5 ' ' '.5     1    seconds st 3.3.8.1.4 ,

V Gev .t 2 a 3,3.s. i. I 1 3626V l llwiuvork c nurMion c) bus Ued ve.41 $ C B 3 8d-I a 3 3.8.i.3 C hs (Lwbvot4ncy sR 3. 3. 3. l 4 t b, Sme. M y :59, 3.3.6. l.2. 4 (d) 'eco45 j

                                                                                                                                                                                    }      ,

ijn:h vn.41 g5A 3. 333.1. 3'.4O' I'3 , C P.70 O BWR/4 STS 3.3-76 Rev. O, 09/28/92

RHR Shutd:wn Cooling System-Cold Shutdown 3.4J ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown B.1 Verify reactor 1 hour from coeling subsystem in coolant circulating discovery of no operation. by an alternate reactor coolant method. circulation AND AND i No recirculation pump in operation. Once per 12 hours thereafter AND B.2 Monitor reactor Once per hour coolant temperature. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY , SR 3.4. 1 Verify one RHR shutdown cooling subsystem 12 hours or recirculation pump is operating. i l O BWR/4 STS 3.4-23 Rev. O,09/28/92

RCS P/T Licits 3.4.}# 4 3.4 REACTOR COOLANT SYSTEM (RCS) 0k 3.4. M RCS Pressure and Temperature (P/T) Limits 4 LCO 3.4.M RCS pressure, RCS temperature, RCS heatup and cooldown g rates, and the recirculation pump starting temperature , requirements shall be maintained within h limits, specifierL m um m-APPLICABILITY: At all times. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1 Restore parameter (s) 30 minutes Required Action A.2 to within limits. shall be completed if this Condition is AND entered.

     ----------------------              A.2     Determine RCS is        72 hours acceptable for Requirements of the                         continued operation.

LC0 not met in MODES 1, 2, and 3. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met. B.2 Be in MODE 4. 36 hours (continued) 3.4-24 Rev. O, 09/28/92 O BWR/4 ST

RCS P/T Lisits 3.4./ i

                                                                                                               ' h ACTIONS (continued)                                                                                                            !

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE--------- C.1 Initiate action to Immediately , Required Action C.2 restoreparameter(s) ; shall be completed if to within limits. - this Condition is entered. AND C.2 Detemine RCS is Prior to i Requirements of the acceptable for entering MODE 2 LCO not met in other than MODES 1, 2 operation. or 3 P.3 lk : i and 3. , SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4. .1 -------------------NOTE-------------------- Only required to be performed during RCS

               ;--              heatup and cooldown operations and RCS inservice leak and hydrostatic testing.                                                              ,

a. W esyt,RCS pressure ERCS temperatur ,,q E 30 minutes '

P.D b. RCS heatup and cooldown ratessere witninx ghe limits specified in th-- FA PMureaM-1 W S.4.4 '2 )

                                                                                                                                }ki (sici s too'r in aos 1. kow ouiadM                                                                 ,
                                                                                                                                   'j SR  3.4.         .2      Verify RCS pressure and RCS temperature are                          Once within                       i within the criticality limits specified in                           15 minutes                       !

m N "' Z. p;pe 3.9 9. g] prior to control rod

                  '4/)                    E-withdrawal for
          ~\

the purpose of achieving criticality (continued) i O BWR/4 ST 3.4-25 Rev. O, 09/28/92 I

                                                                                                                            .         l

RCS P/T Linits 3.4.)2 l

                                                                                     @4              l SURVEILLANCE REQUIREMENTS          (continued)

SURVEILLANCE FREQUENCY Cl SR 3.4.]#.3 --------------------NOTE------------------- p 3l Only required to be met in MODES 1, 2, 3, 7 q' f% and+ 4hith 2: peig]m rc;ct;r :ter dre prers"~ % dvdng ' l\ i Verify the difference between the bottom 0 c^ ri t"- 1/ head coolant temperature and the reactor 15 minutes pressure vessel (RPV) coolant temperature zu - d is Qn the lia.ita sph;i;md ... the TTLR. startup of a re culation f.30 ($ Ht *F] _ 9 SR 3. 4. Mf.4 -------------------NOTE-------------------- Only required to be met in MODES 1, 2, 3, P.b L M Verify the difference between the reactor f - e = H " rr coolant temperature in the recirculation 15 minutes : loop to be started and the RPV coolant ri r " Each temperature is "i+" - U= '~ ;s s --Mird (sTartup of a O':M. z go.p recirculation

                                                                        ' pump .               i f.30 9

SR 3 . 4. W. 5 -------------------NOTE-------------------- Only required to be perfonned when i e' tensioning the reactor vessel head bolting - 1, ),k

         \        studs.

Verify reactor vessel flange and head 30 minutes flange temperatures are 'iihi- 6; 'i=9. 6 dp t!#ir" 4" !"? " : 's. y 74.

FJ--} ut

(@ iw e) 1+ (continued) 1 BWR/4 ST 3.4-26 Rev. O, 09/28/92

1 RCS P/T Liaits I 3.4.pf 9 p, k s SURVEILLANCE REQUIREMENTS (continued) N SURVEILLANCE FREQUENCY l 9 SR 3.4.W.6 -------------------NOTE-------- --------- Not required to be performed until 30 minutes after RCS temper ture 5

in n[.k MODE 4. p , 3o M f

                                                                                                           \    ,
                   --...-....-_....____ ....                         .. -           at                          r Verify reactor vessel flange and" head                               30 minutes            /

flange temperatures are - ith' ' :=. m f

                           'i-...s... F-.

1 7 c ' F.) } %1 h30 i 90 r3 1 ut. 9 SR 3. 4. W. 7 -------------------NOTE------------------- Not required to be perfomed until 12 hours p,b after RCS temperature s 0 'F in MODE 4. i OG m ' i ut Verify reactor vessel lange head 12 hours flange temperatures are

                                                              'e    -- :-.i h                              (

g '

-j: , . : n; 0 y_ 7c,

Dj j ui i :

h.30 g 3a 'Fj j us O _ P t l t 6 O Rev. O,09/28/92 BWR/4 ST 3.4-27 i

I RCS P/T Limits 3.4.9

G I EFPY 8 to 12 14 16

1200 1 ADJUSTED CORE BELTLINE, 1/4 T FLAW Q $ 1000 > g O I 5 ^ a j p*30 2 800 , "e / \e A

O a 000 E

VF.RTICAL LIMIT LINE FOR PRESSURE ABOVE 20% HYDROTEST (312 pas), 400 mASED ON 10CFR50 APPENOlX 0 REOulREMENT OF (RTNDT + 90*F), FLANGE REGION RTNDT = 16*F I BOLT FRELOAD TEMPERATURE OF 79'F BASED ON RECOMMENDED 200 , (RTway + o0er) FoR 0.24-iN. FLAW IN CLOSURE FLANGE REGION, RTNOT

1FF 0

0 100 200 300 RPV METAL TEMPERATURE (OF) Figure 3.4.9-1 (page 1 of 1) Temperature / Pressure Limits for ' Inservice Hydrostatic and Inservice Leakage Tests HATCH UNIT 1 p q 3.4- 27A 1

l RCS P/T Limits

3.4.9 O ;

1600 VALID TO IS EFFECTIVE FULL POWER YEARS OF OPERATION 1400

         -    1200 O                                                                                         A b

z l

n. 1000 ;

o I ADJUSTED CORE BELTLINE lAl 1/4 7 FLAW, RTNOT = 10*F E IRRADIATION SHIFT = 133*F 7 g 800 . 5 Y b N s-a

>=

a iLE 400 FEEDWATER NOZZLE TEMPERATURE

                                         / LIMITRESULTS FORADJUSTED 1/4 TTOFLAW  40*F RT NOT(BWR/S I

200 MINIMUM OPERATING TEMPERATURE

                             /               OF 78'F BASED ON REP-NDED (RTNOT + GO'F) FOR 0.24 IN. FLAW IN CLOSURE FLANGE REGION, t

RTNDT

18'F 0 1 I O 100 200 300 400 500 600 MINIMUM VESSEL METAL TEMPERATURE (OF) 1 Figure 3.4.9-2 (page 1 of 1) ,

Temperature / Pressure Limits for Non-Nuclear Heatup, I Low Power Physics Tests, and Cooldown Following a Shutdown l l HATCH UNIT I 1 3.4- % l

RCS P/T Limits 3.4.9 1600 0 VALID TO 16 EFFECTIVE FULL POWER YEARS OF OPERATION 1400 79

8. JL E 1200 5

z I O e

-                                                                                '0 2 1000 uJ
                                                                                 - { .30 E                                             ADJUSTED CORE BELTLINE.

1/4 T FLAW, RTNDT

I@E' IRRADIATION SHIFT = 123*F E 800 II d

a 5 E gi l 600 W SM M 400 f FEEDWATER NOZZLE TEMPERATURE

LIMIT FOR 1/4 T FLAW (SWR /6 RESULTS ADJUSTED TO 40*F RTNOTI 200 MINIMUM OPERATING TEMPERATURE LIMIT OF 78 8F FROM 10CFR60 APPENDIX O

              /              REOuiREMENT TNAT <Tuiu - RTuoy + e0ar i.

FLANGE RTNDT*18 E l 0 100 200 300 400 500 600 MINIMUM VESSEL METAL TEMPERATURE (OF) Figure 3.4.9-3 (page 1 of 1) Temperature / Pressure Limits for Criticality 9 HATCH UNIT 1 3.4-d7b

l I RCS P/T Limits 1 3.4.9 e k 1600 A' A 1400 . r e a

       ^                                 w i                              ,'

O 1200 e '

A z

5 . ; I Q. i S 1000 , P.3Q ; N10 A' - CORE BELTUNE

        $                                                   AFTER ASSUMED 32 EFPY g  800                                              SHIFT FROM AN INITML O                                                   WELD RTcOF -SO*F U

b E O E z 600 A - SYSTEM HYDROTEST UMIT wrTH rUEL IN VESSEL t: 2 3

                                                       - VESSEL DISCONTINUTTY N 400                                                 uurTs
       @                                               = = CORE BELTLINE WITH to         312 Psc  .-                                32 EFPY SHIFT

n. 1 200 gotyyp CURVE A* IS NOT UMITING po*r FOR INFORMATION ONLY CURVE A IS VAUD TOR 32 EFPY OF OPERATION O , , ,

0 100 200 300 400 500 600 ; 64NIMUM REACTOR VESSEL METAL TEMPERATURE fr) Figure 3.4.9-1 (page 1 of 1) I Temperature / Pressure Limits for

 /s             Inservice Hydrostatic and Inservice Leakage Tests l

l i HATCHUNITfOh 3.4-@[D l j

RCS P/T Limits 3.4.9 I I 1600 l B' 8 1400 . o k a h k v 1200 , r \ a b e z 3 7 30 1 S 1000 , _i

                                   '                                                      1 g

B' - CORE BELTUNE g AFTER ASSuuED 32 EFPY m 800 shirt rROM AN INITIAL

      $                                                       WELD RTaOF -50'F O

b z 8 - Eg$$HEATUP/ l z 600 t-s

f J

                                                         - VESSEL DSCONDNUITY w                                                        UulTS g 400                                              - = CORE BELTLINE WITH m

y 312 esc ) 32 EFPY SHIFT m 1 ' 200 CURE B' IS NOT UMITING soLTUP po*r # [ FOR INFORMATION ONLY CURVE B IS VAUD FOR 32 EFPY OF OPERATION g l i i 0 100 200 300 400 500 600 MIN:UUM REACTOR VESSEL METAL TEMPERATURE fF) Figure 3.4.9-2 (page 1 of 1) Temperature / Pressure Limits for Non-Nuclear Heatup, Low Power Physics Tests, and Cooldown Following a Shutdown HATCH UNIT 2 1gg 3.4 c27 h

RCS P/T Limits 3.4.9 O 1600 C' C 1400 .i i 9 8

    '{1200-                                  ,"                                            j g O                                      t b

x

k n ,

                                                                                                  )
     ~                                  ,                                                      /      .

J W C' - CORE BELTUNE g in AFTER ASSUMED 32 EFPY y 500 SHIFT FROM AN INfTIAL

                                                                                          ]f WELD RTaOF -50*F O
                                                           ~

600 u a i t: l 3

      "                             f
                                                        - VCESEL DISCONTINUITY uum 400
      $                           f                     = - CORE BELTLINE WITH 32 EFPY SHIFT D         312 Psc          J i         pp             (
                                /

CURVE C' f5 NOT LIMITING poo r

                           /
                             /                          FOR INFORMATION ONLY CURVE C 15 VAUD FOR 32 EFPY OF OPERATON O             f O         100      200                300         400       500        600 MNIMUM REACTOR VESSEL METAL TEMPERATURE fF)

Figure 3.4.9-3 (page 1 of 1) Temperature / Pressure Limits for Criticality G V HATCH UNIT 2 On\y 3 4-27f

Reactor Steal Dome Pressure i

3. 4.J4' 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.M' Reactor Steam Dome Pressure p 10 C0 3.4. Thereactorsteamdomepressureshallbes[1020 psig.

fa) APPLICABILITY: MODES I and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within dome pressure to limit. within limit. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l IO SR 3.4.M.1 Verify reactor steam dome pressure is 12 hours 1020Ppsig. 5 O BWR/4 STS 3.4-28 Rev. O, 09/28/92 l

        % l.lN17 l    NO ION                                     SGT System y                                 3.6.4.3   "

j 3g 4 [.p w,f q

                                               ,buel SURVEILLANCE REQUIREMENTS SURVEILLANCE                         FREQUENCY (he9me c d3 SR   3.6.4.3.1    Operate eachpui subsystem for a continuous hours with heaters 0[    31 days              '

operating p,1 SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP

TestingProgram(VFTP).

pegared) p. TQ SR 3.6.4.3.3 Verify each SGT subsystem actuates on an (18) months ; actual or simulated initiation signal. ;

e , SR 3.6.4.3.4 Verify each SGT filter enn4+r Lypa>> [tal months d irr ca.. uw opWd and the fan started. N O l

l

                                                                                      .l O                                                                                         ,

BWR/4 STS 3.6-55 Rev. O, 09/28/92 i

                                                                                            - Opre bn 2 k o#  J/                                                   >[ Secondary), Containment 3.6.4.1 m

7 3.6 CONTAINMENT SYSTEMS - b

                                                   - O eea 6nh                  -

3.6.4.1 (Secondary) Containment - g O12d 2 ise co's datf

                                            /    /)                                   A     tt 6a.:tz.meILI h N C0 3.6.4.1             The secondaryh containment shall be OPERABLE.

P. L APPLICABILITY: MODES 1, 2, and 3 Duran movement of irradiat ' el assemblies in the

                                      ,seco           n         nt, During CORE A                ,

[3 Durin tions with a poten vessel (OPDRVs). r draining the reactor 4 cd Q ACTIONS v M#d , CONDITION REQUIRED ACTION COMPLETION TIME i (Ao t on e oF fa th (Ort 6 l oad unit 2) t(Secondary}cAmf 5 / A.1 Restorej{ secondary - 4 hours f2containmentfinoperable. containmen to i- "^DE 1, 2. er A OPERABLE status. O B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time cf C:nditi:n " - AND not met. B.2 Be in MODE 4. 36 hours t condary] C.1 --------NOTE-- ------ cent t inoperable LCO 3.0.3 is not during movem f applicable. irradiated fuel --------------- --- assemblies in the [ secondary] d movement of Immediately containment, during irradia el CORE ALTERATIONS, or assemblies in during OPDRVs. [ secondary] containment.

AND d / (continued) S 3.6-46 Rev. O, 09/28/92 BWR/4 STS

        $ b(Jg-f L V d Y                                                               SGT System (

N[Clogc wf 3 . 6. 4 .'3 spcina, ~hael SURVEILLANCE REQUIREMENTS j < SURVEILLANCE FREQUENCY

                                                                                                             }
                                                                                                             ?

OperateeachkGTsubsystemfora}10 31 days SR 3.6.4./.1 j continuoushoursfwithheaters 7 operating & ! SR 3.6.4.8.2 Perform required SGT filter testing in In accordance accordancewithN.theVentilationFilter with the VFTP o'7 Testing Program (VFTP). k unt / a ud unit 2.Y SR 3.6.4.Y.3 Verify each SGT subsystem actuates on an f18(months i actual or simulated initiation signal.

5R 3.6.4.5.4 e r i fy ;;;h SET H1+ar raa1er by;;;;

                                ,                                             [i6] munua                 -

d-a can De opened and tne i n at:-+ed g3 - . r i A i i i O O, 09/28/92 i BWR/4 STS 3.6-55 Rev. l l i

   - X & ,f 2 f/ary b y                                                                        DN       S     l SGT System      j

[. 1 7 4Ifc 6 57 P J *4M $c.My 3.6.41-duwbue} (g 3.6 CONTAINMENT SYSTEMS 3.6.4.A Standby Gas Treatment (SGT) System % - uw$2-) & LC0 3.6.4.) wo} SGT subsystems shall be OPERABLE. A APPLICABILITY: -MGDE5 1, 2, and 3,- De. . n mv v em.. ef .. .ediet d fuel = s t e-bl i n in the 500:nd r;') "nn+ninment. __

                            -Dering 00"E ALTEn m nwe During operations with a potential for draining the reactor vessel (0PDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ynet 23 A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable. to OPERABLE status. B. i<equired Ad4en-and B.1 Be in MODE 3. 17 MW O associated Completion Time of Condition A AN not met in MOD N p- B.2 Be in MODE 4. 36 hours [. , Required Action and ------------NOTE-------------

 $         associated Completion             LCO 3.0.3 is not applicable.

Time of C;nditi;n A ----------------------- ----- not met,& ring GAbc g)

           - .  ...s . 6 w.  ....u.o M      Y.1             PlaceOPERABLEJSGT         Immediately fee! ::::dlie: S the                             subsystem in

{ secondary] $ operation. catainment, during CORE-ALTERATIONS M r QR

          .during-0PDRVs.

(continued) i O l BWR/4 STS 3.6-53 Rev. O, 09/28/92 ; l

k ad 2 Va_risa (--OPPRVs SGT System

    'f p Al(chug 3 pd                                                                                                                     3.6.4 5 M*<

D ACTIONS

                            $(#cifhlll J

CONDITION REQUIRED ACTION COMPLETION TIME

4. (continued) k.2TSuspend movemenhof Imediately S

irradiated fuel 8 5 5 '"b "" ' " \ l i [ secondary , containment. - AN C.2)2 Suspend RE Imediately ALTERATION . iAN -

                                                               . C . 2.0 - Initiate action to                              Imediately d           suspend OPDRVs.

[utrit 2 ) I

t. Two/SGT subsystems D.1 --------NOTE--------- . 'N inoperable. W LCO 3.0.3 i OC ef:::rt of ...ediated feel a>>cmL;;e; ir, t.k

                            -{sewndary] -
                                                                )'

applicable, conta4nment. 6 -4na - Suspen vement of Imme stely'

                             ^GRE AtTERAT E 3, m -                       irradiated                   i                               N           ,

dur1Tig1PDRVs----- assemblies in l [ secondary] containment. N ; N

                                                                   .2    Suspend                                           Immediately            !

ALTERATIONS. t QD S i 4r& Initiate action to Immediately suspend OPDRVs.

                                    */'M Ac1W                 cddeA f   & R-o'l,I                    C.D                  AAY a

O BWR/4 STS 3.6-54 Rev. O, 09/28/92

     .)e ((p,f 'L Vaf]Ic>?                                                   nPPRV.s SGT System f15clov$esNot                                                            3.6.4.

NN gg ;pc,(f7 vu W d SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.i.1 OperateeachbTsubsystemfor )J10)t} 31 days A continuous hours with heaters QE g operating SR 3.6.4. 2 Perform require [SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP TestingProgram(VFTP). 3 7 m l uth t 2) Verify each SGT subsystem actuates on an X18[ months SR 3.6.4J.3 actual or simulated initiation signal.

    *    $ R ':.5.d.3.",  Verify ::ch SGT (M +se cooler hvnace        T10] r--th
  ,3A                           r es- M vpenea and the fan stai Ld r l

l l l O BWR/4 STS 3.6-55 Rev. O, 09/28/92

(-Echchu 4 - SGT System L12 kwe 3.6.4.r g ag,y,souroA11 *L>wa 3 g RVEILLANCE REQUIREMENTS SURVEILLANC, M FREQUENCY I L reoume d' un d t wd um 7 2) SR 3.6.4.T.1 Operateeach/SGTsubsystemfora 10 31 days l continuoushoursfwithheaters #2-9 operatingp lum f l u d und 2) SR 3.6.4.I.2 Perfom requiredlSGT filter testing in In accordance accordance with 't'he Ventilation Filter with the VFTP 9 TestingProgram(VFTP

                                      } tc9wtcd f/hn' /aud und 2)

SR 3.6.4. 3 Verify each 18 months actual or s/SGT subsystem imulated actuates initiation on an signal. A_ u- .' 5R 3.6.4.3.4 Verify each SGT f4'ttr ;,,v.cr uyvess [1 Al Jnonths can be opened and the fan started. o a t 6 4 BWR/4 STS 3.6-55 Rev. O, 09/28/92

h5 ystemand[HS l 3.7.2 ; l [ ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME I

f. Required Action and 1 Be in MODE 3. 12 hours O

p associated Completion Time of Condition AND'(4 h .ne 84 not met. 3c, p[ ' f

                                                 /.2       Be in MODE 4.         36 hours
             'Both[PSMsubsystems inoperable for reasons other tha Condition ytand     *.

fd 03 N l NUH[ inoperable.hrr-

                ..._____u__      .%
                   .       u
                                                                                'N days

( Aub onu. wi&ir (2. haves SURVEILLANCE REQUIREMENTS rw wa4e SURV ILLANCE $2 vi

      \_R 3.                                                                                    _      \
                        .1      rify th water leve of each              SW)      24 h rs co ing tow           basin is        ] ft.                                  -

q - W ^4

    /     SR 3.7.2.i         Verify the water level n each PSW pump              2 0 h. . .,

well of the intake structure}Ais a f 60. ft Mmean sea level}}. [imsQ] R 7.2. rify e avera water mperatur of hou [u is [ ) *r. \ (continued) BWR/4 STS 3.7-5 Rev. O, 09/28/92

                                                                                 'hSW[Systemand[HS 3.7.2 SURVEILLANCE REQUIREMENTS                  (continued)

SURVEILLANCE FREQUENCY

3. 2.4 erate e [PSW] oling tower an for 31 day

 ' f.4                               t    5] minu     .                                                           \

L SR 3.7.2./ -------------------NOTE-------------------- , Isolation of flow to individual components ( of SYS M S J ' doesnotrender/PSW)$Systeminoperable. [ Ver eac h s se anu p er 31 days

                               /     operated, and automatic valve in the flow paths servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position, is in the correct position.

CO

                                                          /.                             v SR 3.7.2. Verify each     S    subsystem actuates on an actual or simulated initiation signal.

[8 months O BWR/4 STS 3.7-6 Rev. O, 09/28/92 l

C%;;:td) Programs, and Manuals 60I g S h @ d-.; d Program nd Manuals d Inservice Testing Program (continued) P.9

          @[h              ASME Boiler and Pressure Vessel Code and applicable Addenda l

terminology for Required Frequencies inservice testing for performing inservice ' activities testino activities l Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days 0 b Fvary 9 r.th: 't least ;r.:: per 270 t ys R34 Yearly or annually At least once per 366 days M -ania!!y ;r cvery b A+ 2 y;;7; 1aaet anee ame 713 days h b g. The provisions of SR 3.0.2 are applicable to the above required (requencies for performing inservice testing O activities, P, The provisions of SR 3.0.3 are applicable to inscryice h d. d". testing activities; and , h d t. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to. supersede the requirements of any i

         . 7d.t33- Ventilation Filter Testino Program (VFTPL              P,9     ['((h)7 l,    ,

l f,f, 7 IAprogramshall established to implement the followin required testing of Enginee Safety Feature (ESF) lter ventilat P.i % stems at the freque es specified in [Regu ory Guide arid 4n accordance'with [ latory Guide 1.52 R sion 2; ASME N510-1 9;andAG-1].

f. lb _

QHsekr} (continued) BWR/4 STS 5.0-25 Rev. 0, 09/28/92

l 1 Qr,ccdscer Program and Manuals

                        #6O                                                                             6/'.1
                                                                                                                            , ,f 1

5 @ cdv. d Program and Manuals l

                .r.O
    @W3[ @            '

INSERT a. D3---o- 'M Ventilation Filter Tntina Procram (VFTP)

Demonstrate for each of the ESF systems that an inplace test (continued) of the HEPA filters shows a penetration and system bypass y, p. 2. < 0.05}% when tested in accordance with(Regulatory Guide 1.52, Revision A, and ASME N510-198 at the system M g owrate specified belop 10Mb ogl% f ESF Ventilation System p.% flowrate(cf _

A 56 F Sy5 eFm 3000 +> 4000 b ?Ra cer%) Room Y ' h i X En a o m .no.l 11so To r7.ro c nT.e1( Mc4Ec) 5ys7 ~. (G h b. Demonstrate for each of the ESF systems that an inplace test O'#mgbypass of the charcoal adsorber shows a penetration and system

                                      <M)% when tested in accordance with TRegulatory                                       ,a 3          Guide 1.52. Revision 3,,andASMEN510-198$. at the syste Un h         '

flowrate specified beloQ. {Saction) .1 _ESF Ventilation Syst,gg g.% Ter sysre - CysTn Flowrate (cI-) Too f vooo _/n cRE c. y .g o { a .g o OP.1 . ., c. Demonstrate for each of the ESF systems that a laboratory M *, " S b a test of a sa le of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52. Revision & shows th ASME tir/0-l%9;j methyl todide penetration less than the value s~pecified fecflor 15 a nj below when tested in accordance with fASTM D3803-198 a temperature of s $30'CF and greater than or equal to the g gg relativehumiditylspecifiedbelow. ESF Ventilation System

Penetration (99

Mo) C

                                             .CG V 7.5 sre ~ ~                               O.2 Cfo
                                                                                                    ~

o5M i P.1 _MC rec hs% 7' % _ 9, %_?h]A 5 /g _

                                                                                                                          /A (continued)

BWR/4 STS 5.0-26 Rev. O, 09/28/92

6c+r4 Program / nd Manuals 5

5. Acer:0 Programp nd Manuals P.9 ['

Ventilation Filter Testina Procram (VFTP) (continued) [ ReviewertWote: Allowable pen ration =[100% fficiency f' charcoal credited ' staff safety methylfodh fety factor . luation]/ y 8L Safety ctor=;5ll systems with hea rs.

7. for stems without hea rs. , )

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and

           -                             the charcoal adsorbers is less than the value specified below when tested in accordance with dat s ter" C: Me 1.0G C:nuk,12.29ASMEN510-198yatthesystemf'owrate specifled belo
                                                                         )    gg             .g , g,      pz ESF Ventilatio      ystem            Delta P             Flowrate S                                                 U Tg;, ~               37co c v- pA
                                                          =

cyw-

                                                                             @ a ,. m h um.

c > j [W c j O -l E Demonstrate that the heaters for each of the ESF system ~ dissipate the value specified below [+ -i when tested in i accordancewith[ASMEN510-1989]. K ESF Ventilation System Jattage gy NSbY y '7 feaok ' 15 fo 2o kW y ((vi)O Ul) P1 !_ ' The provisions of SR 3.0.2 and SR 3.0.3 are applicable.to the VTTP test frequencies. , Explosive Gas and Storage Tank Radioactivity Monitoring Progr This program provides cont: or potentially exo9eteve cas _ P, mixtures contained in the) te-Gas Ho p System Xthe q

                             'o        dioact gity co ined in               storage           or fed %to th g of        treatme h syst         d the quantity of radioacuvit
      %;n u            ,       contained in unprotect _          tdoor liquid storace tanksh he op as TneTaa*J                ous ra logy ctivity q ntities shall anch ec determinedfolowingh y

al Position ETSB 11 , '

Post ted Radi ti Release to Waste Ga stem Leak __ iA (continued) BWR/4 STS 5.0-27 Rev. O,09/28/92 i

i l Grec:drcQ Program and Manuals g . S?. I Gf l- _ s- ,

5. Sccda:QPrograms- nd Manuals l

l g .7.2.10 Explosive Gas and Storage Tank Radioactivity Monitorino Procram

                    .r.g3      (continued)                                                   W ure"]. T       iquid radwaste      ntities shal be deter)Kq R 2_O L acc     nce with        dard Review Pla     Section 1 . 3, "Posth Qted   '

(Radioa ve Release e to Tank rm41om=] . j The program shall include:

a. The limits for the concentrations of hydrogen fd cryge' in

           ,                        thefiWane Gai . eld @ Sy:tcyl and a surveillance program to m 'a co,,17,,ser            ensure the limits are maintained. Such limits shall be a7#p., f,,g y ,7             appropriate to the system's design criteria (i.e., whether
       #Y5A                         or not the system is designed to withstand a hydrogen explosion);

surveillance ogram to ensure that the quantityq f [b. ioactivity con ined in [each gas % rage tank an fe P,2 I int he offgas trea nt system] is les than the amo that d result in a le body exposure t 0.5 ren t any indiv al in an unres eted area, in the vent of ; u_ncontrolle 41 ease of the ta 'scontents];an j A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radw tanks that are not surrounded by liners, dikes, or wall , pable-of holding the tanks' contents and that do not ha nk overflows and surrounding area drains connected to the fiquidgadwastegTreatmentfysteeKis less than the amount sat would result in concentrations less than the limits of 10 CFR 20. Apper. dix B. Table E , Column 2, at the nearest P. potable water supply and the nearest surface water supply in - an unrestricted area -in the event of an uncontrolled release of the tanks' contents. i The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the ! Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. :

              " 7.2.b ) Diesel Fuel Oil Testino Proar E-b' y                       A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be establis . The program shall include sampling and testing requiremen ga d P.9 (continued)

BWR/4 STS 5.0-28 Rev. 0, 09/28/92

Reporting Requirements f GP. I gg 5 Reporting Requirements

EI8 Q.a.1 a

rr" e a rre-t: (=tfrrT 0.1.? Radioactive Effluent Release Renort

                                  ~
                 .Cf .3)               ---.-----
                                                       -------..----..------NOTE-------------------------------
                                                                                                                          ~

A single submittal may be made for a multiple unit station. The , Csubmittal should combine sections common to all units at the i f, station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from M i each unit. ' The Radioactive Effluent Release Report covering the operation of i the unit shall be submitted in accordance with 10 CFR 50.36a. The ; report shall include a suunnary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in , conformance with 10 CFR 50.36a an 10 CFR 50, Appendix I, Section IV.B.1. 9 M b.1.b Monthly Doeratino Reports mnAm sh- . Routine reports of operatin riencey, includin valves,$g documentation of !11 challenges ae

                                                  .shall be submitted on a month y basis no later than the to the; l

15th of each month following the calendar month covered by the report.

                    . . .M           C0RE OPERATING LIMITS REPORT (COLRL
      .GP. )       .f.4,5                                                            '

a. Core operating limits shall be established prior to each '

reload cycle, or prior to any remaining portion of a' reload cycle, and shall >e documented in the COLR for the following: ndividual specifications limit st be referenced here. hataddresscore\peratin

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

(continued)

BWR/4 STS 5.3-35 Rev. O, 09/28/g2 i i

Reporting Requirements Gf. I 5 $ Reporting Requirements Q .s. M D CORE OPERATING LIMITS REPORT (COLR) (continued) IdentifykeTopical rt(s) by number, title, date, nd N staff a oval docume or identify t staff Safet fgSEgr F Eva qation Rep t for a plan pecific methe gy by NRC q_letteK (nd date.

c. The core operating limits shall be determined suh that all I applicable limits (e.g., fuel thermal mechanical limits, core thermal h Systems (ECCS)ydraulic limits, nuclearlimits,limits Emergency such as SDM, Core Cooling transient g analysis limits, and accident analysis limits) of the safety analysis are met.

i d. The COLR, including any midcycle revisions or supplements, 3 shall be provided upon issuance for each reload cycle to the NRC.

N Reactor Coolant Systest(RChPRESSURE AND TkRATURE LIMITS

           -5.0.1.7

{EPON (PTLRL The RCS p sure and temperature limits, including hea and cooldown rat criticality, and hyd tatic and leak te limits, i shall be estab hed and documented in PTLR. [Theindiv al l Specifications t address the reactor ves 1 pressure and temperature limits d the heatup and cooldo ates may be , eferenced.] The ana ical methods used to det ine the p sure and tem)erature imits including the heat and cooldown rate hall be tiose ly reviewed and approve the NRC in [ Top' al Report (s)previ

                                                      , numbe       title, date, and NRC sta             pproval document, r staff safety evalu ion report for a plant spe fic
                         , methodology        NRC letter and dat              The reactor vessel press e                   -

1 and temperatur limits, including th e for heatop and cooldown l es, shall be termined so that all licable limits (e.g., hea wn limits, and inse e leak and hydrostatic limits,) coo' testin limits of th nalysis are met. Th TLR, including revision r supplements hereto, shall be pro ed upon issuanc or each re or vessel f1 cy period. M #-

      @ .0.2               seem1 =6hy                  E^s -f

[

cial Re rts may required ering insp tion, test, d mai enance tivities. These spe ' I reports a determined (continued) h BWR/4 STS 5.0-36 Rev. 0,09/28/92

RIporting Requirements 6 5., Reporting Requirements 6fk ( :.';.P -E:::f:L 5::rt, k;; tin.dB a individua'l b is for h unit and the preparationM T.2- su ttal are des ated in e Technical Sp fications. A y

                                                                                / $pecia Reports shaN be submitt in accorda                                                                                                  with 10 CFR 50.4l fwithin                                                                                   e time perio' specified fo each repo he followi                                                              Special Re rts shall be                mitted:

f

a. In the eve i an ECCS is' ctuated and in ts water in the CS in MODE 2, or 3, a ecial Report s 1 be prepa e( '

a submitted thin 90 day describing the e cuestancesK the actuation an the total a umulated actuati cycles to date. The curren value of the sage factor for h affect safety inj tion nozzle all be provided this /\ Special ort whene its value e eeds 0.70. x/_Qn)

                                                       \                                                                                                                                                   _

__ /If an)n'd'ividual emergency dies'el generat (EDG

                                                       / /                                                                                                             nences four,or more           rid failures                 a 25'                               ds, thp(e fai                               alid fai and a                            es
                                       .3                                                                         /expertene                                                        that                     ti    period s) 11 be O                                                                                                                                                                                                                                                      ,
                                                                                '/ reported wi hin                                                                                            s. Repo        on EDG f $1ures shal
                                                       /

include nforpt, tion nded~in lator e19 / { vitfron 3, Regulatory Position rh+ 4;y* Gu7,eleto,d. -/ 1 g,, j p m fe,. rum,r%.<. m f_

                                                                                             ) When a                                                                            M is required by 6 rditier                          :- !?                       ,
                                                                                                ~

LC0 3.3. 3.1 " Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the / following 14 days. The report shall outline the preplanned I alternate method of monitoring, the cause r ? the - inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. 6f.1 (. & 1%T hee;Jax b%l1elm CPAm) reiwh% Ryn A O - 6 ks BWR/4 STS 5.0-37 Rev. 0,09/28/92 I l 1

RGeordRetentio) 5.10 _0 ADMINISTRATIVE CONTROLS 5.10 cord Retention 5.10.1 T following records shall be retained fo at least 3 years:

a. Al icense Event Reports required by 10 R 50.73;

b. Records changes made to the procedures requ ed by Specificat 5.7.1.1; and

c. Records of radi tive shipments.

5.10.2 The following records shall retained for at least 5 years:

a. Records and logs of unit ope tion covering time intervals at each power level;

b. Records and logs of principal main nance activities-inspections, repair, and replacement f principal items of equipment related to nuclear safety;

c. Re rds of surveillance activities, inspect ns, and '

cali tions required by the Technical Speci ations (TS) [andth Fire Protection Program];

d. Records of s led source and fission detector leak ts and results; and

e. Records of annual sical inventory of all sealed source material of record. ,

The following records shall be 5.10.3 tained for the duration of the

unit Operating License: a. Records and drawing changes ref ting unit design ifications made to systems and uipment described in the FS; b. Records f new and irradiated fuel inven , fuel transfers, nd assembly burnup histories;; c. Records of ra tion exposure for all individua entering ,

radiation contro areas; hf,I (continued) g l BWR/4 STS 5.0-38 Rev. O,09/28/92 % __________-_A

i i 1 NUREG 1433 COMPARISON DOCUMENT - BASES 1 l l O l i i

PAM Instrumentation B 3.3.3.1 BASES LCO A Suporession Pool Water TemDerature (Continued) that, there is a group of 'ensors within a 30 ft line of sightxof eac elief va e discharge location. N 1 N Thus, s,in groups o ensors ar suffici,e N omMotoy eitch

                                                                                                        \          '

relief' valve.di s arge'loca on. Eac roup of foA sens6rs . ! 34 [3Myg includes two monitoring sors forf n rmal suppr)ession9ool' temp %raturl two sensors for- . The outputs for PAM (cardled sensors a recorde fon four ependent recorders ' the ! control com (chan'nels A apd C are reduodant to annels B i and D, respecJi9ely). A11 four of theie reco rs must be ) ! [ ofOPERABLE the relief valvetoMurnish two' channels discharge locaUions. These recordersof/PAM indicatio ' are the primary indication used by the operator during an ! accident. Therefore, the PAM Specification deals ! specifically with this portion of the instrument channels.

     , [sa7 b]         e APPLICABILITY    The-PAM instrumentation LCO is applicable in MODES I and 2..                               l These variables are relate ( to the diagnosis and preplanned O                         actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2.                    In MODES 3, 4, and 5, j

i plant conditions are such that the likelihood of an event i that would require PAM instrumentation is extremely low; j therefore, PAM instrumentation is not required to be j OPERABLE in these MODES. 1 i i i ACTIONS Note 1 has been added to the ACTIONS to. exclude the MODE ! change restriction of LCO 3.0.4. This exception allows ! entry into the applicable MODE while relying on the ACTIONS ! even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to diagnose an accident using alternative instruments and methods, and the low probability of an event requiring these instruments. Note 2 has been provided to modify the ACTIONS related to ! PAM instrumentation channels. Section 1.3, Completion ! Times, specifies that once a Condition has been entered, j subsequent $seems, subsystems, components, or variables

          'p g, oft.5)  expressed in the Condition discovered to be inoperable or j

O y .\ (continued) I BWR/4 STS B 3.3-67 Rev. O, 09/28/92 l l

PAM Instrumentation B 3.3.3.1 BASES ACTIONS not within limits, will not result in separate entry into (continued) the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable PAM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable PAM Function. A.1 When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channels (or, in the case of a Function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of.an event requiring PAM instrumentation during this interval, b .

                                                             -  p.b Li If a channel has not been restored to OPERABLE status in 30 days, this Required Action specipss initiation of action           ,

in accordance with Specification 5. 14 "rM :- =*e which requires a written report appr:d i., th; b.. :t; g' l 7= _....... .;::] to be submitted to the NRC. This report i b.\l

              /  discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions.

This action is appropriate in lieu of a shutdown requirement, since alternative actions are identified before loss of functional capability, and given the likelihood of plant conditions that would require information provided by this instrumentation. C.1 [y worQ WhenoneormoreFunctionshavetwdrequiredchannelsthat are inoperable (i.e., two channels inoperable in the same , 1 (continued) BWR/4 STS B 3.3-68 Rev. O, 09/28/92 i

PAM Instrumentation B 3.3.3.1 BASES i ACTIONS C_,_}. (continued) Function), one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of l 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an accident occur. f ondit is ied by Note that7 txqiud hydroge itor c s. tion ro ' des np appMriat quired ions fo o inopera le h ogenj r monitor channels.,

~.

When two drogen monitor ch gnels are inoper le, one h)drogenmo or c nel must by stored to O BLE status - within 72 hour he 72 ho o tion T gis sed on the I w prob li of t.be occurrenc of LOCA th ould 1),\) gener b drogen in 'unts capable o xceeding th flamm3bri tylimpi,the ngth of., time a r the even a at opey4 tor a i would be re ' red to preven drogen accumulatto rom exceeding-this limit; and the availability of the hydrogen recombiners, the Hydrogen Purge System, and the Post Accident Sampling System. - Ks.1 {9 This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. The applicable e Condition referenced in the Table is Function dependent ' Each time an inoperable channel has not met -asFRequirei

            %g Action       of Condition associated    CompletionCTime

.- 0,has es expired, ep% :._ik, and the Condition 1 is f@' '

entered for that channel and provides for transfer to the appropriate subsequent Condition. (continued) BWR/4 STS B 3.3-69 Rev. O, 09/28/92 l

PAM Instrumentation B 3.3.3.1 BASES h ' ACTIONS f_d (continued) 'S For the majority of Functions in Table 3.3.3.1-1, if any Required Action and associated Completion Time of Condition C er 0 :re not met, the plant must be brought to a MODE in which the LCO not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

                % .1                                             g> .\           y4
                                                                  '-" ' --d        ry Since alternate means of monitoring
                         =t;innat area radiation have been developed and teste        g/%

the Required Action is not to shut down the plant, but ,@ h rather to follow the directions of Specification 5.0. .a i These alternate means may be temporarily installed if the i k nonnal PAM channel cannot be restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels. AS holed a t %e beamr ma o f the SRu 3' w ~ .i SURVEILLANCE he following SRs apply to each PAM instrumentation Function REQUIREMENTS n Table 3.3.3.1-1. psrRT C SR 3.3.3.1.1 Perfonnance of the CHANNEL CHECK once every 31 days ensures _that a cross failure of instrumentation has not occurred. A CHANNEL CHECK isfa comparison of the parameter indicated on bgpth{ one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it

 ;                                                                        (continued)

BWR/4 STS B 3.3-70 Rev. O, 09/28/92

ECCS Instrumentation B 3.3.5.1 BASES-ACTIONS , G.1 and G.2 (continued) E j ion Org-<.hannels and one-er.JpotA Function-5.g ~channels) re in b l_ e . #

                                                                                                                     = - -
                                                                                                                                       .l Qt**hF           In this situation (loss of automatic initiation capability),

the 96 hour or 8 day allowance, as applicable, of Required ; {f Action G.2 is not appropriate, and all ADS valves must be e declared inoperable within I hour after discovery of loss of ! ADS i ni ti a ti on capabi l i ty . f".; r r - ...,. . . a ,,,.. . ... . . q ' rstates that Require tion G.1 is only applicable for Functions 4.c, '

                                                          ,        4.g, 5.c, 5.e, 5.T M nd                          g.                   ;

Rpquired G.1 is n p1' to F o 4.h

                              .and 5.h         i        Iso requirj e             into         s Conditiojidf a                          '

cha'nnel in these ten's is inope since.they are ! nA3 the'Ma'nual Initi_ation ionsgdarenot,assumedinany ! T 1 accident ransient analysist Thus, a total loss of manual initi n capability for 96 hours or_8 days (as Lilowed by Reaui ed Action G.2) is allowed.J - The Completion Time is intended to allow the operator time !

                             -to evaluate and repair any discovered inoperabilities. 'This                                               i Q

v Completion Time also allows for an exception to the nomal

                                " time zero" for beginning the allowed outage time " clock."

For Required Action G.1, the Completion Time only begins j upon discovery that the ADS cannot be automatically initiated due to inoperable. channels within similar ADS trip : system Functions as described in the paragraph above. The l 1 hour Completion Time from discovery of loss of initiation i capability is acceptable because it minimizes risk while : allowing time for restoration or tripping of channels. l i Because of the diversity of sensors available to provide ! initiation signals and the redundancy of the ECCS design, an : allowable out of service time of 8 days has been shown to be l acceptable (Ref. 5) to pemit restoration of any inoperable i channel to OPERABLE status if both HPCI and RCIC are ! OPERABLE (Required Action G.2). If either HPCI or RCIC is i inoperable, the time shortens to 96 hours. If the status of i HPCI or RCIC changes such that the Completion Time changes j from 8 days to 96 hours, the 96 hours begins upon discovery ; of HPCI or RCIC inoperability. However, the total time for ! an inoperable channel cannot exceed 8 days. If the status l of HPCI or RCIC changes such that the Completion Time , changes from 96 hours to 8 days, the " time zero" for ; beginning the 8 day " clock" begins upon discovery of the ; t (' """) O ! BWR/4 STS B 3.3-131 Rev. O, 09/28/92 l L

ECCS Instrumentation B 3.3.5.1 BASES h ACTIONS G.1 and G.2 (continued) inoperable channel. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events. H.1 With any Required Action and associated Completion Time not met, the associated feature (s) may be incapable of performing the intended function, and the supported feature (s) associated with inoperable untripped channels must be declared inoperable immediately. G _) SURVEILLANCE L;: c': we Certain Frequencies ar on approved REQUIREMENTS , topical reports. Iif oraci '- ee to use these Frequencies, tje e lisens must just ncies as j (.M required 4y'The staff SER for the topical report.

                                                                                  - _)

As noted in the beginning of the SRs, the SRs for each ECCS instrumentation Function are found in the SRs column of Table 3.3.5.1-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for A 1 LtL\ , p & Conditions and Reouired Actions may be delaperformance 8lof required

to 6 hours as follows: (a) for Functions 3.c f, -

    .\              and (b) for Functions other t an 3.c a3.f.4Ehd)        provided the associated Functiongmaintains        initiatio f Upon va P"g completion of the Surveillance, or expiration of the 6 hour Of g        E   I allowance, the channel must be returned to.0PERABLE status                I g 30 4p          or the applicable Condition entered and Required Actions i            0g taken.                                 iability analysis f 0gctt           (Ref. 5) assumption th:t-C h; r:This Note is based on theg:     d the averag required to perforn channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary.

(continued) BWR/4 STS B 3.3-132 Rev. O, 09/28/92

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND 1. Main Steam Line Isolation (continued) Most MSL Isolation Functions receive inputs from four channels. The outputs from these channels are combined in a one-out-of-two taken twice logic to initiate isolation of all main steam isolation valves (MSIVs). The outputs from the same channels are arranged into two two-out-of-two ogi trip systems to isolate all MSL drain valve Bush SL drain line has two isolation valves with o two-out-of-two y logic system associated with each valves (TNSEAT C D 0 (INSEAT W . The exceptions to this arrangement are the Main Steam Line 7l Flow-High Function and Area .aneM>r#ppene$e4 Temperature Functions. The Main Steam Line Flow-High Function uses

                 @       16 flow channels, four for each steam line. One channel from each steam line inputs to one of the four trip strings.

Two trip strings make up each trip system and both trip systems must trip to cause an MSL isolation. Each trip string has four inputs (one per MSL), any one of which will trip the trip string. The trip strings are arranged in a one-out-of-two taken twice logic. This is effectively a one-out-of-eight taken twice logic arrangement to initiate isolation of the MSIVs. Similarly, the 16 flow channels are O. connected into two two-out-of-two logic trip systems (effectively, two one-out-of-four twice logic), with each trip system isolating one of the two MSL drain valves en=the

ietiJ L. 1!r.;. m--(._2W SEAT O _

The Main Steam Tunnel Temperature-High Function receives input from 16 channels. The logic is arranged similar to the Main Steam Line Flow-High Function. The Turbine

/-*

Builabg Area Temperature-High Function receives input from j 64 chan gnels. The :r+.L ar; r n-- f 'r. ; one-out-of-(kv) i JNSE( LT E7 thirty-two taken twice logic trip system to isolate all U MSIVs. Similarly, the inputs are arranged in two one-out-of-sixteen twice logic trip systems, with each trip system isolating one of the two MSL drain valvesg-- d-M MSL Isolation Functions isolate the Group 1 valves.

2. Primary Containment isolation Most Primary Containment Isolation Functions receive inputs from four channels. The outputs from these channels are k (continued)

BWR/4 STS B 3.3-151 Rev. O, 09/28/92

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND 2. Primary Containment isolation (continued) arranged into two two-out-of-two logic trip systems. One trip system initiates isolation of all inboard primary containment isolation valves, while the other trip system initiates isolation of all outboard primary containment isolation valves. Each logic closes one of the two valves on each penetration, so that opeyra 'on of either lou isolates the penetration. M (Jyggg7 p The exception to this arrangement is the Drywell Radiation-High Function. This Function has two channels, whose outputs are arranged in two one-out-of-one logic trip systems. Each trip system isolates one valve per associated penetration, similar to the two-out-of-two logic described above. - Primary Containment Isolation Drywell Pressure-High and Reactor Vessel Water Level-Low, Level 3 Functions isolate the Group 2, 6, 7, 10, and 12 valves. Reactor Building and Refueling Floor Exhaust Radiation-High Functions isolate the Group 6, 10, and 12 valves. Primary Containment Jsolation Drywell Radiation-High Function isolates the*-) m containment purge and vent valves.

3. 4. Hioh Pressure Coolant iniection System Isolation and Reactor Core Isolation Coolina System isolation Most Functions that isolate HPCI and RCIC receive input from two channels, with each channel in one trip system using a one-out-of-one logic. Each of the two trip systems in each .

isolation group is connected to one of the two valves on each associated penetration. The exceptions are the HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High and Steam Supply Line Pressure-Low Functions. These Functions receive inputs from four turbine , exhaust diaphragm pressure and four steam supply pressure channels for each system. The outputs from the turbine exhaust diaphragm pressure and steam supply pressure channels are each connected to two two-out-of-two trip systems, ach trip system isolates one valve per associated penetration.

                        .UE SElLT 6 (continued)    l BWR/4 STS                      B 3.3-152                Rev. O, 09/28/92 1

6 INSERT F for oroposed BASES B 3.3.6.1 , The TIP ball valves isolation does not occur until the TIPS have been fully ! s retracted (The logic also sends a TIP retraction signal). ' t INSERT G for proposed BASES B 3.3.6.1 ! Additionally, each trip system of the Steam Line Flow-High Functions receives input from a low differential pressure channel. The low differential pressure channels are not required for OPERABILITY. t t O l t l l 2asar To 7$g3.25-1f5;2_

I l Primary Containment Isolation Instrumentation B 3.3.6.1 APPLICABLE Hich Pressure Coolant Iniection and Reactor Core Isolation SAFETY ANALYSES, Coolina Systems Isolation LCO, and APPLICABILITY 3.a. 4.a. HPCI and RCIC Steam Line Flow-Hioh (continued) Steam Line Flow-High Functions are provided to detect a break of the RCIC or HPCI steam lines and initiate closure of the steam line isolation valves of the appropriate ! system. If the steam is allowed to continue flowing out of the break, the reactor will depressurize and the core can uncover. Therefore, the isolations are initiated on high flow to prevent or minimize core damage. The isolation - action, along with the scram function of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Specific credit for these Functions is not assumed in any FSAR accident analyses since the bounding analysis is performed for large breaks such as recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding. The HPCI and RCIC Steam Line Flow-High signals are initiated from transmitters (two for HPCI and two for RCIC) O that are connected to the system steam lines. Two channels of both HPCI and RCIC Steam Line Flow-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude thejisolation function. The Allowable Values are chosen to be low enough to ensure that the trip occurs to prevent fuel damage and maintains ____ the MSLB event as'the bounding event. r y g g fL T L These Functions isolate the Group 3 and 4 valves, as 7 l appropriate. l 3.b. 4.b. HPCI and RCIC Steam Supolv Line Pressure-Low Low MSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue operation of the associated system's turbine. These isolations are for equipment protection and are not assumed in any transient or accident analysis in the FSAR. However, they also provide a diverse signal to indicate a possible ( system break. These instruments are included in Technical Specifications (TS) because of the potential for risk due to (continued) BWR/4 STS B 3.3-163 Rev. O, 09/28/92 l l - .. _ -__ _ _ _

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 3.b. 4.b. HPCI and RCIC Steam Supply Line Pressure-low SAFETY ANALYSES, (continued) LCO, and APPLICABILITY possible failure of the instruments preventing HPCI and RCIC initiationsi(%f. & ~[lere&e, #e3 me<4 Crrke, odd W O'52 l

            / [ Q. bg               Anc Po IN 5 *.b e4 The HPCI and RCIC Steam Supply Line Pressure-Low signals are initiated from transmitters (four for HPCI and four for                                     i V[ ((let.7)              RCIC) that are connected to the system steam line.                              Four channels of both HPCI and RCIC Steam Supply Line f.[p           Pressure-Low Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are selected to be high enough to prevent damage to the system's turbine. These Functions isolate the Group 3 and 4 valves, as appropriate. 3.c. 4.c. HPCI and RCIC Turbine Exhaust Diaphraam Pressure-Hioh High turbine exhaust diaphragm pressure indicates that the pressure may be too high to continue operation of the associated system's turbine. That is, one of two exhaust diaphragms has ruptured and pressure is reaching turbine casing pressure limits. These isolations are for equipment protection and are not assumed in any transient or accident analysis in the FSAR. These instruments are included in the p TS because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations, pg. 3}m %Ie & c +k<q meef CrQc u %f % u Rc, ,5

g. l, h to% Sbed, .

              *g            - '

4'[ Thesignals HPCI andareRCIC Turbine initiated from Exhaust transmitters Diaphragm (four forPressure-High HPCI and

               ")        (J four for RCIC) that are connected to the area between the rupture diaphragms on each system's turbine exhaust line.

Four channels of both HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument l failure can preclude the isolation function. 1 The Allowable Values are Ai.gh enough to prevent damage to the system's turbine. Low a (continued) l BWR/4 STS B 3.3-164 Rev. O, 09/28/92

..- . . . - . . - - . - . . . - . . - . . _ - . . . . ~ - . . ~ . . . .

           ~CO M N-INSERT L for orooosed BASES B 3.3.6.1                                         b        _

The Allowable Values correspond to s 215 inches water column for HPCI and s ! 190 inches water column for RCIC, which are the parameters monitored on i control room instruments. l i i l l I i i i

r r r O l t t h t i I l I e i h i O l soseLT To l 3 3 ,3 - 16,3 . i

INSERT 6 3.3.8.1 Backaround Section oaae 83.3-219) O Each 4.16 kV emergency bus has a dedicat Ved annunciator fed by two ow voltage : relays and their associated time delays. The logic for the annunciation ' function is arranged in a two-out-of-two configuration. ,l 3.3.8.1 Backaround Section JUnit 2)

                                                                                             \  l Each 4.16 kV emergency bus has a dedicated low voltage annunciator fed by two       i relays and their associated time delays. The logic for the annunciation function is arranged in a one-out-of-two configuration.                             '

I b I b O ! I L e O , LtMIT _1 A@ bOIT 2_ a,

t LOP Instrumentation i B-3.3.8.1 i

                                                                       ~                 ~

BASES f-f^ j APPLICABLE / The specific Applicable l SAFETY ANALYSES, Applicability discussions are listed LCO, and , Function basis. APPLICABILITY (continued) _. r 1. 4.16 kV Emeroency Bus Undervoltaae (Loss of Voltaae) offsite power may be completely lost to the respective proper operation of the applicable equipment. Therefore, power to DG power when the voltage on the bus drops below the Loss of Voltage Function Allowable Values (loss of voltage with a short time delay). This ensures that adequate power will be available to the required equipment. The Bus Undervoltage Allowable Values are low enough to to ensure that power is available to the required equipment. voltages, but short enough to ensure that power is_ available O r Two channels of 4.16 kV Emergency Bus-Undervoltage (Loss of required to be OPERABLE when the associated DG is required can preclude the DG function. (Two channels input to each Sources-Operating,",and 3.8.2, "AC Sources-Shutdown," for Applicability Bases ~for the DGs. " y ! < Safety Analyses, LCO, and - below on a Function by ; ! ' .--- l LAsIofvoltageons4.16kVemergencybusindicatesthat ' emergency bus and is unable to supply sufficient power for ' the power supply to the bus is transferred from offsite ( ; ; ! : i prevent inadvertent power supply transfer, but high enough . l The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to nomal > to the required equipment. l Voltage) Function per associated emergency bus are only ; l to be OPERABLE to ensure that no single instrument failure : I of the three DGs.) Refer to LC0 3.8.1, "AC ; ' ._

2. 4.16 kV Eneroency Bus Undervoltace (Dearaded Voltaae)

A reduced voltage condition on a 4.16 kV emergency bus !- indicates that, while offsite power may not be completely ! lost to the respective emergency bus, available power may be : insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function. Therefore, power supply to the bus is transferred from offsite power to onsite DG power when the voltage on the bus' drops below the Degraded Voltage Function Allowable Values 4 g ( _ t, _ d) BWR/4 STS B 3.3-221 Rev. O, 09/28/92 1

LOP Instrumentation B 3.3.8.1 BASES APPLICABLE / 2. 4.16 kV Emeraency Bus Undervoltace (Deoraded Voltaae) SAFETY ANALYSES, (continued) LCO, and APPLICABILITY (degraded voltage with a time delay). This ensures that . adequate power will be available to the required equipment. _ f, - /

                              -       The Bus Undervoltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough b
           /
                        ]             to ensure that sufficient power is available to the r:;;ir:d
                                      ;;.ie.;.t. The Time Delay Aljogble Values are long enou Jh MMD
 .g                                  G y ovide ~ time ford he offsite p              supp4f   to recherto.
                        ~

a he uired

   /b Oc        .N4e pD --
                                                                                   ~

5 yt y 4= v4 any m ..\ Two channels of 4.16 kV _ Emergency Bus Undervoltage (Degraded i

    'TM         :,,4,5 4L,_         ) Voltage) Function per associated bus are only requi. red to be tede.A L 4 t,.t st. .t            OPERABLE when the associated DG is required to be OPERABLE LA A.44. ,,s                al:   to ensure that no single instrument failure can preclude the cL . 6pg&                         DG function. (Two channels input to each of the three av3llM)iay,(+Le                   emergency buses and DGs.)     Refer to LCO 3.8.1 and LCO 3.8.2 L.f4ac . ., r.9,tm                  for Applicability Bases for the DGs.

_ @EPTO, 3, ( MELT 5 ACTIONS A Note has been provided to modify the ACTIONS related to

LOP instrumentation channels. Section 1.3, Completion

                    ..\               Times. specifies that once a Condition has been entered, subsequent hin:, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or dWishg                   not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable LOP instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable LOP instrumentation channel.                                        ~

A.1

                                                                     '                               P.70
                                                                         /              .6 ar 2 WithoneormorechannelsoffFunctioninoperable,the                            l Functionits"notx apaD4e of periortpng.pfe in3cnded Jdnctipn_

Therefore, only 1 hour is allowed to restore-the inoperable P. u, Aes mo+ ~6Ja '.. '.4M;. = pull-L, b A_ ass.c.41 e.--we y N tcontinued) BWR/4 STS B 3.3-222 Rev. O, 09/28/92

INSERT 5 L3.8.1 Acolicable Safety Analysis Section_ (Unit 2) m (page B 3.3-222) l s i

3. 4.16 kV Emeroency Bus Undervoltaae (Anticipatory Alarms) i A reduced voltage condition on a 4.16 kV emergency indicated that, while offsite power is adequate for normal operating conditlons, available power may be marginal for some equipment required for LOCA cunditions. Therefore, the anticipatory alarms actuate when the 4.16 kV bus voltages approach the minimum required voltage for normal; i.e., non-LOCA, conditions. This ensures that manual actions will be initiated to restore the bus voltages or to initiate a plant shutdown. i One channel of the 4.16 kV emergency bus undervoltage (Anticipatory Alarm)- ,

function per the associated bus is only required to be OPERABLE when the asso:iated DG is required to be OPERABLE. (Two channels input to each of the three emergency buses.) 3.3.8.1 Aeolicable Safety Analysis Section.(Unit 1) , (page B 3.3-222) Q b

3. 4.16 kV Emeraency Bus Undervoltaae (Anticipatory Alarms)

A reduced voltage condition on a 4.16 kV emergency indicated that, while , offsite power is adequate for normal operating conditions, available power may be marginal for some equipment required for LOCA conditions. Therefore, the ' anticipatory alarms actuate when the 4.16 kV bus voltages approach the minimum ' required voltage for normal; i.e., non-LOCA, conditions. This ensures that manual actions will be initiated to restore the bus voltages or to initiate a - plant shutdown. Two channels of the 4.16 kV emergency bus undervoltage (Anticipatory Alarm) l function per the associated bus are only required to be OPERABLE when the associated DG is required to be OPERABLE. (Two channels input to each of the I three emergency buses.) l I O E UNIT 1 AND UNIT 2 REVISION [

                                                                                                                                . LOP Instrumentation B 3.3.8.1 BASES ACTIONS                                  A.1     (continued) channel to OPERABLE status.[If the inope ble ch nel 5nm pe restorea to 0F                         BLE st us withi the a owabl]e M                         o    of servic'e time, the annel m t be pla d in e tri ed conditivn per Requi ed Actio A.I. P1 ing t e inop able chann' 1 in trip uld cons vatively ompen te for th inoperabi 'ty, resto                         capabili            to acco odat a
            % p%,.             " y 4                  IfEngle

lure (wit in the LOP instrumen tion), a allo

             @M aW                                  j       op ation            continu J plac the ch nel in t ip (e.g.,

Alterna ly, if i is not de ' red s in the ase where i 8'h # 4,ycl placin the ch nel in ip would sult in DG pis aMk P Me s Require initiat n),Co ition B must be ent red and

                                           # ll          g Action        taken. f--                                                                                      t N^                                       The Completion Time is intended to allow the operator time                                                    4
        ' ,         /M-                                     to' evaluate and repair any discovered inoperabilities. The
                                                   's 1 hour Completion Time is acceptable because it minimizes l
  /                                                         risk while allosiing time for restoration or tripping of                                           l          ,
/           - - _ -

channels. / 3 td5E_BT_d- > ' pno i

                                            ^               If any Required Action and associated Completion Time are not met, the associated FunctionjiyhopfapableJK) gl
         %-                                     ,__.{mTorminer twintevfded#uncttenA Therefore, tne associated                                                    ,

M s) is declared inoperable inanediately. This requires

M s M " . 3 *.'"3 ' entry into applicable Conditions and Required Actions of bhh ePk#7 LCO 3.8.1 and LCO 3.8.2, which provide appropriate actions

          & 4k mac d=4                                      for the inoperable DG(s).

te-.ge-ey bus u , - - - -

P.w SURVEILLANCE / As noted at the beginning of the SRs, the SRs for each LOP \ REQUIREMENTS instrumentation Function are located in the SRs column of / Table 3.3.8.1-1. "

                                                      ~ The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated                                                -

Conditions and Required Actions may be delayed for up to M hours provided the associated Function maintains p .72 , M initiation capabilitw,4Unon concletion of the Surveillance,

                                                          ' or7xpiration of the X ho r allowance, the channel must be]                                                     ,

e,8 J F- A -s t J 1 m. h I p:~:4: k i :4:3+.J d : - e 6m&b:t h pb.vla. m-c% A A t,ch& 2. .e4 a, sh.A . f 6 3 coLs..

                                                                                                                             ~4sd yo:4e ;A n. bw m.                bI w-      . A. a w. L,c cL y Q d .a, ra ch s :e.                                         . g o cces m O                                                                                                                                                continued)
                  . . .           .~ w a ,

BWR 4 STSN B 3.3-223 bRev. O, 09/28/92 g or EdtrC I N a nm). d 2) osui 7 Ctanu_(cg-h0A C6ph h A > 1

(. g' [ 0#d N buct)Ovl 3 b k G Gn1 e sm] B 3.3.8.1 LOP Instrumentation ] f W N CtMnWilCAfdof Cl6h5 'to %e bg \p@ M] BASES (k CLA-N-CL(Y0be54 (onf(Cm t'ncj -Se, g g% q g N Q M {X D o\e- C h t9Hir4 6rd SDtf rdinc ) SURVEILLANCE returned to OPERABLE M atus of the applicaDie Etndition REQUIREMENTS nd Re quired Actions taken. ' (continued) enteredg \ - SR 3.3.8.1. (ulkrc-Of dAtuacdch Performance of he CHANNEL CHECK once eyery 12 hours ensures A thatagrossfaijureofinstrumentationAhasnotoccurred. paris f the para er indicat d on CHA EL CHECK isf ne hannei to imila arameter on er channe . It 's d on th ssump ' n that inst nt channel onito ' g he sa arame should rea pproximatel e sa alue. Sig icant iations bet en the instr nt cha els coulc (/J an ind tion of excdsive instru t drift i channels )or something even more serfous.( A CHANNEL CHECK oneofthe)

                      ~ will detect cross channel failured thus, it is key to verifying the instrumentation continues to operate properly Di i    go etMY             between each CHANNEL CALIBRATION.

f /R70) [ Md Agpeliient teri re d rmin by t pljt'staf a ~ayed Afi a c inati of t chann ins menteuncert intifs

                         'nci ing in cation and re abili .J If a channel is outside the match criteria, it may be an indication that the instrument has drifted outside its limit.                                             l I 2L The Frequency is based upon operating experience that demonstrates channel failure is rare. Thus, perfonnance of the CHANNEL CHECK ensures that undetected outright channe)oQ failure is limited to 12 hours. The CHANNEL CHECK supplements less formal, but more frequent, checks of                  (ggg    .

channels during nomal operational use of the displays

                      ' associated with channels required by the LCO.

T SR 3.3.8.1.2 -

                       'A CHANNEL FUNCTIONAL TEST is performed on each required hI           channel to ensure that the entire channel will perform the intended function.o (b b        The Frequency of 3          ays is based on operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given function in any 31 day interval is a rare event.

(continued) BWR/4 STS B 3.3-224 Rev. O, 09/28/92

INSERT C Unit 2) CTIONS (continued) l Each 4.16 k bus has a dedicated annunciator fed by two relays and their associated time delays in a one-out-of-two logic configuration. Only one O relay and its associated time delays is required to be OPERABLE. Therefore, the loss of the required relay or time delay renders Function incapable of performing the intended function. Since the intended function is to alert personnel to a lowering voltage condition and the voltage reading is available for each bus on the control room front panels, the Required Action is verification of the voltage to be above the annunciator setpoint (nominal) hourly. Unit 1) ACTIONS Each 4.16 kV bus has a dedicated annunciator fed by two relays and their associated time delays in a two-out-of-two logic configuration. Both relays and their associated time delays are required to be OPERABLE. Therefore, either required relay or time delay renders Function incapable of performing 4 the intended function. Since the intended function is to alert personnel to a lowering voltage condition and the voltage reading is available for each bus on the control room front panels, the Required Action is verification of the voltage to be above the annunciator setpoint (nominal) hourly. G U O E UNIT 1 AND UNIT 2 REVISION [ I

RHR Shutdown Cooling System-Cold Shutdown B 3 4.hg 4> : BASES ACTIONS B.1 and B.2 (I (continued) With no RHR shutdown cooling subsystem and no recirculation , pump in operation, except as permitted by the LCO Not , an until RHR or recirculation pump operation is re-established, ' an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature. The I hour Completion Time is based on the coolant circulation function ; and is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours thereafter. This will provide assurance of continued d' temperature monitoring ca ity. I'1 During the period when the reactor coolant is being C MP J , gti an alternate thod (other than by the ]

                              ,circulatedby/hutdownfooli required RHR                       yste@ , the reactor coolant ILgyd pr.cssge I"           tenneraturermust be periodical y monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.

O SURVEILLANCE SR 3.4.f. REQUIREMENTS ' This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is , determined by the flow rate necessary to provide sufficient l decay heat removal capability. The Frequency of 12 hours is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.

                                                         'IhJ REFERENCES             -Nene.--               .

l. 4-{lNSEET 64t*

O ; BWR/4 STS B 3.4-45 Rev. O, 09/28/92 ! i

RCS P/T Linits B 3.4. N q B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4..W RCS Pressure and Temperature (P/T) Limits l BASES

BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

p. 30 Eko MD

{TMs SpecMach;nc = contains P/T limit curves for heatup,kooldo ,h inservice leakage and hydrostatic testing, and -=t= ' a the

                                                               + - temperature. The hbh-@ maximum e144 curve criticality.

rate of change provides of reactor limits for both cotand

                                                                  'up                     h P. it Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). 'The vessel is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel. 10 CFR 50, Appendix G (Ref.1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section III, Appendix G (Ref. 2). The actual shift in the RTwor of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref.4). The operating P/T limit curves will be adjusted, (continued) < BWR/4 STS B 3.4-46 Rev. O, 09/28/92

=--

RCS P/T Linits B 3.4.M. BASES BACKGROUND as necessary, based on the evaluation findings and the (continued) recemendations of Reference 5. The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the ' span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions. The heatup curve represents a different set of restrictions - than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls. - The criticality limits include the Reference 1 requirement that they be at least 40'F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic O testing. The consequence of violating the LCO limits is that the RCS ' has been operated under conditions that can. result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event ' these limits are exceeded, an evaluation must be perfomed to detemine the effect on the structural integrity of the RCPB components. ASME Code, Section XI, Appendix f. (Ref. 6), provides a recomended methodology for evaluating an operating event that causes an excursion outside the limits. APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature ; eate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a ' condition that is unanalyzedg :%

                     - ...--- -u   . . . . . . . . . . . . . , ...

fx x x ' e tSince

                                                                     "/T li;;;its. 9 Hethe%rP/T th limits are not derived from any DBA, there are no acceptance                      }

(continued) BWR/4 STS 8 3*4~47 Rev. O, 09/28/92 g_ s .,,,- ~ _. , m ,, 6

                                                                                   .~. , ,        .w       A

l i RCS P/T Licits B 3.4.t6-BASES APPLICABLE limits related to the P/T limits. Rather, the P/T. limits SAFETY ANALYSES are acceptance limits themselves since they preclude (continued) operation in an unanalyzed condition. RCS P/T limits satisfy Criterion 2 of the NRC Policy b Statemen{(Reir. CI h) k: ' LCO The elements of th' LCO are: ' figures 3 A 4-14~1 349-2 + , ed ~ i

a. RCS pressure 'temperatu e[and heatu) or cooldown rate.s , y i are within the limits sped iime ;.-J: = PTL"2 h ig Rc5 ht4 uotsown, an inss<viet iah and NMeHuHc. 4cshg;

b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel l V) coolant ismithin th: l'-it Of th; PTLR during i 5 lU recirculation pump1 startup/ e4.id, === L % ^ - Ti l f g

r ;@ ../ .r=np s .._ . . , _ _ .. . ,% 3 il p. 30' c Pa he j y in temperature the respective difference between recirculation looptheand reactor in thecoolant puimany,q ! o r vessel meet:; th: 'irit :' u.s ricR during* pump

                                                         " T            ^- = ^ : ^ ^^:  ~ ': v "q
                 -        ~stiFtup g i.

ad

                                      .. ".... y. " *y'i
                                                             . eg . ;

d. RCS pressure and temperature are within the P30 criticalitydlimits l @c specified

                                                  -Q.pr, u A-3    inI'he ."TL"/

erme*4iff;Peter*^4 fo acMeq. g

                                                                                                      ~7'   vif

e. The reactor vessel flange and the head flange temperatures are within the li;it: # +" "L .when d ut i i@mt, min 3 Wp. reactor vessel head bolting studs r; t:n:f red.

lD These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide l margin to nonductile failure.

                                                                              /s          '3 The rate of change of temperature limits                    trob t e themal                 ,

gradient through the vessel wall and used as inputp for l calculating the heatup, cooldown, and inservice leakage and i hydrostatic testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves. Violation of the limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses (continued) BWR/4 STS B 3.4-48 Rev. O, 09/28/92

RCS P/T Limits B 3.4. W BASES ACTIONS C.1 and C.2 (continued) Operation outside the P/T limits in other than MODES 1, 2, , and 3 (including defueled conditions) must be corrected so I that the RCPB is returned to a condition that has been i verified by stress analyses. The Required Action must be , initiated without delay and continued until the limits are i restored. i i Besides restoring the P/T limit parameters to within limits, ! an evaluation is required to detemine if RCS operation is allowed. This evaluation must verify that the RCPB /f integrity is acceptable and must be completed before 2.12 l approaching criticality or heating up to > 'F. evera ; methods may be used, including comparison with pre-analyzed : transients, new analyses, or inspection of the components. i pb ASME Code, Section XI, Appendix E (Ref. 6), may be used to l support the evaluation; however, its use is restricted to I evaluation of the beltline. ! JMSGM6SI - . SR 3.4.MT.1 9 h' P'3 SURVEILLANCE REQUIREMENTS Verification that operation is within limits is h- I required every 30 minutes when RCS >ressure and temperature , conditions are undergoing planned cianges. This Frequency i is considered reasonable in view of the control roon j indication available to monitor RCS status. Also, since : temperature rate of change limits are specified in hourly l increments, 30 minutes pemits a reasonable time for i assessment and correction of minor deviations. i Surveillance for heatup, cooldown, or inservice leakage and I hydrostatic testing may be discontinued when the criteria ! given in the relevant plant procedure for ending the j activity are satisfied. ; This SR has been modified with a Note that requires this -l Surveillance to be performed only during system heatup and i cooldown operations and inservice leakage and hydrostatic { testing, i MCS m ' (continued) BWR/4 STS B 3.4-51 Rev. O, 09/28/92 l

RCS P/T Limits , B 3.4.19- I BASES m 9 Va b SURVEILLANCE SR 3 . 4. Mr. REQUIREMENTS (continued) A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical. Perfoming the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal. SR 3.4. .3 and SR 3.4.16.4 p,3o Differential temperatures within the applicable ensure that thermal stresses resulting from the startup of imits ld an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. are satisfied. Perfoming the Surveillance within minut@esbefore starting the idle recirculation pump provides adequate ! assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start. An acceptable means of demonstrating compliance with the g g g temperature differential requirement in SR 3.4.Mr.4 is to compare the temperatures of the operating recirculation loop a the idle 1000. (-ga w ww.g. 4 han f,D 4.7.3 %s S been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 49

                                                                - mie 7:::tn L .. dv-s pis..... :.O g.p In MODE 5, the l

overall stress on limiting components is lower. Therefore, AT limits are not required. SR 3.4. .5. SR 3.4. .6. and SR 3.4. .7 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits l (continued) l BWR/4 STS B 3.4-52 Rev. O, 09/28/92 l l

RCS P/T Lit::its B 3.4.W BASES SURVEILLANCE SR 3.4. .5. SR 3.4. .6. and SR 3.4. .7 (continued) REQUIREMENTS during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits. The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel o head bolting studs to ensure that once the head is tensioned J f .. u the limits are satisfied. When in MODE 4 with RCS A B4 Up @ temperature s TA'F, 30 minute checks of the flange temperatures are required because of the reduced margin to Q the limits. When in MODE 4 with RCS temperature 5 monitoring of the flange temperature is required ev%ery s o(, vs I 12 hours to ensure the temperature is within the limits ga w d specified in the PTLR. The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour p [.4 e-Frequency is reasonable based on the rate of temperature change possible at these temperatures. (JNSETLT &53ah REFERENCES 1. 10 CFR 50, Appendix G. W.

2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G. g w5 g,enif ramq 7,

l

3. ASTM E 185-82,[ July 1982. Tes4s Ter QM -Wokr Co*Id Nul*W %

y % gggy .

4. 10 CFR 50, Appendix H.

5. Regulatory Guide 1.99, Revision 2, May 1988.

5. ASME, Boiler and Pressure Vessel Code, Section XI, ._..

l Appendix E. ,

                                                .  . ~ . .... ., ..__ _ . . . . . .                                              (      1 p.7 FSAR, Section                      5.1.26f}. .. 6mr t>,,
                                                                                              ,       4        g                 [      l a

E'r1%IN_rF#TBf3( lg l O BWR/4 STS B 3.4-53 Rev. O, 09/28/92 L_________-_______________-______-____________________________________ .-

Reactor Steam Dome Pressure B 3.4..W m to j# b ' B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4. K Reactor Steam Dome Pressure lo BASES ._ P.b BACKGROUND The reactor steam dome pressure is an assumed initial condition of desion basis accidents and transients + sheepn assumed value in the determination of compliance

                    %  'with reactor pressure vessel overpressure protsetion

{J criteriaj adis also f APPLICABLE Thereactorsteamdomepressureofs[020 s an SAFETY ANALYSES initial condition of the vessel overpressure protection analysis of Reference 1. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety / relief valves, during the limiting pressurization transient. The detemination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved. Reference 2 also assumes an initial reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (see Bases for LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"). Reactor steam dome pressure satisfies the requirements of Criterion 2 of the NRC Policy Statemen -, M 31 ~ Qarguswt paie<b.amolysW Q LCO The specified reactor steam dome pressure limit of 5 1020f psig ensures, the plant is operated within the assumptions of the tr&:h..; m,;.ly:::. Operation above the limit may result in a ers..: M-t response more severe than analyzed. APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these (continued) BWR/4 STS B 3.4-54 Rev. O, 09/28/92 l

INSERT bNA SR 3.4.9.5 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting

       =tuds.      SR 3.4.9.6 is modified by a Note that requires the ui- g  Surveillance to be initiated 30 minutes after RCS temperature           l 5   *F in Mode 4. SR 3.4.9.7 is modified by a Note that requires        17 4t- 100 the Surveillance to be initiated 12 hours after RCS temperature
       $    'F in Mode 4. The Notes contained in these SRs are necessary   ls(A

,,, go specify when the reactor vessel flange and head flange [f\ temperatures are required to be verified to be within the limits d1*" go specified,@ .. r- lj 0 i O g met,T TD 6 3.4-53

INSERT B53 b Unit I _y 8. George W. Riverbark (NRC) Letter to J. T. Beckham, Jr. (GPC), Amendment 126 to the Operating License, dated June 22, 1986. Unit 2 y 8. Kahtan N. Jabbour (NRC) Letter to W. G. Hairston', III (GPC), Amendment 118 to the Operating License, dated 1 January 10, 1992. l

9. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993. ,

3 i O . l I i l O . a= To

 #a tta 92                      8 3.4-53 l

J

   -Jc 0N0             &IU                                                                                                            3  f B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3       Standby Gas Treatment (SGT) System BASES BACKGROUND              The SGT System is required by 10 CFR 50, Appendix A. GDC 41,
                                 " Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials th leak from the primary containment into the (secondar containment following a Design Basis Accident (DBA)                                                 are yk g.1 filtered and adsorbed prior to exhausting to the environment.

g gl pd The SGT Syste onsists of two fully redundant subsystems, each with its own set of t;:t.M. dampers, charcoal filter W'g g train, and controlsy Each charcoal filter train consists of (components listed in p y g 'g order of the direction of the air flow):

a. A desister or moisture separator; (80

b. An electric heater;

c. A prefilter;

d. A high efficiency particulate air (HEPA) filter; hD e. A charcoal adsorberp(IB5E/27 A) f

f. A second HEPA filter; and '

oriol vwe g :s.tri'_;:1 fan. The sizing of the SGT Syst equipment and compone s is : based on the results of an infiltration analysjs, as well as an exfiltration analysis of the $secondaryktontainment. 3 . 27 The internal pressure of the SGT Syste Q oundary region is , maintained at a negative pressure of 10.25). inches water r gauge when the system is in operation, which represents the internal pressure required to ensure zero exfiltration of PZ air from the bui d

                               ...inyaten...ylngwhenjxgsedtoag0)euphwind, m-
                                                                       .. e.c . . . . . . . . . . . . . . , .

The desister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity l 4 (continued) BWR/4 STS B 3.6-109 Rev. O, 09/28/92 UNii/

    -       -_           -                                 . _= _- .                        _ _ _

l l I Il"IY l Y63[dd SGT System B 3.6.4.3 BASES ( Qpp ( aa d 3) 1 BACKGROUND of the airstream to less than 470}% (Re - 21. The prefilter (continued) removes large particulate matter, while he HEPA filter removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorber remove $ gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhau from the charcoal g adsorber. g g ,,. g ! 04fE ! " The SGT Systes6(automatically starty and operate $ in response k-) pt.2 to actuation signal indicative of conditions or an accident that could re peration of the system. Following initiation, harcoal filter train fans start. Upon verification that subsystems are operating, the redundant 3 subsystem is n lly shut down. [ @ jarf) etc9mcd) APPLICABLE The design basis for th)eGT Syst s to mitigate the SAFETY ANALYSES consequences of a loss of coolant accident and fuel handling accidents (Ref> 2). For all events analyzed, the SGT System is shown to be a@ omatically initiated to reduce, via filtration and ads tion, the radioactive material released to the environment. g3 ) The SGT System satisfies Criterion 3 of the NRC Policy , Statement j

                                                                           ' ojo LCO G) aid))LP tY Followin a DBA a Tinimum of                     T subsyst is required p, s     to           n th secondaryd containment at a negative QP.
                             - pressure w          respect to the environment and to process fg gaseous releases. Meeting the LC0 requirements for OPERABLE subsyst         ensures operation of at least                   SGT d566                subsys in the vent of a single active failure. g                       F 17 p.
                                                 ~

SE(LT C }

                                                                               'oCA (L.@)

APPLICABILITY In MODES 1, 2, and 3, a 884 could ea to a fission product l release to primary containment that leaks to secondary l containment. Therefore, SGT System OPERA 8ILITY .w required ! during these MODES. Es t / ud 2. g In MODES 4 and 5, the probability and consequences of d eve esse 6s are reduced due to the pressure and temperature

o. l M limitations in these MODES. Therefore, maintaining the SGT (continued)

BWR/4 STS B 3.6-110 Rev. O, 09/28/92 (JNITl

g (lvd / /e/ifod SGT System B 3.6.4.3

         ,N BASES ACTIONS        ).1                .3   (continued) gh-     operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a                       J sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3. 3 REQUIREMENTS Operating each{SGT k y J-su system for e {10] continuous hours ensures that {k'i a:y:^= are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation ith the heaters on 'eutc; ti: h::te r -- -'-- + - e;: :tMKfor afl0)f continuous hours every 31 days eliminates afoisture on the adsorbers and HEPA filters. Th 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system. SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter '- Testing Program (VFTP). The SET Sy;t ; filter te t: tr: 3

                   - c: rd:::: with P.e;"l:t:ry 0;id: 1.52 ("d. 4. The VFTP

['@ includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). . Specific test frequencies and additional infomation are discussed in detail in the i VFTP. p.27) i SR 3.6.4.3.3 This SR verifies that eac GT subsystem starts on receiot of an actual or simulated initiation signal./While this 7 ; Surveillance can be perfonned with the reactor at power,

operating experience has shown that these components u _ 11y ,

pass the Surveillance when >erfonned at the f18Emonth* a p, % Frequency _.TThe LOGIC-sYsTE0 FUNCTIONAL It.sT in SR 3.3.6.2,6=- (overlaps this SR to provide complete testing of the safety f (continue BWR/4 STS B 3.6-113 Rev. O, 09/28/92 UNIT / 1

l I d S>< f ) - t/65IM SGT Systea B 3.6.4.3 : O[,2J BASES SURVEILLANCE SR 3.6.4.3.3 (continued) REQUIREMENTS __ to Wuio Therefore, the Frequency was found to be r*dggd acceptable from a reliability standpoint. It fe-g% _ SR 3.6.4.3.4 # , This SR veri at the filter cooler bypass r can be opened and the fan s . This ensu the ventilation mode of SGT Sys on is available, p)y While this Surveillanc power, operatin e per rience has shown tha with the reactor at se components usually e Surveillance when performed at [1 nth Frequency, which is based on the refueling le. erefore, the Frequency was found to be acceptable from a

            /      reliability standpoint.                                                  N REFERENCES         1.      10 CFR 50, Appendix A GOC 41.

2. FSAR,Section((0.2.3]. g

().M 3. =;mn. , .

                                        .:n :.::, nn . [2] .

7

               -- 3, uni b 2 FSA E, Se e hon. C.2. 3.

l i

                                                                                              \
                                                                                              \

O BWR/4 STS B 3.6-114 Rev. O,09/28/92 UNIT /

g jp)b 55W -O sraRa3 i SGT System ().27 GI(CIw T **5 B364YI Sfeerke fy War W eb B 3.6 CONTAINMENT SYSTEMS B 3.6.4./ Standby Gas Treatment (SGT) System-O para % 9 7 BASES BACKGROUND The SGT System is required by 10 CFR 50, Appendix A. GDC 41,

                                                                        " Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into theXsecondar           p, t containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the envi ronment.

M (/M l*"t (kdhe GT Systemj onsists of two fully redundant subsystems, each with its own set of het=:rk, dampers, charcoal filter train, and controls., M Each charcoal filter train consists of (components listed in l order of the direction of the air flow): f.@ a. A demister or moisture separator;

b. An electric heater; A

c. A prefilter; i i

d. A high efficiency particulate air (HEPA) H 1 ter- l pao p(ocharcoal adsorbe q u-14I.rs 9.s m .4 w D e

c h.l 6 k h umf1 '

f. A second HEPA filter; and M A

g. A centrifugal fan. hM The sizing of the SGT System 3 equipment and components is based on the results of an infiltration analysis, as well as ;

4 ld_ an exfiltration analysis of the secondaryk containments p. t. W'I 7 j The internal pressure of the SG Systensboundary region is maintainedatanegativepressureof40.25iincheswa gauge when the system is in operation, which represents the internal pressure required to ensure zero exfiltrati_on of airfromthebuildingwhenexposedtoaX10fsphwind,

                                                                    -binin; et en en;!: ;f [4';]" t: th: 591df=;_

b' The desister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity A j (continued) BWR/4 STS B 3.6-109 O, 09/28/92 Rev. lJN t T I.

.fr{} y d [ W S io d trwfiu O SGT Systea ; \f. 2.7 A If d a g e # B3.6.4.% l Spec klf, "N*'A ' BASES BACKGROUND of the airstream to less than)[70 % efs. 2f. The filter (continued) removes large particulate matter, while the HEPA filter removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorbetremoveX gaseous Ay6 elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal 1 adsorber.

           "'{ lQ UO     Th SGT System 5 automatically startr, and operateX in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following initiation, both charcoal filter train fans start. Upon verification that-b+th-subsystems are operating, the redundant subsystem is normally shut down.                 -

(ecmred I%9 ift0 _ h'S gec,a The design basis for the'SGTisSystey,w APPLICABLE to mitigate- the SAFETY ANALYSES consequences of a loss of coolant accident and feel hand!%g accWnt; (Refs. 2). For sit eventt enely:d, the SGT Systesit5 g shown to be automaticall initiated to reduce, via filtration and adsorption, the radioactive material released J to the environment. los jgsErLT C The SGT System satisfies Criterionc 3 of the NR Statemen . I'$ j _ s g- 3,m LCO U"g lJ FollowingaSBA,[aminimumofontSGTsubsystem5Nequir to maintain the,) seconda containmentfat a negative (A~'4 y pressure with respect to the environment and to process gaseous releases Meeting the LCO requirements for two4u e OPERABLE subsystems ensures operation of at least SGT subsystessin the event of a single active failure. pserf h u"*

                                                            ~           ,          ~

g A 46.W Q1o,ff -J S d V APPLICABILITY In MODES 1, 2, and 3, a-9BA could lead to a \ fission product release to primary containment that leaks toisecondary containmentr. during Therefore, SGT Systems 0PERABILITY p required these MODES. g fy gg m fii uts 4 ano o, the prv' ability and consequei a vi inese 9g i event are reduced due to limitat in these MODES. pressure and tempe efore. maintaint ure the SGT-6 (continued) BWR/4 STS B 3.6-110 Rev. O, 09/28/92 ()t)l T ')

                                                                                                                                                                              #*S Y pk T M4p?                                                                                                                                                               SGT System 4ff C       eM                                                                                                                                                      B 3.6.4.1 7

5pre ;-Rce lly d"hed (] v BASES ACTIONS D. D.2. and 0.3 (continued')s operat ns. Therefore, in eith ase, inability to suspend movemen of irradiated fuel assemb s would not be a sufficien eason to require a react 7 re ed tina I *9 %:& P. :u, SURVEILLANCE SR 3.6.4.1.1 pt REQUIREMENTS k) Operating each SGT subsystem for a {f 7 continuous hours 10 J~ a -- are OPERABLE and that all ensures that [h M] associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation g'with the heaters

                             --;:::tr:}}(for           on 'e.tenti; ty10)(continuous                                     hours                   heater                              every;;;'n;  t: nintiM{

31 days Opt eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system. SR 3.6.4. .2 This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing 4=- Pro : l::: gram

(VFTP).

rfth Re;Th: tery SCTtid:Syst= 1.";2 filter (h f. int:3).

tre The VFTP

                  / .Ga includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). . Specific test frequencies and additional information are discussed in detail in the VFTP.

1 SR 3.6.4.1.3 , This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal.JTihile this % Surveillance can be performed with the reactor at power, I operating experience has shown that these components usully ) pass the Surveillance when serformed at the 418 F oontn @ 3 c - Frequenev- iThe L 65L ST31FA FUNCTIONAL TEST in SR 3.3.6.2 fverlaps'this SR to provide complete testing of the safe h h$h a (continued) r ar BWR/4 STS B 3.6-113 Rev. O, 09/28/92 UN17 :

1 ,fg yest 2. VUSION -0 kg Al(cLgo 3.6.4. Speer%l{y AbN*0 ? ' BASES h SURVEILLANCE SR 3.6.4.3.3 (continued) REQUIREMENTS Mfunctiond Therefore, the Frequency was found to be

                #       acceptable from a reliability standpoint.

hCl \SR 3.6.4.3.4

s SR verifies that the filter cool bypass damper can be operied and the fan started. This ensur that the ventilation mode of SGT System operation s available. While this Surveillance can be performed w sth the reactor at power, operating experience has shown that (ese components l'3 > usually pass the Surveillance when performed at the [18] month F'reguency which is based on the refteling cycle. Therefore,theheque,ncywasfoundtobeacceptabefroma reliability stand int. . E 7 REFERENCES 1. 10 CFR 50, Appendix A

                      > < Udos \ F1 hR, Set k u 5~.3 , GDC 41.

3 2'. FSAR,Section16.2.3). &

f. 36 4. (=sM , See % a o s.t vs W f R:3:1at:ry 0;id: 1.52, Rev. 2 1 9

p,u+4 O BWR/4 STS B 3.6-114 Rev. O, 09/28/92 l UNIT 2 l

       .k Vm t M elsid                                                                        -0PORS j.h           , sad BASES ACTIONS             E . D.2. and D.3 (continu                                             N ope    ions. Therefore, in eithe             ase, inability t suspend moveme     of irradiated fuel assembli would not be a sufficien        ason to require a reactor utdown.                           i B       lp Pu SURVEILLANCE           SR   3.6.4.'1.1 REQUIREMENTS                                                            .p Operating each SGT subsystem for n ensures that 9-" L] -*yr+- are                   (continuoushours0(,k PERABLE associated controls are functioning properly. It also and that al  4 ensures that blockage, fan or motor failure, or excessive             l vibration can be detected for corrective action. Operatio-y{with the heaters on (est::: tic 5ter cy:ltag t: :ir,t:i_
                                  " 4--t--S for ty10)t continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.

SR 3.6.4. 2 UsN 1 This SR verifies that the required GT filter testing is performed in accordance with the Ventilation Filter Testing

                     *p       'l accurueme
                                  . Program    -

(VFTg

                                                 ....... ......, y-
                                                                   ---- {Sgtgig.]t:;jrcirm
                                                                         ..-_ .---       n    The VFTP        -

includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP. 9 SR 3.6.4.1.3 QJd 2 This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. +wniie snis 7 Surveillance can be performed with the reactor at power, , operating experience has shown that these compo,neAnual}y j pass the Surveillance when performed at the f17 month Mn1 i Frequen he LUbR bThlt.M FUNCTIONAL TEST in SR 3.3.6.2.6 r aps this SR to provide complete testing of the safety _U (continued) BWR/4 STS B 3.6-113 Rev. O, 09/28/92 On G Z

g M *L Ml3J OMV SGT System /

  , p ,II ckays ni                                                              83.6.4.J S$eei<e.[fy uusbC"b                                                             b     ,

BASES SURVEILLANCE SR 3.6.4.3.3 (continued) REQUIREMENTS woe g nction) Therefore, the Frequency was found to be p,,.6 acceptable from a reliability standpoint. f't* . f SR 3.6.4.3.4 T is SR verifies that the filter ooler bypass damper an be op ed and the fan started. This nsures that the j vent lation mode of SGT System ope tion is available. I While this Surveillance can be perfo d with the reacto at power, perating experience has shown that these componen s f

         . l) P             usually h ss the Surveillance when per reed at the

[18] monthMrequency, which is based on he refueling cycle Therefore, th Frequency was found to be cceptable from a reliability stNndpoint. REFERENCES 1. 10 CFR 50, ".ppr. 9 8 . GDC 41. I X. FSAR,Section'{6.2.3). g h6 c. R voi.iorv Guiu. 1.32, R.,. @ Kroad bC) 2, OAA ySeehuJ f G l I O j BWR/4 STS B 3.6-114 Rev. O, 09/28/92 UNil $

    & ll41 SYA                                                                              fll"3 Ss1 systea B 3.6.4.

(f RUchcwye3wa+ S/edRes[(y &be el BASES C C ACTIONS 4.1, A.2 Ain'lh A3. (continued) V operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

1 SURVEILLANCE SR 3.6.4. 1 REQUIREMENTS ; Operating each SGT subsys em for a continuous hours > ensures that Bets sf:;:tas are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operatio q ' ; hr[with the heaters

                              +7e-""-d)(for          on (satesetic h:;ter cyclirii days a 10)Leontinuoushoursevery31        te ::.i-t:f         '

eliminates moisture n the adsorbers and HEPA filters. The - 31 day Frequency was developed in consideration of the known - reliability of fan motors and controls and the redundancy ; available in the system. ,

                                                             "~+ " "~ d '

O sa3.e.412 : This SR verifies that the requiredb6T filter testing is ; performed in accordance with the Ventilation Filter Testing Program (VFTP). -The SOT Syste; filici tu ne ;r: 4 cc:d- e wi+h Daani>+new r,uide 1.52 (naf ?); The VFTP Dff,o [ = includes testing HEPA filter performance, charcoal adsorbe; efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and 1 following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP. . SR 3.6.4JI.3 d b' l This SR verifies that each/SGT subsystem starts on receint of an actual or simQted initiation signal. awnsle this N

                                                                                      ~      ~

Surveillance can be performed with the reactor at power, operating pass experience the Surveillance whenhas shownatthat

                                                              >erformed       these the';f18 mont components 9       usu ,

f ine LOFIL sT51tPI FUNCTIONAL It3T in SR 3.3.ti.24 ' over aps this SR to provide compl?te testing of the safety (continued) ; BWR/4 STS B 3.6-113 Rev. O, 09/28/92 JN T 2

              - . ~ .

W Mdf I 2. 4/stod -teg' 7,.* Sar Q a((cLys wf B 3.6.4 SfectReaffy Nu~ Q,) BASES 9 - SURVEILLANCE SR 3.6.a.'i.3 (continued) REQUIREMENTS M Therefore, the Frequency was found to be

            #            acceptable from a reliability standpoint.

6

e 3 6.4.3.4 This R verifies that the filter cooler byp ss damper can be opene nd the fan started. This ensures th t the ventilat n mode of SGT System operation is a ilable. While this rveillance can be performed with t e reactor at i I power, operat g experience has shown that these omponents [ k 31 usually pass th Surveillance when performed at th ' [18] month Freque , which is based on the refueli cycle. j g? Therefore, the Freq cy was found to be acceptable f ma ; reliability standpoint. _ J REFERENCES - --1 , 0 cen 30, - n d i " , n u .- ie WriI FS M2, Svee noa S,3 ,

p. $c

2. FSAR,Section(6.2.31 3 < FsM, See%> r g.n.4 y

3. Regulatory co ue :.:2, ncf. [2] .

9 7&peC-i 1 BWR/4 STS B 3.6-114 Rev. O, 09/28/92 uN i T 2.

1 NUREG 1433 COMPARISON DOCUMENT - JUSTIFICATION O FOR DEVIATION O ' I l O

JUSTIFICATION FOR DEVIATION FROM NUREG 1433 () ITS: SECTION 1.0 - USE AND APPLICATION PLANT SPECIFIC DIFFERENCES i P.1 The Plant Hatch Unit 1 Improved Technical Specifications ! (ITS) do not include this definition. The definition is i only used in one Specifications in the NUREG, LCO 3.3.5.1, in a Surveillance Requirement. Since the Surveillance ' Requirement was not added to the Plant Hatch Unit 1 ITS, the definition was also not added. Refer to the Discussion of , Changes to NUREG 1433 for ITS: Section 3.3-Instrumentation, i for further discussion. P.2 The Plant Hatch Unit 1 ITS do not include this definition. The definition is only used in two Specifications in the NUREG, LCOs 3.3.6.1 and 3.3.6.2, in Surveillance ' Requirements. Since the Surveillance Requirements were not added to the Plant Hatch Unit 1 ITS, the definition was also not added. Refer to the Discussion of Changes to NUREG 1433 . for ITS: Section 3.3-Instrumentation, for further i discussion. P.3 LINEAR HEAT GENERATION RATE (LHGR) and MAXIMUM FRACTION OF ("' LIMITING POWER DENSITY (MFLPD) are not utilized in the Plant Hatch Units 1 and 2 ITS; therefore, these definitions are deleted. P.4 Brackets are removed and the proper plant specific optional wording /value is used, consistent with the current Plant Hatch Units 1 and 2 Technical Specifications. P.5 Brackets are removed and the optional wording is used. ; P.6 The proper FSAR section/ chapter and title, and proper specification /LCO numbers have been provided for each unit. P.7 Typographical / grammatical errors are corrected. P.8 The utilization of a Pressure and Temperature Limits Report (PTLR) requires the development, and NRC Staff approval, of l detailed methodologies for future revisions to P/T limits. At this time, Plant Hatch does not-have the necessary methodologies submitted to the NRC for review and approval. l Discussions with the NRC related to developing vendor generic methodologies have indicated that providing the limits specifically within the Technical Specifications, i rather than pursuing an approved PTLR presentation, is . acceptable. Therefore, the proposed presentation removes j references to the PTLR and proposes that the specific limits i and curves be included in the P/T Limits Specification (LCO 5O 4.5.9). i HATCH UNITS 1 and 2 1 REVISION E i _--_______l

l JUSTIFICATION FOR DEVIATION PROM NUREG 1433 ITS: SECTION 1.0 - USE AND APPLICATION GENERIC APPROVED /PENDING CHANGES TO NUREG 1433 GA.1 Change approved per package BWR-06 Item C1, 5/20/93 and BWR-06, Revision 1, Item C.3, 6/29/93. GA.2 Change approved per package BWR-05 Item C.1, Rev. 1, 6/29/93. T {M O HATCH UNITS 1 and 2 f REVISIONf{

i JUSTIFICATION FOR DEVIATION FROM NUREG 1433 ITS: SECTION 3.3 - INSTRUMENTATION m _ i PLANT SPECIFIC DIFFERENCES P.72 l (continued) Similar rationale applies for secondary containment isolation, ECCS initiation, and DG-LOP initiation. See Generic Change BWR-18 C.88 which has been approved. ! P.73 Clarification of " initiation capability" has been provided in I the Bases for the LOP instrumentation. This clarification includes the entire sequence of events which must be performed - by the logic for the LOP instrumentation to satisfy its intended Function. The reference to "DG initiation" was potentially [ confusing in the ITS and was changed to the generic " initiation" L with the appropriate clarification provided in the Bases. GENERIC APPROVED /PENDING CHANGES TO NUREG 1433 GA.1 Changed to be consistent with NUREG change package BWR-18, Items [ C.2 Rev 1, C.18 Rev 1, C.19, C.20, C.21 Rev 1, C.22, C.23 Rev 1, C.24 Rev 1, C.25, C.28, C.29, C.30, C.32, C.33, C.34, C.35, C.36, C.37, C.38, C.39 Rev 1, C.40, C.41, C.42, C.43, C.44, and C.45 C.88 Rev 1. Approved 3/39/94 and 5/6/94. ( , GA.2 Change approved per package BWR-01A, Item C.1, 3/20/93. GA.3 Changed to be consistent with NUREG change package BWR-19, Items , C.1, C.2, C.4, C.7, C.8, and C.9, approved 3/29/94. GA.4 Change approved per package BWR-06 Item C.9, Revs. 2 and 3, 10/13/93. GA.5 Change approved per package NRC-02, Items C.15 and C.21, ! 5/20/93. l i GA.6 Changed to be consistent with NUREG package BWOG-09, Item C.26, Rev 1 6/8/94. . i The final status of items annotated in the NUREG 1433 Comparison i Documents as GP is shown here. i [ i HATCH UNITS 1 AND 2 13 REVISION E

JUSTIFICATION FOR DEVIATION FROM NUREG 1433 () ITS: SECTION 3.4 - REACTOR COOLANT SYSTEM PLANT SPECIFIC DIFFERENCES (continued) P.28 The reason for this LCO is to ensure the initial assumption of the overpressure protection analysis is met. Therefore, the words have been modified to enforce this reason. P.29 To alleviate confusion (since the two units are different with regards to the location of all the transmitters and indicators) and to improve clarity of this Specification, this sentence has been deleted. P.30 The utilization of a Pressure and Temperature Limits Report (PTLR) requires the development, and NRC Staff approval, of detailed methodologies for future revisions to P/T limits. At this time, P367t Hatch does not have the necessary methodologies uubmitted to the NRC for review and approval. Discussions with the NRC related to developing vendor generic methodologies have indicated that providing the limits specifically within the Technical Specifications, rather than pursuing an approved PTLR presentation, is acceptable. Therefore, the proposed presentation removes references to the PTLR and proposes the specific limits and curves be included in the P/T Limits Specification (LCO O 3.4.9) and its Bases. ' P.31 The basic premise of SR 3.0.1 is that an SR is a requirement for the LCO compliance at all times during the Applicability or other specified condition (in this case, only MODES 1, 2, 3, and 4), not just at the time required for performance as specified in the Frequency. However, in the case of these two SRs, the intent is not consistent with the presentation. The recirc AT limits do not apply at all times in MODES 1, 2, 3, and 4. The temperature limits are only intended for the event of the pump start itself. The revised presentation corrects this oversight. P.32 The Bases wording added in BWR-18 regarding low power or loop flow is not included in this LCO and is not in the CTS. Therefore, this wording is not applicable. b t (~T i s_/ . HATCH UNITS 1 AND 2 5 REVISION E !

J 17IFICATION FOR DEVIATION FROM NUREG 1433

                '1S: SECTION 3.4 - REACTOR COOLANT SYSTEM              >

GENERIC APPROVED /PENDING CHANGES TO NUREG 1433 GA.1 Changed to be consistent with NUREG change package BWR-18 Items C.2 Rev 1, C.37, C.47 Rev 3, C.48, C.49, C.50, C.51, C.52, C.53, C.54, C.55, and C.56. Approved 3/29/94, 5/6/94 and 6/8/94. GA.2 Changed to be consistent with NUREG change package BWR-12 Item C.1, except minor changes were made in the LCO (deleting the ":" and the "a." and moving the words to be part of the sentence) for consistency. Approved 6/29/94. GA.3 Changed to be consistent with NUREG change package BWOG-02 Item C.3, rev 1. Approved 3/29/94. GA.4 Change approved per package BWR-03 Items C.1 (and its rev.

1) and C.3, 3/20/93.

GA.5 Changed to be consistent with NUREG change package BWR-19 Items C.10 and C.11. Approved 3/29/94. The final status of items annotated in the NUREG 1433 Comparison /~N Documents as GP is shown here. O HATCH UNITS 1 AND 2 ((b g REVISION E

JUSTIFICATION FOR DEVIATION FROM NUREG 1433 . ITS: SECTION 3.6 - CONTAINMENT SYSTEMS ( PLANT SPECIFIC DIFFERENCES (continued) l P.29 All secondary containment penetration flow paths have two ! isolation valves, therefore this Note is unnecessary and has been deleted. l r P.30 Since there are more installed RHR pumps than are required , to meet LCO 3.6.2.3, and more installed SGT subsystems than ! are required to meet LCO 3.6.4.3 (Unit 1) and LCO 3.6.4.9 (Unit 2); the word " required" has been added to the i applicable places, consistent with its use throughout the NUREG. t P.31 Comment number not used. , P.32 This SR has been deleted since the Hatch design does not ! include a filter cooler bypass damper and fan. ; P.33 A Note has been added to the ACTIONS to allow inspection of ! the Unit 1 hardened vent rupture disk while Unit 2 is operating. This inspection will cause both the Unit 1 SGT ! subsystems to be inoperable, thus the allowance is needed to continue operating Unit 2 while this inspection is being ; performed. Without this allowance, a dual unit shutdown O would be required. l It is expected that the hardened vent inspection from ; hanging of the first tag of the clearance through ! completion of the inspection and restoration to operability ; will take aproximately 12 hours. Should, during the course , of the inspection, there be an emergent need to return one of the tagged-out SGT trains to service, this action l (bolting the rupture disk flange, racking the breaker, and i removing clearance tags) should, at most, take approximately 1 hour. P.34 The proposed ACTION D added per NUREG change package BWR-04, Item C.8 has not been added into the Unit 2 ITS since -l it is not needed. The change was made to the NUREG because : a Condition allowed two SGT subsystems to be inoperable (NUREG Condition D), and confusion existed as to which requirement applied if the Unit was in MODE 1, 2, or 3 at i the same time fuel was being moved. Since the Unit 2 LCOs i have been split up with the MODES 1, 2, and 3 requirements in a different LCO than the handling of irradiated fuel ; assemblies in the secondary containment, the confusion does not exist. In the Unit 2 ITS, no ACTIONS exist in the j Operating LCO (MODES 1, 2, and 3) to allow two SGT , subsystems to be inoperable; therefore, LCO 3.0.3 will I apply, just as the proposed NUREG Condition requires. In l . addition, this proposed NUREG ACTION only applies for MODES I'

 .       1, 2, and 3. Therefore, it is not needed in the other two Unit 2 SGT System LCOs.

HATCH UNITS 1 AND 2 6 REVISION b

I I JUSTIFICATION FOR DEVIATION FROM NUREG 1433 (s) ITS: SECTION 3.7 - PLANT SYSTEMS PLANT SPECIFIC DIFFERENCES (continued) P.21 The proposed Plant Hatch testing is not all in accordance with RG 1.52, but it is in accordance with the VFTP as stated in the previous sentence. Thus the reference to RG 1.52 has been deleted. P.22 Reference 5, RG 1.52, does not specify the Frequency for this test. The proper basis for the Frequency has been added. The basis is consistent with other Bases for similar Surveillance Frequencies. P.23 The NRC staff approved Hatch licensing basis is "well within the limits of 10 CFR 100", thus this extra statement is not needed and has been deleted. P.24 The words have been changed to be consistent with the Plant Hatch design. P.25 The analysis that assumes the Main Turbine Bypass System to i be OPERABLE is the feedwater controller failure to maximum demand event transient analysis. Therefore, the Bases have been changed to reflect this analysis, ()) c P.26 Plant Hatch has an specific analysis for a fuel handling accident in the spent fuel storage pool (in addition to the analysis for a fuel handing accident in the RPV). The Plant Hatch 1.4 ensing basis is not "i 25% of 10 CFR 100 exposure guidelines NUREG-0800," but is well below the guideline doses of 10 CFR 100 and met the exposure guidelines of NUREG-0800. Therefore, the Bases have been modified. P.27 These words have been added to clarify that the boundary is not necessarily required to be leak-tight, but is required to meet the leak tightness requirements of SR 3.7.4.4 (i.e., leakage can occur as long as a 0.1 inch pressure is maintained in the control room). Also, an allowance to open control room access doors for entry and exit has been added since the design of the boundary only has one door. P.28 The words are changed consistent with the Plant Hatch analysis, described in the Applicable Safety Analyses section. P.29 "Or systems" has been added to the Note for SR 3.7.2.2 for clarification and consistency, such that the Note reads: g " Isolation of flow to individual components or systems does ('j not render PSW System inoperable." The NUREG 1433 Bases for this SR states: ... isolation of the PSW System to , components or systems may render those components or l HATCH UNITS 1 AND 2 4 REVISION A

r i JUSTIFICATION FOR DEVIATION FROM NUREG 1433 ; O ITS: SECTION 3.7 - PLANT SYSTEMS .i PLANT SPECIFIC DIFFERENCES (continued) ( P.29 (continued) l l systems inoperable, but does not affect the OPERABILITY of i the PSW System. As such, when all PSW pumps, valves, and ! piping are OPERABLE, but a branch connection off the main l header is isolated, the PSW System is still OPERABLE." An ' evaluation performed by the Hatch Architect Engineer confirms that an isolation of PSW flow to any system or ; component would not create a flow perturbation that would ! result in starving another system or component of PSW flow ! (i.e., flow imbalance resulting from system or component isolation is not a problem at Hatch.)

i I I f t

r i t l l l HATCH UNITS 1 AND 2~ f Zj[ REVISIONf(( l 1

i JUSTIFICATION FOR DEVIATION FROM NUREG 1433 i , ITS: SECTION 3.10 - SPECIAL OPERATIONS PLANT SPECIFIC DIFFERENCES ! P.1 Brackets have been removed and the proper value/words have j been used for each of the two units. Also, Bases changes + were made to be consistent with the Specifications. P.2 Due to the design of the Units 1 and 2 secondary containments, these changes have been made to reference the proper LCO for Unit 2. Refer to the Discussion of . Changes for Section 3.5, " Containment Systems." ! i P.3 Typographical /gramnatical errors have been corrected. ; 1 P.4 The allowances provided by these Specifications are not I needed at Plant Hatch; consequently, they have been i deleted. P.5 The proper references have been provided. P.6 The Startup Test Program has been completed at Plant Hatch; thus, a reference is not needed. P.7 Changes were made to provide for consistency with other Specifications.

P.8 Changes were made to either provide additional information or clarity, or to incorporate plant-specific terminology. P.9 The proposed hydrostatic testing requirement for RCS I temperature is > 212*F (plant specific value). Therefore a sentence saying that the requirement may eventually be greater than 200*F is unnecessary. The last sentence in the fourth paragraph has been deleted since this LCO is ! not exempting the Safety Limit from being met during a hydrostatic test. Therefore the Safety Limit is required to be met in accordance with SL 2.1.2. The temperature requirements are included in LCO 3.4.9. Additionally, a l sentence regarding the system test pressure is added. P.10 Since Plant Hatch does not have a reactor high water level , scram or a suppression pool makeup system; these - references have been deleted. t P.11 The previous sentence states that, the rod patterns assumed in the safety analysis may not be preserved. This sentence has been changed to state that a special CRDA analysis "may be" required. P.12 The correct power level (corresponding to the analysis , value) is 10% RTP. As written, the power level

 \                 corresponds to the low power setpoint.

1 HATCH UNITS 1 AND 2 1 REVISION E

4 p,a .

         -.~,%s  E.  . A,*s-w4     AA-4-m---4         4.-- .W-Je--  d+4 mm   aJ ,*< +   *-4   >,J L - - -+a* -.m rm-  X-i i

l JUSTIFICATION FOR DEVIATION FROM NUREG 1433 () ITS: SECTION 5.0 - ADMINISTRATIVE CONTROLS i PLANT SPECIFIC DIFFERENCES (continued) , P.17 Two notes are added to the NUREG for the Ventilation Filter i Testing Program. Note 1 implements a current Technical ! Specification Clarification concerning the impact of ! certain types of painting, buffing and grinding, and ! welding on Standby Gas Treatment System filters. l Evaluations have determined that the use of water based l paints, and the performance of metal grinding, buffing or l Welding is not detrimental to the charcoal filters of the : SGT System, either prior to or during operation. Note 2 is ! added to allow, in the future, the use of environmentally l friendly refrigerants equivalent to those specified in ASME i N510-1989, without requiring a Technical Specification I change. ; P.18 The actual allowable flowrates for the two systems are written out in the proposed ITS rather than specifying a i

10 range.

P.19 The actual allowable wattage values for the SGT System I heaters are written out in the proposed ITS rather than - specifying at 10% range. P.20 The provisions in the NUREG for Waste Gas Systems are for PWRs and not applicable to Plant Hatch. Quantities of ' radioactivity contained in all outdoor liquid radwaste tanks meeting the conditions of NUREG 5.7.2.14.e are i determined in accordance with the specified surveillance program. The sentence in the introductory paragraph is not j necessary to specify a method to determine liquid radwaste quantities. P.21 These provisions are only for PWRs and are.not applicable l for Plant Hatch. 1 P.22 The current Plant Hatch Technical Specifications do not contain requirements for testing of new fuel oil. Instead of the NUREG requirements for this testing, the Plant Hatch ; ITS contains requirements in current use to ensure i acceptability of the new fuel oil. , l' P.23 Current Technical Specifications for testing of fuel oil in the storage tanks is conducted on a 92 day interval. This current interval has been found acceptable for use at Plant Hatch and is being retained in the proposed ITS. 1 HATCH UNITS 1 AND 2 3 REVISION Ei

I JUSTIFICATION FOR DEVIATION FROM NUREG 1433 j () ITS: SECTION 5.0 - ADMINISTRATIVE CONTROLS j l PLANT SPECIFIC DIFFERENCES (continued) i P.24 The prcA T sions of SR 3.0.2 and SR 3.0.3 would have been ! applicable to the diesel fuel oil testing provisions if : they had been left in the LCOs of Section 3.8. Since these [ Section 3.0 provisions are not generally applicable to Administrative Controls, then the applicability must te specifically stated in Section 5.0 provisions. ; i P.25 The description of the entry conditions into the SFDP are t clarified and generalized to assure that they include all possible required entry conditions. See Also BWR 25.C3 which was approved 6/7/94. ; l P.26 A clarification is added to include in the annual ! occupational radiation exposure report, only those other personnel for whom monitoring was required. This change f does not modify the present intent of the NUREG. j P.27 A clarification is added to the monthly operating report requirement to state that the safety / relief valves are ! those for " main steam." This change does not modify the present intent of the NUREG. 7 P.28 Comment number not used. L t P.29 Reference to LCO 3.3.3.1 for Post Accident Monitoring Instrumentation is all that is necessary to locate this I special reporting requirement. P.30 Changes to be consistent with plant specific terminology. i P.31 Changes to clarify the control room command function and ! shift crew composition for a dual unit plant with a common control room. l P.32 A direct reference to 10 CFR 50.54 for determination of minimum shift crew composition is added. Without this - reference, ITS 5.2.2.b could be incorrectly construed to define these requirements. ; i l l

l (:) HATCH UNITS 1 AND 2 4 REVISION

i JUSTIFICATION FOR DEVIATION FROM NUREG 1433 O ITS: SECTION 5.0 - ADMINISTRATIVE CONTROLS PLANT SPECIFIC DIFFERENCES (continued) P.33 The diesel generator accelerated test frequency ' requirements are relocated in their current licensing bases ! form to plant procedures. A plant procedure implements the current Technical Specifications requirements, as well as the requirements and responsibilities for tracking emergency DG failures for the determination and reporting of reaching trigger values specified in NUMARC 87-00. These requirements are more restrictive than those specified in NUREG 1433. P.34 The staff's October 25, 1993 correspondence proposed a change to 5.2.2.f and 5.1.2 to require the person satisfying the control room command function (5.1.2) and the (Operations Manger) to have an active SRO license. Current Specifications only require that the [ Operations Manager] " hold" an SRO license. For consistency with 5.1.2 : (as accepted in BWR-09, C.2), 5.2.2.f has been clarified that the operations manager shall hold an active or inactive SRO license. r P.35 Changes to the program description have been made to s eliminate the references to ASME XI and applicable Addenda as required by 10 CFR 50.55a. The deletion of item (a) of ; section 5.7.2.12 is consistent with the NRC letter dated October 25, 1993 regarding section S.O. Frequencies listed 1 in the NUREG which are not in the CTS have been deleted. ' P.36 NUREG 1433 is modified consistent with Hatch current ' Technical Specifications which in turn are consistent with 10 CFR 20. Plant specific nomeclature is used where applicable. P.37 Consistent with the current Hatch Technical Specifications the limitation on liquid and gaseous effluents have been modified. The SER for these limitations were issued with Unit 1 Amendment 190 and Unit 2 Amendment 129. , P.38 The uti]ization of a Pressure and Temperature Limits Report (PTLR) requires the development, and NRC Staff approval, of :' detailed methodologies for future revisions to P/T limits. At this time, Plant Hatch does not have the necessary methodologies submitted to the NRC for review and approval. Discussions with the NRC related to developing vendor i generic methodologies have indicated that providing the ' limits specifically within the Technical Specifications, rather than pursuing an approved PTLR presentation, is acceptable. Therefore, the proposed presentation removes O'" references to the PTLR and proposes the specific limits and curves be included in P/T limits Specification LCO 3.4.9. HATCH UNITS 1 AND 2 g([( REVISION E

f- JUSTIFICATION FOR DEVIATION FROM NUREG 1433 ITS: SECTION 5.0 - ADMINISTRATIVE CONTROLS PLANT SPECIFIC DIFFERENCES (continued) l P.38 (continued) Removal of the PTLR description in the Administrative Controls section also results in renumbering the PAM i Report. Generic changes associated with the PTLR are not shown. ! t GENERIC APPROVED /PENDING CHANGES TO NUREG 1433 , GA.1 Changed to be consistent with NUREG change package BWOG-09, Items C.1 through C.4, C.6 through C.16 and C.19 through C.24, 3/29/94. ; GA.2 Change approved per package WOG-06, Items C.1, C.5, and , C.7, 3/20/93. GA.3 Change approved per package BWR-06, Item C.7, S/20/93. GA.4 Change approved per package NRC-02, Item C.22, 5/20/93. ; [ Final status of items annotated in the NUREG 3433 Comparison Documents as GP is shown here. , k [ O HATCH UNITS 1 AND 2 (( REVISION E}}

ML20073G963

Text

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