Text

Proposed Tech Specs Re Elimination of Selected Response Time Testing Requirements from TS
	ML20082L333
Person / Time
Site:	Hatch
Issue date:	04/14/1995
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20082L329	List: HL-4789, Application for Amend to License NPF-5 Re Elimination of Selected Response Time Testing Requirements from TS; ML20082L333;
References
	NUDOCS 9504210113
	Download: ML20082L333 (40)
	v • d • e

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a c . Enclosura 3 l Edwin I. Hatch Nuclear Plant - Unit 2 Request to Revise Technical Specifications: [

Response Time Testing !

Page Change Instructions l

Replace the following pages of the Technical Specifications with the new enclosed page.

Eagg Instruction 3.3-6 Replace 3.3-41 Replace -

3.3-42 Replace -

3.3-43 Replace !

3.3-44 Replace 3.3-55 Replace .;

3.5-6 Replace l 3.5-6a Add j 3.5-6b Add i

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l Hl.-4789 E3-1 9504210113 950414 PDR ADOCK 05000366 P PDR

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RPS Instrumentation 3.3.1.1 l l

l SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.16 ------------------NOTES------------------

1. Neutron detectors are excluded.

2. For Functions 3 and 4, channel :

sensors are excluded. !

3. For Function 5, "n" equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.

Verify the RPS RESPONSE TIME is within 18 months on a >

limits. STAGGERED TEST BASIS l i

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b HATCH UNIT 2 3.3-6 t:s-psh.ichs.i.sunii21.p.c.spropo asl35-495 -

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. RPS Instrumentaticn i 3.3.5.1 l i

SVRVEILLANCE REQUIREMENTS i

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 f 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 initiation capability.

SURVEILLANCE FRl Q'JENCY 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 i

HATCH UNIT 2 3.3-41 k:1-psh.ichsit sunii2spropo as135-495

o ECCS Instrumentation 3.3.5.1 Table 3.3.5.1 1 (page 1 of 6)

Emergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWASLE 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 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 I

b. Drywell 1,2,3 4(b) B SR 3.3.5.1.1 s 1.92 pois Pressure - High S!t 3.3.5.1.2 SR 3.3.5.1.4 SL 3.3.5.1.5 I

c. Reactor Steam Dome 1,2,3 4 C SR 3.3.5.1.1 e 390 pois Pressure - Low SR 3.3.5.1.2 and (Injection Permissive) SR 3.3.5.1.4 s 476 pois SR 3.3.5.1.5 '

1 4(a), $(a) 4 8 SR 3.3.5.1.1 2 390 psig SR 3.3.5.1.2 and SR 3.3.5.1.4 5 476 psis SR 3.3.5.1.5 1

d. Core Spray Pum 1,2,3, 1 par E SR 3.3.5.1.1 a 570 gpm Discharge Flow - Low subsystem SR 3.3.5.1.2 and (Bypass) 4(a), $(a) SR 3.3.5.1.4 s 745 spm SR 3.3.5.1.5 1

2. Low Pressure coolant Injection (LPCI) System

a. Reactor vessel Water 1,2,3, 4(b) B SR 3.3.5.1.1 t -113 inches Level - Low Low Low, SR 3.3.5.1.2 Level 1 4(*), 5(a) SR 3.3.5.1.4 SR 3.3.5.1.5 (continued)

(a) Wher, associated stbsystem(s) are required to be OPERABLE.

1 (b) Also required to initiate the associated diesel generator (DG) and isolate the associated plant service I water (PSW) turbine building (T/8) isolation valves.

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I HATCH UNIT 2 3.3-42 k:\wpihatch\its\ unit 2\ specs \ proposed \135 - 495

.. ECCS Instrumentatien 3.3.5.1 Table 3.3.5.1 1 (page 2 of 6)

Emergency Core Cooling System Instru.entation I

k APPLICAgLE CONDITIONS MODES REQUIRED REFERENCED j OR OtHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOW 48LE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE .

2. LPCI System (continued) 1,2,3 4(b) B SR 3.3.5.1.1 s 1.92 pois +

b. Drywell SR 3.3.5.1.2 Pressure - High SR 3.3.5.1.4 -

SR 3.3.5.1.5 t I

c. Reactor Steam Dome 1,2,3 4 C SR 3.3.5.1.1 t 390 pelg 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 l l

4(a),5(a) 4 8 SR 3.3.5.1.1 2 390 psig SR 3.3.5.1.2 and SR 3.3.5.1.4 5 476 peig SR 3.3.5.1.5 1

d. Reactor Steam Dome 1(c) 2(c)

, , 4 C SR 3.3.5.1.1 t 335 psig Pressure - Low SR 3.3.5.1.2 (Recirculation 3ICI SR 3.3.5.1.4 Discharge Valve SR 3.3.5.1.5 Permissive)

e. Reactor vessel Shroud 1,2,3 2 B SR 3.3.5.1.1 2 202 inches ,

Level - Level 0 SR 3.3.5.1.2 !

SR 3.3.5.1.4 SR 3.3.5.1.5

f. Low Pressure Coolant 1,2,3, 1 per C SR 3.3.5.1.4 . .

Injection Ptap pung SR 3.3.5.1.5 -

Start - Time Delay 4(a), $(a) i Relay ;

Pumps A,8,0 t 9 seconds and s 11 seconds Ptmp C 5 1 second (continued) l (a) When associated sub,.ystem(s) are required to be OPERABLE. I (b)- Also required to initiate the associated DG and isolate the associated PSW T/B isolation valves.

(c) With associated recirculation pinp discharge valve open. i i

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HATCH UNIT 2 3.3-43 L:swpsh.ichsit suni2s.Pec.spropo eds135-495 i

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. ECCS Instrumentatien 3.3.5.1 i

Table 3.3.5.1 1 (page 3 of 6)

Emergency Core Cooling System Instrtmentation 1

APPLICABLE CONDITIONS j MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWASLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

2. LPCI System (continued) ,

s. Low Pressure 1,2,3, 1 per E SR 3.3.5.1.1 a 1675 som !

Coolant Injection Ptap subsystem SR 3.3.5.1.2 and Discharge Flow - Low 4(a), $(s)' SR 3.3.5.1.4 s 2215 spa (Bypass) SR 3.3.5.1.5

3. High Pressure Coolant >

Injection (HPCI) System

a. Reactor vessel Water 1, 4 8 SR 3.3.5.1.1 2 47 inches Level - Low Low, SR 3.3.5.1.2 e Level 2 2(d), 3(d) SR 3.3.5.1.4 SR 3.3.5.1.5 I

b. Drywell 1, 4 s SR 3.3.5.1.1 s 1.92 psig Pressure - Nigh SR 3.3.5.1.2 2(d) 3(d)

, SR 3.3.5.1.4 SR 3.3.5.1.5 I

c. Reactor vessel Water 1, 2 C SR 3.3.5.1.1 s 56.5 inches Level - High, Level 8 SR 3.3.5.1.2 2(d), 3(d) SR 3.3.5.1.4 SR 3.3.5.1.5 1

d. Condensate Storage 1, 2 0 SR 3.3.5.1.3 2 2.61 ft Tank Level - Low SR 3.3.5.1.5

e. Suppression Pool Water 1, 2 0 SR 3.3.5.1.1 s 154 inches Level - High SR 3.3.5.1.2 2(d),3(d) SR 3.3.5.1.4 SR 3.3.5.1.5 (continued)

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

(d) With reactor steam dome pressure > 150 psig.

t HATCH UNIT 2 3.3-44 k:\wp\hstch\its\ unit 2\ spec s\ proposed \13 5- 495

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. Priaary. Containment Isolation Instrumentation !

3.3.6.1 l SURVEILLANCE REQUIREMENTS l

_______________________--------------NOTES------------------------------------ !

1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary 4

Containment Isolation Function. -

2. When a channel is placed in an inoperable status solely for performance of l 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 l maintains isolation capability. ;

l O'RVEILLANCE FREQUENCY ;

i t

SR 3.3.6.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

i SR 3.3.6.1.2 Perform CKKtN~L FUNCTIONAL TEST. 92 days !

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i SR 3.3.6.1.3 Perform CHANNEL CALIBRATION. 92 days >

1 SR 3.3.6.1.4 Perform CHANNEL FUNCTIONAL TEST. 184 days !

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SR 3.3.6.1.5 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months ,

i SR 3.3.6.1.7 ------------------NOTE- ---------------- ,

Channel sensors are excluded. l l Verify the ISOLATION SYSTEM RESPONSE TIME 18 months on a is within limits. STAGGERED TEST BASIS i h

HATCH UNIT 2 3.3-55 k:\wp\ hatch \its\ unit 2\ spec sipregn> sed \ l 35-495 i

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ECCS--Op: rating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.9 -------------------NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test. >

Verify, with reactor pressure s 165 psig, 18 months the HPCI pump can develop a flow rate 2 4250 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.10 -------------------NOTE--------------------

Vessel injection / spray may be excluded.

Verify each ECCS injection / spray subsystem 18 months actuates on an actual or simulated automatic initiation signal.

SR 3.5.1.11 -------------------NOTE--------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or 18 months simulated automatic initiation signal.

SR 3.5.1.12 -------------------NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify each ADS valve opens when manually 18 months actuated.

(continued)

HATCH UNIT 2 3.5-6 L:\wp\ hatch \its\ unit 2\ spec s\ proposed \l 3 5 - 495 l

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ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)- !

1 SURVEILLANCE

[ FREQUENCY j SR 3.5.1.13 -------------------NOTE--------------------

L ECCS injection / spray initiation .

instrumentation response time may be t t

assumed from established limits.

1 Verify each ECCS injection / spray subsystem 18 months i ECCS RESPONSE TIME is within limits.

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ECCS-Opcrating 3.5.1 P

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RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.16 ------------------NOTES------------------

1. Neutron detectors are excluded.

3 K. For Function 5, "n" eguals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.

Verify the RPS RESPONSE TIME is within 18 months on a limits. STAGGERED TEST ,

BASIS

2. for func$lon s 3 and 4, e kan,e I Sensor 5 are e,Kcluded.

HATCH UNIT 2 3.3-6 Amendment No. 135 G

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 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 CAllBRATION. 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 ' ?.5 ! 6 Verify th: ECCS RESPONS: T"i: i; with4n 18 mentbr "" a limits. STAGGEREL TE5T-4A6Mr-

,=

HATCH UNIT 2 3.3-41 Amendment No. 135

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.,_ ECCS Instrumentation 3.3.5.1 I

Table 3.3.5.1-1 (page 1 of 6)

Emergency Core Cooting System Instrumentation i

l APPLICABLE CONDITIONS j MODES REQUIRED REFERENCED i OR OTHER CHANNELS FROM l SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE i FUNCTION CONDlil0NS 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 inches Level - Low Low Low, SR 3.3.5.1.2 Level 1 4(a), $(a)

SR 3.3.5.1.4 1 SR 3.3.5.1.5 l

- 4 N=4=&mEmind. !

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 i N

l

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

4(*), 5(a) 4 8 SR 3.3.5.1.1 a 390 psig. !

SR 3.3.5.1.2 and SR 3.3.5.1.4 5 476 psig l SR 3.3.5.1.5

I ? ! " .1 -

d. Core Spray Punp 1,2,3, 1 per E SR 3.3.5.1.1 t 570 spm Discharge F tow - Low subsystten SR 3.3.5.1.2 and (Bypass) 4(a), $(a) sR 3.3.5.1.4 s 745 spm SR 3.3.5.1.5 .j l

2. Low Pressure Coolant injection (LPCI) System !

a. Reactor Vessel Water 1,2,3, 4(D) B SR 3.3.5.1.1 t--113 inches Level - Low Low Low, SR 3.3.5.1.2 Levet 1 4(a) $(a)

, SR 3.3.5.1.4 SR 3.3.5.1.5

dBamam&4 bb (continued)

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

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

HATCH UNIT 2 3.3-42 Amendment No. 135 ,

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. ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 2 of 6)

E'ergency Core Cooling System Instrumentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE

2. LPCI System (continued) 1,2,3 4(b) B SR 3.3.5.1.1 5 1.92 psig

b. Drywell SR 3.3.5.1.2 Pressure - High SR 3.3.5.1.4 SR 3.3.5.1.5 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 5 476 psig SR 3.3.5.1.5 4(a),$(a) 4 8 SR 3.3.5.1.1 e 390 psig SR 3.3.5.1.2 and SR 3.3.5.1.4 5 476 psig SR 3,.3.5.1.5

- ,,.1

d. Reactor Steam Dome 1ICI,2ICI, 4 C SR 3.3.5.1.1 2 335 psig Pressure - Low SR 3.3.5.1.2 l (Recirculation 3(c) SR 3.3.5.1.4 Discharge Valve SR 3.3.5.1.5 l

1 Permissive)

e. Reactor vessel Shroud 1,2,3 2 B SR 3.3.5.1.1 1 202 inches Level - Level 0 SR 3.3.5.1.2 !

SR 3.3.5.1.4 j g, SR 3.3.5.1.5 '

l

f. Low Pressure Coolant 1,2,3, 1 per C SR 3.3.5.1.4 Injection Pump l ptanp SR 3.3.5.1.5 1 Start - Time Delay 4(83,5(8)

Relay Pumps A,B,D t 9 seconds and s 11 seconds '

Pump C 5 1 second (continued)

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

(b) Also required to initiate the associated DG and isolate the associated PSW 1/B isolation valves.

(c) With associated recirculation pump discharge valve open.

HATCH UNIT 2 3.3-43 Amendment No. 135

. ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 6)

Emergency Core Cooling System Instrtanentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACilDN A.1 REQUIREMENTS VALUE

2. LPCI System (continued) 1,2,3, 1 per E SR 3.3.5.1.1 2 1675 spm

g. Low Pressure subsystem SR 3.3.5.1.2 and Coolant Injection Pump 4(a), $(a) SR 3.3.5.1.4 5 2215 spm Discharge Flow - Low (Bypass)

SR 3.3.5.1.5

3. High Pressure Coolant injection (HPCI) System

a. Reactor Vessel Water 1, 4 B SR 3.3.5.1.1 m -47 inches Level - Low Low, SR 3.3.5.1.2 Level 2 2(d), 3(d) SR 3.3.5.1.4 SR 3, . 3, . 5,.

.. 1. 5~

b. Drywell 1, 3.3.5.1.1 4 B SR s 1.92 psig Pressure - High SR 3.3.5.1.2 2(d) 3(d)

. SR 3.3.5.1.4 SR 3, . 3, . 5,. .1. 5

c. Reactor vessel Water 1, 2 C SR 3.3.5.1.1 s 56.5 inches Level - High, Level 8 SR 3.3.5.1.2 2(d), 3(d) SR 3.3.5.1.4 SR 3.3.5.1.5

-- :.:. . .i

)

d. Condensate Storage 1, 3.3.5.1.3 1

2 D SR a 2.61 ft '

Tank Level - Low SR 3.3.5.1.5 !

e. Suppression Pool Water 1, 2 SR 3.3.5.1.1 I

D s 154 inches !

Level - High SR 3.3.5.1.2 2(d), 3(d) SR 3.3.5.1.4 l

1 SR 3.3.5.1.5 l (continued)

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

(d) With reactor steam dome pressure > 150 psig.

l 1

i HATCH UNIT 2 3.3-44 Amendment No. 135

Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation function.

2. When a channel is placed in an inoperable status sol.ely for performance of required Surveillances, entry i: to 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 ,

i maintains isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.1.3 Perform CHANNEL CALIBRATION. 92 days SR 3.3.6.1.4 Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.6.1.5 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months i'

SR 3.3.6.1.7 ------------------NOTE-------------------

y ba ^c

o: diction detect;r:_______________________________'eded.

Verify the ISOLATION SYSTEM RESPONSE TIME 18 months on a is within limits. STAGGERED TEST ,

BASIS 0 0nne$ itnSofS OfC CKCSUE*N' HATCH UNIT 2 3.3-55 Amendment No. 135 l

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' ECCS - Operating 3.5.1 <

l 1

1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY !

i SR 3.5.1.9 -------------------NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l

after reactor steam pressure and flow are '

adequate to perform the test.

Verify, with reactor pressure s 165 psig, 18 months the HPCI pump can develop a flow rate 2: 4250 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.10 -------------------NOTE--------------------

Vessel injection / spray may be excluded.

Verify each ECCS injection / spray subsystem 18 months actuates on an actual or simulated i automatic initiation signal.

SR 3.5.1.11 -------------------NOTE--------------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or 18 months simulated automatic initiation signal.

SR 3.5.1.12 -------------------NOTE--------------------

N:t required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify each ADS valve opens when manually 18 months actuated.

Il'rtseri d- ---JA 4

I HATCH UNIT 2 3.5-6 Amendment No. 135 l

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ax s n SQ r

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SR 3.5.1.13 -------------------NOTE--------------------

ECCS injection / spray initiation instrumentation response time may be assumed from established limits.

Verify each ECCS injection / spray subsystem 18 months ECCS RESPONSE TIME is within limits.

i Enclosure 4 Edwin I. Hatch Nuclear Plant - Unit 2 Request to Revise Technical Specifications:

Response Time Testing Bases Changes I

RPS Instrumentation B 3.3.1.1 ,

BASES SURVEILLANCE SR 3.3.1.1.14 REQUIREMENTS (continued) The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function uses an electronic filter circuit to generate a signal proportional to the core THERMAL POWER from the APRM neutron flux signal. This filter circuit is representative of the fuel heat transfer '

dynamics that produce the relationship between the neutron '

flux and the core THERMAL POWER. The time constant is specified in the COLR and must be verified to ensure that the channel is accurately reflecting the desired parameter.

The Frequency of 18 months is based on engineering judgment considering the reliability of the components.

SR 3.3.1.1.15 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LC0 3.1.3), and SDV vent and drain valves (LCO 3.1.8),

overlaps this Surveillance to provide complete testing of the assumed safety function.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply 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 that these components usually i pass the Surveillance when performed at the 18 month t Frequency.

SR 3.3.1.1.16 '

This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME acceptance criteria are included in Reference 10.

Note 1 allows neutron detectors to be excluded from RPS l l RESPONSE TIME testing because the principles of detector j operation virtually ensure an instantaneous response time.

l (continued)

HATCH UNIT 2 B 3.3-31 Uwp\ hatch \its\ unit 2\ba ses\ proposed \0 - 495 !

RPS Instrumentation >

B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.16 (continued)

REQUIREMENTS Note 2 allows channel sensors for Reactor Vessel Steam Dome Pressure - High and Reactor Vessel Water Level - Low, Level 3 (Functions 3 and 4) to be excluded from RPS RESPONSE TIME testing. This allowance is supported by Reference 12 which concludes that any significant degradation of the channel sensor response time can be detected during the ;

performance of other Technical Specifications SRs.

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HATCH UNIT 2 B 3.3-31b 6 swp\ hatch \its\ unit 2\ bases \proponed\0-495 l

l 4 RPS Instrumentation l 8 3.3.1.1 BASES SURVEILLANCE SR 3.3.1 1.16 (continued)

REQUIREMENTS RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS l Frequency to be determined based on four channels per trip system, in lieu of the eight channels specified in Table 3.3.1.1-1 for the Main Steam Line Isolation Valve-Closure Function. This Frequency is based on the logic interrelationships of the various channels required to produce an RPS scram signal. This Frequency is consistent i with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.

REFERENCES 1. FSAR, Section 7.2.

2. FSAR, Chapter 15,

3. FSAR, Section 6.3.3.

4. FSAR, Supplement SA.

5. FSAR, Section 15.1.12.

6. NED0-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.

7. FSAR, Section 15.1.38.

8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.

9. NED0-30851 P- A, " Technical Specification Improvement Analyses for BNR Reactor Protection System,"

March 1988.

10. Technical Requirements Manual.

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

12. NED0-32291, " System Analyses for Elimination of Selected Response Time Testing Requirements,"

January 1994.

HATCH UNIT 2 B 3.3-32 knwp\ hatch \its\ unit 2\ bases \ proposed \0- 4 95 l

. ECCS fnstrurentatisn f> 3.3.5.1 BASES APPLICABLE ECCS instrumentation satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES, Statement (Ref. 6). Certain instrumentation Functions are l LCO, and retained for other reasons and are described below in the APPLICABILITY individual Functions discussion.

(continued) ,

The OPERABILITY of the ECCS instrumentation is dependent l upon the OPERABILITY of the individual instrumentation >

channel Functions specified in Table 3.3.5.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). Table 3.3.5.1-1, footnote (b), is added to l show that certain ECCS instrumentation Functions are also required to be OPERABLE to perform DG initiation and 1 actuation of the PSW T/B isolation.

l Allowable Values are specified for each ECCS Function specified in the table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not ;

exceed the Allowable Value between CHANNEL CALIBRATIONS.

Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis, where applicable. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined, accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate prote. tion because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

(continued)

HATCH UNIT 2 B 3.3-105 k:\wp\hakh\its\ unit 2\ bases \ proposed \0 - 495 i

1 ECCS Instrumentatien B 3.3.5.1 1

l BASES )

SURVEILLANCE SR 3.3.5.1.5 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LC0 3.5.1, LCO 3.5.2, LC0 3.7.2, LCO 3.0.1, and LCO 3.8.2 overlaps this Surveillance to complete testing of the assumed safety function.

The 18 month Frequency is based on the need to perform this :

Surveillance under the conditions that apply 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 that these components usually <

pass the Surveillance when performed at the 18 month Frequency.

1 (continued)

HATCH UNIT 2 B 3.3-133 k:\wp\ hatch \its\ unit 2ihases\ proposed \0-495

. ECCS Instrumentaticn .!

B 3.3.5.1 ;

BASES (continued)

REFERENCES 1. FSAR, Section 5.2. ,

, 2. FSAR, Section 6.3.

3. FSAR, Chapter 15.

4. NEDC-31376-P, "Edwin I. Hatch Nuclear Power Pl ant, -

SAFER /GESTR-LOCA, Loss-of-Coolant Accident Analysis,"

December 1986.

5. NEDC-30936-P-A, "BWR Owners' Group Technical Specification Improvement Analyses for ECCS Actuation Instrumentation, Part 2," December 1988.

l

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

HATCH UNIT 2 B 3.3-134 k:\wp\ hatch \its\ unit 2\ bases \proposeds0- 495

Primary Containment Isolation Instrumentatien B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.7 (continued)

REQUIREMENTS ISOLATION SYSTEM RESPONSE TIME acceptance criteria are ,

included in Reference 6. This test may be performed in one measurement, or in overlapping segments, with verification that all components are tested.

i A Note to the Surveillance states that the channel sensors are excluded from ISOLATION SYSTEM RESPONSE TIME testing.

The exclusion of the channel sensors is supported by Reference 8 which indicates that the sensors' response times ,

are a small fraction of the total response time. Even if the sensors experienced response time degradation, they would be expected to respond in the microsecond to ,

millisecond range until complete failure.

ISOLATION SYSTEM RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. This Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience that shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.

i REFERENCES 1. FSAR, Section 6.3.

2. FSAR, Chapter 15.

3. FSAR, Section 4.2.3.4.2.

4. NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," !

July 1990.

5. NEDC-30851P-A Supplement 2, " Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.

6. Technical Requirements Manual.

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

8. NED0-32291, " System Analyses for Elimination of ,

Selected Response Time Testing Requirements,"

January 1994.

HATCH UNIT 2 8 3.3-174 k:swpsh tchsii.sumt2sh..e.sproposeds0-495

, i ECCS - Operating B 3.5.1 BASES l SURVEILLANCE SR 3.5.1.12 (continued) -

SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.5.1.13 This SR ensures that the ECCS RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis. Response time testing acceptance criteria are included in Reference 14. A Note to the Surveillance states that the instrumentation portion of the response time may be assumed from established limits. The exclusion of the instrumentation from the response time surveillance is supported by Reference 15, which concludes that instrumentation will continue to respond in the microsecond to millisecond range prior to complete failure.

The 18 month Frequency is based on the need to perform the Surveillance under the conditions that apply 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 that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1. FSAR, Section 6.3.2.2.3.

2. FSAR, Section 6.3.2.2.4.

3. FSAR, Section 6.3.2.2.1.

4. FSAR, Section 6.3.2.2.2.

5. FSAR, Section 15.1.39.

6. FSAR, Section 15.1.40.

7. FSAR, Section 15.1.33.

(continued)

HATCH UNIT 2 B 3.5-16 t:swes h.ichsii.sunii2sh. .sprono.easl35-495

ECCS -- Operating B 3.5.1 BASES REFERENCES 8. 10 CFR 50, Appendix K.

(continued) .

9. FSAR, Section 6.3.3.

10. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 ,

SAFER /GESTR-LOCA Loss-of-Coolant Analysis,"

December 1986. .

11. 10 CFR 50.46.

12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.

(NRC), " Recommended Interim Revisions to LCOs for ECCS .

Components," December 1, 1975. :

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

14. Technical Requirements Manual.

15. NED0-32291, " System Analyses for Elimination of Selected Response Time Testing Requirements,"

January 1994.

l i

l l

HATCH UNIT 2 B 3.5-16a kawpsh.ichsii. son;i2sh.u.sproco a1135-495 l

. i

[CCS - Op:. rating B 3.5.1 i

This page intentionally left blank.

o i

HATCH UNIT 2 8 3.5-16b k:\wr eatch\its\ unit 2\ bases \pgmd\l35-495 l

. RPS Instrumentation ,

B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.14 REQUIREMENTS (continued) The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function uses an electronic filter circuit to generate a signal proportional to the core THERMAL POWER from the APRM neutron flux signal. This filter circuit is representative of the fuel heat transfer dynamics that produce the relationship between the neutron flux and the core THERMAL POWER. The time constant is specified in the COLR and must be verified to ensure that the channel is accurately reflecting the desired parameter.

The Frequency of 18 months is based on engineering judgment considering the reliability of the components.

SR 3.3.1.1.15 The LOGIC SYSTEM FUNCTIONAL TEST demonstrate :ne OPERABILITY of the required trip logic for a specific ,

channel. The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LC0 3.1.8), I overlaps this Surveillance to provide complete testing of l the assumed safety function.

The 18 month Frequency is based on the need to perform this {

Surveillance under the conditions that apply during a plant i outage and the potential for an unplanned transient if the i Surveillance were performed with the reactor at power.

Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.

SR 3.3.1.1.16 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification l that all components are tested. The RPS RESPONSE TIME l acceptance criteria are included in Reference 10.

TuIron detectors W excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time.

E3 (continued)

HATCH UNIT 2 B 3.3-31 REVISION 0

Insert A Note 2 allows channel sensors for Reactor Vessel Steam Dome Pressure - High and Reactor Vessel Water Level - Low, Level 3 (Functions 3 and 4) to be excluded from RPS RESPONSE TIME testing. This allowance is supported by Reference 12 which concludes that any -

significant degradation of the channel sensor response time can be detected during the performance ofother Technical Specifications SRs.

l i

l l

I

. RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.16 REQUIREMENTS (continued) 3 RPSRESPONSETIMEtestsare[onductedonan18 month STAGGERED TEST BASIS. Note A requires STAGGERED TEST BASIS Frequency to be determined based on four channels per trip system, in lieu of the eight channels specified in Table 3.3.1.1-1 for the Main Steam Line Isolation Valve-Closure Function. This Frequency is based on the logic interrelationships of the various channels required to produce an RPS scram signal. This Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.

REFERENCES 1. FSAR, Section 7.2.

2. FSAR, Chapter 15.

3. FSAR, Section 6.3.3.

4. FSAR, Supplement 5A.

5. FSAR, Section 15.1.12.

6. NED0-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.

7. FSAR, Section 15.1.38.

)

8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1,1980.

9. NED0-30851-P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System,"

March 1988.

10. Technical Requirements Manual.

l

11. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993. l I 3.

UG Do- 322 91. %+em Analvses for Elimin<+ien of Selec. fed Ke.sponse %eTn4:n3 R9 vow.,tz

Tann ry 1944 HATCH UNIT 2 B 3.3-32 REVISION O l

~

\

(

ECCS Instrumentation <

B 3.3.5.1 l l

BASES b

l i

APPLICABLE ECCS instrumenta ion satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES, Statement (Ref. Certain instrumentation Functions are LCO, and retained for other reasons and are described below in the '

APPLICABILITY individual Functions discussion.

(continued) ,

The OPERABILITY of the ECCS instrumentation is dependent i upon the OPERABILITY of the individual instrumentation l channel Functions specified in Table 3.3.5.1-1. Each '

Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal i, b= ; I iabi[3 5.1-1 footnote (b), is added to show that certain ECCS instrumentation Functions are also required to be OPERABLE to perform DG initiation and actuation of the PSW T/B isolation.

Allowable Values are specified for each ECCS Function specified in the table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS.

Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device '

(e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis, where applicable. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined, accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation ,

uncertainties, process effects, calibration tolerances, l instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

(continued)

HATCH UNIT 2 B 3.3-105 REVISION 0

4 ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE SR 3.3.5.1.5 REQUIREMENTS (cont?nued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LC0 3.5.1, LC0 3.5.2, LCO 3.7.2, LCO 3.8.1, and LCO 3.8.2 overlaps this Surveillance to complete testing of the assumed safety function.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply 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 that these components usually pass the Surveillance when performed at the 18 month Frequency.

M 3.3.5.1.6 De le+e l This S ures that the individual channel res se times are less th r equal to the maximum valu ssumed in the accident analys . Response time tes acceptance criteria are include '

Referenc .

ECCS RESPONSE TIME tests c cted on an 18 month STAGGERED TEST BASI is Frequen is consistent with the typical industr ueling cycle and is ed upon plant operating e ience, which shows that rando ilures of instru ation components causing serious respo time d ation, but not channel failure, are infrequent ccurrences.

f l (continued)

HATCH UNIT 2 B 3.3-133 REVISION O li i

- l 1

( ECCS Instrumentation B 3.3.5.1 l

BASES (continued)

REFERENCES 1. FSAR, Section 5.2. l

2. FSAR, Section 6.3.

3. FSAR, Chapter 15.

4. NEDC-31376-P, "Edwin I. Hatch Nuclear Power Plant, SAFER /GESTR-LOCA, Loss-of-Coolant Accident Analysis,"

December 1986.

5. NEDC-30936-P-A, "BWR Owners' Group Technical Specification Improvement Analyses for ECCS Actuation Instrumentation, Part 2," December 1988.

6. Tecnnis.1 ";;;ic;,T;rt; "::::1< 4 (p 1(. NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

r I

I L

RATCH UNIT 2 8 3.3-134 REVISION 0

i

- I g Primary Containment Isolation Instrumentation B 3.3.6.1 1

i BASES I

SURVEILLANCE SR 3.3.6.1.7 (continued) 1

{

REQUIREMENTS '

ISOLATION SYSTEM RESPONSE TIME acceptance criteria are l included in Reference 6. This test may be performed in one !

measurement, or in overlapping segments, with verification '

that all components are tested.

to the Surveillance states that the radiation Replace detecto be excluded from ISOLATION SYSTEM r E

' TIME testing. te is necessary becaus -

N8O difficulty of genera approprj le etector input ;

signal and because the princ detector operation thSeN virtually ensure an i neous re e time. Response times for radi

etector channels sha easured from ,

detector ut or the input of the first electron ent in the channel. !

ISOLATION SYSTEM RESPONSE TIME tests are conducted on an i 18 month STAGGERED TEST BASIS. This Frequency is consistent with the typical industry refueling cycle and is based upon l plant operating experience that shows that random failures '

of instrumentation components causing serious response time degradation, but not channel failure, are infrequent '

occurrences.

4 REFERENCES 1. FSAR, Section 6.3.

2. FSAR, Chapter 15.

3. FSAR, Section 4.2.3.4.2.

4. NEDC-31677P-A, " Technical Specification Improvement '

Analysis for BWR Isolation Actuation Instrumentation,"

July 1990. !

5. NEDC-30851P-A Supplement 2, " Technical Specifications Improvement analysis for BWR lsolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.

6. Technical Requirements Manual. ,

t

7. NRC No.93-102, " Final Policy Statement on Technical !

Specification Improvements," July 23, 1993.

S PJG D O - 32actl. Sy ste m An o /v se s for .

E l im i n 91. o n o f de le cfed Ef.5 pon.5 C 7,lne Teen 9 kmvi'e mcd s , h vvy ' 9 9 4.

HATCH UNIT 2 B 3.3-174 REVISION 0 1 l

6 Insert B '

A *te to the Surveillance states that the channel sensors are excluded from ISOLATION SYSTEM RESPONSE TIME testing. The exclusion of the channel sensors is supported by !

Reference 8 which indicates that the sensors' response times are a small fraction of the total response time. Even if the sensors experienced response time degradation, they would be expected to respond in the microsecond to millisecond range until complete failure.

J f

i i

f b

l l

i l

l

Insert C

.SR 3 5. l. l3 This SR ensures that the ECCS RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis. Response time testing acceptance criteria are included in Reference 14. A Note to the Surveillance states that the instrumentation portion of the response time may be assumed from established limits. The exclusion of the instmmentation from the response time surveillance is supported by Reference 15, which concludes that instrumentation will continue to respond in the micro second to millbecond range prior to complete failure.

The 18 month Frequency is based on the need to perform the Surveillance under the conditions that apply 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 that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

~

ECCS - Operating B 3.5.1 BASES i

i SURVEILLANCE SR 3.5.1.12 (continued)'

REQUIREMENTS )

i SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

1nser'$ l 5 REFERENCES 1. FSAR, Section 6.3.2.2.3.

2. FSAR, Section 6.3.2.2.4.

3. FSAR, Section 6.3.2.2.1.

4. FSAR, Section 6.3.2.2.2.

5. FSAR, Section 15.1.39.

6. FSAR, Section 15.1.40.

1

7. FSAR, Section 15.1.33.

8. 10 CFR 50, Appendix K.

9. FSAR, Section 6.3.3.

10. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Analysis,"

December 1986.

11. 10 CFR 50.46.

12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.

(NRC), " Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

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

4. mk.;e.oI Reavirement3 Maan i .

15 NE DO ~ h aa b y *hi c* t An*'YS'S f*r' GIem; noon d.5e.lected Resgoase Tim e Tes% 9 Ke9 aire m ents " %nv.y I%A.,

HATCH UNIT 2 B 3.5-16 REVISION 0

ML20082L333

Text

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