Text

Proposed Tech Spec Tables & Tech Specs Re Containment & Instrumentation
	ML20063M308
Person / Time
Site:	Brunswick
Issue date:	09/07/1982
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML20063M305	List: ML20063M304; ML20063M308;
References
	NUDOCS 8209100225
	Download: ML20063M308 (41)
	v • d • e

_ __

t m TABLE 3.3.5.2-1 h bN v 2 REMOTE SHUTDOWN MONITORING INSTRUMENTATION M >,3 .

O EO MINIMUM i READOUT QG CHANNELS ggu FUNCTIONAL UNIT AND INSTRUMENT NUMBER LOCATION OPERABLE

~ w as 8$ 1. Reactor Vessel Pressure RSP* 1

,,8g (C32-PI-3332 and C32-PT-3332)

EU 2. Reactor Vessel Water Level RSP* 1 i (B21-LT-N017D-3, B21-LSH-N017D-3)

(B21-L1-3331, B21-LI-R604AX, B21-LT-3331, B21-LT-N026A)

3. Suppression Chamber Water Level RSP* I 4

(CAC-LI-3342 and CAC-LT-3342) i ti o 4. Suppression Chamber Water Temperature RSP* I y (CAC-TR-778-7)

5. Drywell Pressure RSP* 1 (CAC-P1-3341 and CAC-PT-3341)

6. Drywell Temperature RSP* 1 (CAC-TR-778-1,3,4) 4

7. Drywell Oxygen Concentration Local Panel 1 i (CAC-AT-1259-2)

8. Residual Heat Removal Head Spray RSP* 1 g Flow (Ell-FT-3339 and Ell-FI-3339)

E g- 9. Residual Heat Removal System Flow RSP* 1 g (Ell-FT-3338, Ell-FI-3338, and Ell-FY ,3338) n

! g: 10. Residual Heat Removal Service Water RSP* 1 Discharge Differential Pressure (Ell-PDT-N002BX and Ell-PDI-3344) l

Remote Shutdown Panel, Reactor Building 20' Elevation

e N TABLE 4.3.5.2-1 e

h REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS -

9 CHANNEL CHANNEL

. FUNCTIONAL UNIT AND INSTRUMENT NUMBER CilECK CALIBRATION E

U 1. Reactor Vessel Pressure M Q s (C32-PI-3332 and C32-PT-3332)

2. Reactor Vessel Water Level (B21-LT-N017D-3, B21-ISH-N017D-3) NA Q (B21-LI-3331, B21-LI-R604AX, B21-LT-3331, M Q B21-LT-N026A)

3. Suppression Chamber Water Level M R (CAC-LI-3342 and CAC-LT-3342) u 2 4. Suppression Chamber Water Temperature M R u (CAC-TR-778-7)

L

5. Drywell Pressure (CAC-PI-3341 and CAC-PT-3341) M Q

6. Drywell Temperature (CAC-TR-778-1,3,4) M R

7. Drywell Oxygen Concentration (CAC-AT-1259-2) M Q I

8. Residual lleat Removal Head Spray Flow (El1-FT-3339 and E11-FI-3339) M Q

9. Residual Heat Removal System Flow M Q g (El1-FT-3338, E11-F1-3338, and E11-FY-3338) r*

h 10. Residual Heat Removal Service Water Discharge M Q g Differential Pressure (E11-PDT-N002BX .

n and Ell-PDI-3344) 5

g TABLE 3.3.5.2-1 Ei vs REMOTE SilUTDOWN MONITORING INSTRUMENTATION S -

@ MINIMUM i READOUT CitANNELS FUNCTIONAL UNIT AND INSTRUMENT NUMBER LOCATION OPERABLE gj

1. Reactor Vessel Pressure RSP* 1 (C32-PI-3332 and C32-Pf-3332)

2. Reactor Vessel Water Level RSP* 1 (B21-LT-N017D-3, B21-LSil-N017D-3)

(B21-LI-3331, B21-L1-R604AX, B21-LT-3331, B21-LT-N026A) )

3. Suppression Chamber Water Level RSP* 1 (CAC-LI-3342 and CAC-LT-3342)

El

4. Suppression Chamber Water Temperature RSP* 1 u (CAC-TR-778-7)

5. Drywell Pressure RSP* 1 (CAC-PI-3341 and CAC-PT-3341)

6. Drywell Temperature RSP* 1 (CAC-TR-778-1,3,4)

7. Drywell Oxygen Concentration Local Panel 1 (CAC-AT-1259-2)

8. Residual lleat Removal llend Spray RS P* 1

g. Flow (Ell-FT-3339 and Ell-FI-3339) n E. 9. Residual IIcat Removal System Flow RSP* 1 S (Ell-FT-3338, Ell-FI-3338, and Ell-FY-3338)

N 2: 10. Residual lleat Removal Service Water RSP* 1

? Discharge Differential Pressure (Ell-PDT-N002BX and Ell-PDI-3344)

,* Remote Shutdown Panel, Reactor Building 20' Elevation

i .

E! TABLE 4.3.5.2-1 M

h REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS o

CilANNEL CRANNEL FUNCTIONAL UNIT AND INSTRUMENT NUMBER CHECK CALIBRATION E

U l. Reactor Vessel Pressure M Q s2 (C32-PI-3332 and C32-PT-3332)

2. Reactor Vessel Water Level (B21-LT-N017D-3, B21-LSH-N017D-3) NA Q (B21-LI-3331, B21-LI-R604AX, B21-LT-3331, M Q B21-LT-N026A)

3. Suppression Chamber Water Level M R (CAC-LI-3342 and CAC-LT-3342)

{ 4. Suppression Chamber Water Temperature M R (CAC-TF-776-7)

I

$ 5. Drywell Pressure (CAC-PI-3341 and CAC-PT-3341) M Q

6. Drywell Temperature (CAC-TR-778-1,3,4) M R

7. Drywell Oxygen Concentration (CAC-AT-1259-2) M Q

8. Residual Heat Removal Head Spray Flow (Ell-FT-3339 and Ell-F1-3339) M Q

9. Residual Heat Removal System Flow g (Ell-FT-3338, E11-FI-3338, and E11-FY-3338) M Q B

g 10. Residual Heat Removal Service Water Discharge M Q m Differential Pressure (Ell-PDT-N002BX -

N and Ell-PDI-3344) 5

4 ATTACHMENT 2

SUBJECT:

Primary Containment Isolation Valves APPLICABILITY: Unit 2 O

DESCRIPTION OF TECHNICAL SPECIFICATION (TS) CHANGE: Technical Specification Table 3.6.3-1, Primary Containment Isolation Valves: (1) Revise the valve group number from 7 to 8 for the reactor vessel head spray isolation valves, and (2) revise the valve group number from 8 to 2 for the RHR discharge isolation valves to radwaste and the RHR process sampling valves.

BASIS FOR CHANGE: During the transition from custom Technical Specifications to Standard Technical Specifications for Brunswick, a typographical error occurred in the valve group number for the reactor vessel head spray isolation valves. The Brunswi-k-2 Technical Specifications correctly state that the reactor vessel head spr.e valves are in valve group 8. The Brunswick-1 reactor vessel head spray isolation valves also actually receive a group 8 signal; therefore, the valve group number for the Brunswick-1 valves should be revised to correctly show the actual plant design.

The RHR discharge isolation valves to radwaste and the RHR process sampling valves are actuated by isolation signals from either reactor vessel water level low 1 (actuates valve groups 2, 6, 7, and 8) or drywell pressure high j (actuates valve groups 2, 6, and 7). These valves are not actuated by a signal from reactor steam dome pressure high (actuates valve groups 7 and 8) as presently shown by TS Table 3.3.2-1. These valves are shown in the Technical Specifications as group 8 valves. To reflect the correct isolation l

actuations of the RHR discharge isolation valves to radwaste and the RHR process sampling valves, the valve group number for these valves should be revised from 8 to 2.

{

w TABLE 3.6.3-1 (Continued) v3 PRIMARY CONTAINMENT ISOLATION VALVES A

ISOLATION TIME VALVE FUNCTION VALVE GROUP I! (Seconds) g Reactor vessel head spray isolation valves 8 30

" Ell-F022 El1-F023 RilR shutdown cooling supply isolation valves 8 30 E11-F008 '

F11-F009 RilR injection isolation valves 8 30 Ell-F015A, B M RilR discharge isolation valves to radwaste 2 30 l

E11-F040 7 Ell-F049 t;

RilR process sampling valves 2 30 l EI1-F079A, B EIl-F080A, B NOTE 1: See Specification 3.3.2, Tabl.e 3.3.2-1 for isolation signal that operates each valve group.

N n

U W -

n L

ATTACHMENT 3

SUBJECT:

Primary Containment Isolation Valves APPLICABILITY: Unit 1 DESCRIPTION OF TECHNICAL SPECIFICATION (TS) CHANGE: Technical Specification Table 3.6.3-1, Primary Containment Isolation Valves: Revise the valve group number from 8 to 2 for the RHR discharge isolation valves to radwaste and the RHR process sampling valves.

BASIS FOR CHANGE: The RHR discharge isolation valves to radwaste and the RHR process sampling valves are actuated by isolation signals from either reactor vessel water level low 1 (actuates valve groups 2, 6, 7, and 8) or drywell pressure high (actuates valve groups 2, 6, and 7). These valves are net actuated by a signal from reactor steam dome pressure high (actuates valve groups 7 and 8) as presently shown by TS Table 3.3.2-1. These valves are shown in the Technical Specif'ications as group 8 valves. To reflect the correct isolation actuations of the RHR discharge isolation valves to radwaste and the RHR process sampling valves, the valve group number for these valves should be revised from 8 to 2.

g TABLE 3.6.3-1 (continued) en PRIMARY CONTAINMENT ISOLATION VALVES 6 ~

9 ISOLATION TIME VALVE FUNCTION VALVE GROUP I! (Seconds) s Reactor vessel head spray isolation valves 8 30

" l E11-F022 El1-F023 RilR shutdown cooling supply isolation valves 8 30 E11-F008 El1-F009 RilR injection isolation valves 8 30 Ell-F015A, B RilR discharge isolation valves to radwaste 2 30 m E11-F040 l h Ell-F049 RllR process sampling valves 2 30 l Ell-F079A, B Ell-F080A, B

}

NOTE 1: See Specification 3.3.2, Table 3.3.2-1 for isolation signal that operates each valve group.

n .

W I

ATTACHMENT 4

SUBJECT:

Fire Detection Instruments APPLICABILITY: Unit 1 and Unit 2 DESCRIPTION OF TECHNICAL SPECIFICATION (TS) CHANGE: Technical Specification Table 3.3.5.7-1, Fire Detection Instruments: Revise the minimum number of flame, heat, and smoke instruments required operable in their defined fire zones and add additional fire zones that have been established.

BASIS FOR CHANGE: Additional fire detection instruments have been installed or removed from fire zones listed in TS Table 3.3.5.7-1 and selected fire zones encompassed by Technical Specifications have been redefined to improve fire detection and response. The attached revisions to the minimum number of operable detectors are necessary to reflect these changes.

TABLE 3.3.5.7-1 FIRE DETECTION INSTRUMENTS INSTRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE FLAME HEAT SMOKE

1. Reactor Building #1 Zone 1 -17' 0 0 1 Zone 2 -17' 0 0 1 Zone 3 -17' O O 6 Zone 4 -17 0 0 6 Zone 5 20' 0 0 12 Zone 6 20' 0 0 11 Zone 7 20' 0 0 10 Zone 8 50' O 0 11 Zone 9 50' 0 0 15 Zone 10 80' 0 0 8 Zone 11 80' O O 10 Zone 12 98' 0 0 3 Zone 13 117' O O 1 Zone 14 117' 0 0 34 1 Zone 16 77' 0 0 4

2. Control Building Zone 1 70' 0 0 9 Zone 2 49' 0 0 4 Zone 3 49' 0 0 4, Zone 4 49' 0 0 13 Zone 5 49' O O 14 Zone 6 49' O O 6 Zone 7 23' 0 0 3 Zone 8 23' 0 0 3 Zone 9 23d 0 0 25 Zone 10 23' 0 0 24 Zone 11 23' O O 3 Zone 12 23' 0 0 3 Zone 13 49' 0 0 9 Zone 14 49' O O 9 Zone 15 70' 0 1 0 Zone 16 70' 0 1 0

3. Diesel Generator Building Zone 1 2' 0 0 25 Zone 2 2' 0 0 24 Zone 3 50' O O 9 Zone 4 23' 0 0 7 Zone 5 23' 0 0 5 Zone 6 23' 0 0 5 BRUNSWICK - UNIT 1 3/4 3-60 Amendment No.

TABLE 3.3.5.7-1 (Continued)

INSTRUMENT LOCATION MINIMUM INS CRUMENTS OPERABLE FLAME HEAT SMOKE

3. Diesel Generator Building (Cont'd)

. Zone 7 23' 0 0 5 Zone 8 23' 0 0 5 Zone 9 23' 0 0 8 Zone 10 50' O O 9

4. Service Water Building Zone 1 4' 0 0 7 Zone 2 20' 0 0 6

5. A0G Building Zone 1 20' 1 0 0 Zone 2 20' 1 0 0 Zone 3 20' 0 5 1 Zone 4 37' - 49' 2 5 0 BRUNSWICK - UNIT 1 3/4 3-61 Amendment No.

TABLE 3.3.5.7-1 FIRE DETECTION INSTRUMENTS INSTRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE FLAME HEAT SM0KE

1. Reactor Building #2

Zone 1 -17' 0 0 1 Zone 2 -17' 0 0 1 Zone 3 -17' O O 6 Zone 4 -17' O O 6 Zone 5 20' O O 12 Zone 6 20' O O 10 Zone 7 20' 0 0 9 Zone 8 50' 0 0 11 Zone 9 50' 0 0 15 Zone 10 80' 0 0 9 Zone 11 80' O O 10 Zone 12 98' 0 0 3 Zone 13 117' O O 1 Zone 14 117' 0 0 34 Zone 16 77' 0 0 4

2. Control Building Zone 1 70' O O 9 Zone 2 49' 0 0 4 Zone 3 49' 0 0 4 Zone 4 49' 0 0 13 '

Zone 5 49' 0 0 14 Zone 6 49' 0 0 6 Zone 7 23' 0 0 3 Zone 8 23' 0 0 3 Zone 9 23' 0 0 25 Zone 10 23' 0 0 24 Zone 11 23' O O 3 Zone 12 23' 0 0 3 Zone 13 49' O O 9 Zone 14 49' 0 0 9 Zone 15 70' O 1 0 Zone 16 70' O 1 0

3. Diesel Generator Building Zone 1 2' O 0 25 Zone 2 2' O O 24 Zone 3 50' 0 0 9.

Zone 4 23' O O 7 Zone 5 23' 0 0 5 Zone 6 23' 0 0 5 BRUNSWICK - UNIT 2 3/4 3-60 Amendment No.

1 1 _ _ - .

TABLE 3.3.5.7-1 (Continued)

INSTRUMENT LOCATION MINIMUM INSTRUMENTS OPERABLE FLAME HEAT SMOKE

3. Diesel Generator Building (Cont'd) o Zone 7 23' 0 0 5 Zone 8 23' 0 0 5 Zone 9 23' O 0 8 Zone 10 50' 0 0 9

4. Service Water Building Zone 1 4' 0 0 7 Zone 2 20' 0 0 6

5. A0G Building I

Zone 1 20' 1 0 0 Zone 2 20' 1 0 0 Zone 3 20' 0 5 1 Zone 4 37' - 49' 2 5 0 l

BRUNSWICK - UNIT 2 3/4 3-61 Amendment No.

ATTACHMENT 5

SUBJECT:

Safety / Relief Valves APPLICABILITY: Unit 1 and Unit 2 O

DESCRIPTION OF TECHNICAL SPECIFICATION (TS) CHANGE: Technical Specification 4.4.2: Revise the surveillance requirement for demonstrating Safety / Relief

' Valve (S/RV) operability.

BASIS FOR CHANGE: The three-stage S/RVs, which are equipped with bellows, have been replaced with two-stage S/RVs which are not equipped with bellows.

The existing Limiting Condition for Operation (TS 3.4.2) states that the safety function of all reactor coolant system S/RVs will be operable with their lift settings within i 1% of their specified values. The existing Surveillance Requirement (TS 4.4.2) requires the demonstration of S/RV operability through a bellows integrity check once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Since the two-stage S/RVs are not equipped with bellows, the surveillance requirements of TS 4.4.2 are no longer applicable nor can they be performed. The proposed Surveillance Requirement specifies that the S/RVs will be demonstrated operable pursuant to the requirements of TS 4.0.5. Therefore, the lift settings of all the S/RVs will be verified through surveillance testing of their associated solenoids in accordance with the requirements of the ASME Boiler and Pressure Vessel Code.

+

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of all reactor coolant system safety / relief valves shall be OPERABLE with lift settings within i 1% of the following values.*#

4 Safety-relief valves @ 1105 psig.

4 Safety-relief valves @ 1115 psig.

3 Safety-relief valves @ 1125 psig.

APPLICABILITY: CONDITIONS 1, 2, and 3.

ACTION:

a. With the safety valve function of one safety / relief valve inoperable, restore the inoperable safety valve function of the valve to OPERABLE status within 31 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With the safety valve function of two safety / relief valves inoperable, restore the inoperable safety valve function of at least one of the valves to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. With the safety valve function of more than two safety / relief valves l inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD l SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.2 The safety valve function of each of the above required safety / relief valves shall be demonstrated OPERABLE in accordance with the Surveillance Requirements of Specification 4.0.5.

The lift setting pressure shall correspond to ambient conditions of the valves at normal operating temperature and pressure.

From Spring, 1980 until the completion of the T-quencher modification, the safety-relief valve lift settings shall be arranged such that each safety-relief valve pair has a minimum nominal lift setting differential of 20 psi and shall be within 1% of the following values:

2 Safety-relief valves @ 1095 psig 3 Safety-relief valves @ 1105 psig 3 Safety-relief valves @ 1125 psig 3 Safety-relief valves @ 1125 psig BRUNSWICK - UNIT 1 3/4 4-4 Amendment No.

1

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of all reactor coolant system safety / relief valves shall be OPERABLE with lift settings within i 1% of the following values.*

4 Safety-relief valves @ 1105 psig.

4 Safety-relief valves @ 1115 psig.

3 Safety-relief valves @ 1125 psig.

t APPLICABILITY: CONDITIONS 1, 2, and 3.

ACTION:

l

a. With the safety valve function of one safety / relief valve inoperable, restore the inoperable safety valve function of the valve to OPERABLE l status within 31 days or be in at least HOT SHUTDOWN within the next l

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. With the safety valve function of two safety / relief valves inoperable, restore the inoperable safety valve function of at least one of the valves to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. With the safety valve function of more than two safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

[

SURVEILLANCE REQUIREMENTS 4.4.2 The safety valve function of each of the above required safety / relief l

l valves shall be demonstrated OPERABLE in accordance with the Surveillance Requirements of Specification 4.0.5.

I

The lift setting pressure shall correspond to ambient conditions of the valves at normal operating temperature and pressure.

l l

l BRUNSWICK - UNIT 2 3/4 4-4 Amendment No.

1 i

ATTACHMENT 6

SUBJECT:

Safety-Related Hydraulic Snubbers APPLICABILITY: Unit 1 and Unit 2 DESCRIPTION OF TECHNICAL SPECIFICATION (TS) CHANGE: Technical Specification Table 3.7.5-1, Safety-Related Hydraulic Snubbers: Revise the snubber list to reflect typographical corrections, snubber additions, and snubber deletions.

l BASIS FOR CHANGE: Due to plant modifications, certain system lines have been rerouted or removed which have necessitated the relocation and/or deletion of selected safety-related hydraulic snubbers. In addition, certain typographical errorsihave been identified in the snubber numbers and accessibilities/inaccessibilities listed in TS Table 3.7.5-1. The attached revised pages reflect the necessary corrections and revisions to TS Table 3.7.5-1.

l l

l 1

1

~

TABLE 3.7.5-1 (Continued)

E

@ S AFETY-RELATED IIYDRAULIC SNUBBERS

n SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON, LOCATION AND ELEVATION INACCESS IBLE ZONE ** TO REMOVE

@ Instrument Sensing System I y IB21-70lSS 164 Drywell 104' I No No ,

s 701SS 167 104' I No No 70lSS 169 100' I No No 70lSS 170 103' I No No 70lSS 171 99' I No No 70ISS 172 101' I No No 70lSS 175 100' I No No 70lSS 177 94' I No No 70lSS 178 97' I No No 70lSS 179 96' I No No u 70lSS 184 88' I No No D

y Reactor Closed Cooling Water System

,L 1RCC-32SS 30 Reactor Building 55' A No No 32SS 45 60' A No No 36SS 78 54' A No No 37SS 79 54' A No No 39SS80 59' A No No -

38SS 81 54' A No No 7sS 112 57' A No No 48SS 167 59' A ,

No No 48SS 168 58' A No No 48SS 169 60' A No No l 50SS 272 4' A No No i $ 60SS 121 Drywell 17' I No No

@ 60SS 122 16' I No No

@ 65SS 128 7' I No No 65SS 129 9' I No No 7ISS 139 9' I No No 73SS 145 5' I No No 19SS 157 21' I No No 19SS 160 29' I No No

TABLE 3.7.5-1 (Continued)

SAFETY-RELATED llYDRAULIC SNUBBERS * .

$ SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR llIGli RADIATION ESPECIALLY DIFFICULT

@ NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE n Reactor Core Isolation Cooling System M IE51-4SS45 Drywell 31' I No No 8

3SS46 39' I No No

@ 3SS47 39' I No No y 4SS66 39' I No No

- 4SS68 40' I No No 4SS69 40' I No No 4SS70 39' I No No 4SS71 36' I No No 4SS72 31' I No No 4SS73 30' I No No 41SS51 Reactor Building 40' A No No 42SS74 20' A No No

42SS75 20' A No No 42SS76 18' A No No R 42SS77 5' A No No

42SS78 0' A No No y 42SS79 4' A No No ,

g 42SS80 -13' A No No 42SS81 -16' A No No 42SS82 -9' A No No 40SS83 -9' A No No 40SS84 -9' A No No 40SS85 -12' A No No 40SS86 -9' A No No l 40SS87 -15' A No No 40SS88 -13' A No No 4ISS89 41' A No No l 41SS95 -41' A No No 19SSil3 -17' A No No b 19SSil4 -16' A No No

,k 49SS129 0' A No No a l

.if l

TABLE 3.7.5-1 (Continued) .

SAFETY-RELATED HYDRAULIC SNUBBERS

N

@ SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGil RADIATION ESPECIALLY DIFFICULT

@ NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE

~

Fuel Pool Cooling System l

@ IG41-ISS22 Reactor Building 12' A No No y ISS24 38' A No No y ISS30 38' A No No 12SS32 9' A No No 12SS33 9' A No No 15SS37 11l' A No No 20SS76 108' A No No 19SS79 89' A No No l 22SS85 108' A No No 12SS98 88' A No No 6SS111 88' A No No 7SS121 87' A No No R

c-SSS152 82' A No No y Reactor Recirculation System 5

IB32-SSAI Drywell 8' I No No SSBl 81' I No No i

SSA2 11' I No No SSB2 11' I No No ,

SSA3 11' I No No SSB3 1l' I No No SSA4 21' I No No SSB4 21' I No No SSA5 21' I No No SSB5 21' I No No

, SSA6 27' I No No R, SSB6 27' I No No

,9 SS89A 30' I No No

% SSB9B 30' I No No z SSA10 24' I No No

TABLE 3.7.5-1 (Continued) .

SAFETY-RELATED llYDRAULIC SNUBBERS

N SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGli RADIATION ESPECIALLY DIFFICULT h No. ON, LOCATION AND ELEVATION INACCESSIBLE 2.ONE*

TO REMOVE s I, p Reactor Recirculation System (Continued)

, SSAll Drywell (Cont'd) 11' I No No SSBil 11' I No No SSA12A 30' I No No H SSAl2B 30' I No No

- SSB12A 30' I No No SSBl?B 30' I No No Reactor Vessel Instrumentation IPS-3554 Drywell 32' I No No 3558 32' I No No 3561 32' I No No 3562 60' I No No 3567 63' I No No 3570 32' I No No g 3613 32' I No No p 3617A 32' I No No 3617B 32' I No No 4 3751 34' I No No e 3752 34' I No No Reactor Feedwater System IB21-2SS3 Drywell 38' I No No l

2SS4 56' I No No 3SS6 41' I No No 3SS9 39' I No No 3SSil 41' I No No 3SS12 40' I No No 3SS13 61' I No No SSS17 38' I No No g SSS18 56' I No No a

n

TABLE 3.7.5-1 (Continued) .

SAFETY-RELATED llYDRAULIC SNUBBERS

E z SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT h NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE

a

i Reactor Feedwater System (Continued) 6SS20 Drywell (Cont'd) 41' I No No

@ 6SS23 39' I No No U 6SS25 41' I No No s 6SS26 40' I No No 6SS27 63' I No No ISS227 34' I No No ISS228 38' I No No 2SS229 53' I No No 2SS230 62' l I No No 3SS231 40' I No No 3SS232 36' I No No 3SS233 40' I No No 3SS234 48' I No No M 3SS235 63' I No No 4SS236 34' I No No Y 4SS237 38' I No No

$ SSS238 53' I No No SSS239 61' I No No 6SS240 41' I No No 6SS241 36' I No No 6SS242 39' I No No-6SS243 48' I No No 6SS244 61' I No No '

Residual Heat Removal System IEll-90SS267 Drywell 79' I No No ,

90SS268 86' I No No g 90SS271 36' I No No a 90SS274 938 I No No

' E. 90SS275 93' I No No N 90SS277 96' I No No >

S 90SS278 96' I No No x 90SS280 , 101' I No No

- 90SS281 93' I No No 90SS282 10l' I No No

\

t

4 TABLE 3.7.5-1 (Continued) .

s SAFETY-RELATED llYDRAULIC SNUBBERS

3.

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGli RADIATION ESPECIALLY DIFFICULT y NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE '

M M l Residu'ai lleat Removal System (Continued) 8 IEll-132SS264 Reactor Bldg. 31' A No No

@ 21SS296 (Cont'd) 39' A Fo No y 21SS297 39' A No No

- 47SS323 42' A No No 47SS326 42' A No No i

47SS328 42' A No No 49SS330 42' A No No 49SS331 42' A No No 49SS333 42' A No No 49SS334 43' A No No 49SS336 40' A No No 128SS355 42' A No No 49SS359 42' A No No R*

127SS376 59' A No No 128SS387 43' A N, No y 2SS396 5' A No No O 2SS397 3' A No No SSS398 -41' A No No 2SS399 -3' A No No SSS400 -3' A No No 4SS401 -12' A No No SSS402 -12' A No No 3SS403 -11' A No No 6SS404 -12' A No No 8SS405 -14' A No No 6SS406 -14' A No No g 8SS407 -15' A No No o 12SS408 -14' A No No E I16SS409 -9' A No No l

@ 113SS410 -9' A No No R 9SS411 -14' A No No z 109SS412 -14' A No No

. 2SS413 0' A No No 132SS414 43' A No No 1

TABLE 3.7.5-1 (Continued) ,

SAFETY-RELATED !!YDRAULIC SNUBBERS

to SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT h

en NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE 5 l 9 Residual Heat Removal System (Continued)

IEll-39SS460 Reactor BIdg. 10' A No No g 89SS461 (Cont'd) 6' A No No A 53SS462 15' A No No 53SS463 14' A No No 53SS464 14' A No No 53SS465 14' A No No 53SS466 14' A No No 50SS467 14' A No No 50SS468 17' A No No 18SS469 53' A No No 18SS470 43' A No No 89SS480 67' A No No 89SS487 67' A No No u 89SS489 67' A No No D 89SS491 67' A No No u 91SS499 69' A No No b

91SS500 57' A No No 56SS504 14' A No No 56SS505 7' A No No 56SS506 3' A No No 56SS507 3' A No No 56SS508 4' A No No 46SS509 8' A No No 46SS510 11' A No No 46SS511 10' A No No 46SS512 -l' A No No 58SS514 14' A No No N 49SS515 37' A No No E 49SS516 37' A No No 8~ 56SS517 -5' A No No l E

51SS546 32' A No No 51SS547 28' A No No f 115SS549 31' A No No

TABLE 3.7.5-1 (Continued)

I g S AFETY-RELATED HYDRAULIC SNUBBERS *

o h SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT ON, LOCATION AND ELEVATION

$ NO. INACCESSIBLE ZONE ** TO REMOVE 9 i

, Residual Heat Removal System (Continued) g IEl1-54SS 551 Reactor Bldg. 3)' A No No a 54SS 552 (Cont'd) 28' A No No H 98SS 554 32' A No No

~ 58SS 563 7' A No No 58SS 565 13' A No No ,

58SS 566 6' A No No 107SS 573 15' A No No i 69SS 574 6' A No No ;

91SS 575 53' A No No 68SS 577 S' A No No 53SS 596 14' A No No r j 50SS597 26' A No No !

w 3 Service Water System u ISW-133SS 22 Reactor Building -6' A No No b

110SS 35 -5' A No No 174SS 70 42' A No No 173SS 72 14' A No No 142SS 74 40' A No No 142SS 75 40' A No No 142SS 82 70' A No No l

140SS86 45' A No No .

153SS 102 44' A No No !

173SS 110 48' A No No I 173SS 114 70' A No No 103SS 117 41' A No No I !

l g 103SS 121 38' A No No !

l g 103SS 126 60' A No No g 103SS 127 57' A No No

! O , l n

e l

l

TABLE 3.7.5-1 (Continued) .

, S AFETY-RELATED HYDRAULIC SNUBBERS * ,

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HICH RADIATION ESPECIALLY DIFFICULT h NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE

~

9-

, Service Water System (Continued) I h ISW-106SS 131 Reactor Bldg. 60' A No No H 100SS145 (Cont'd) 59' A No No H 100SS 149 60' A No No 106SS 151 59' A No No 106SS 156 60' A No No 142SS 163 60' A No No 142SS 164 8' A No No 142SS 165 8' A No No 175SS 166 42' A No No 140SS 167 42' A No No 142SS 168 58' A No No 142SS 169 71' A No No M" 174SS 174 42' A No No i 173SS 175 30' A No No Y

133SS 177 -5' A No No d 100SS 193 62' A No No 138SS 210 56' A No No !

106SS 211 60' A No No 106SS 212 60' A No No 106SS 213 59' A No No 106SS 214 60' A No No ,

127SS215 17' A No No 123SS 216 17' A No No N

9 B

n W

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

i 4

b l

to I

TABLE 3.7.5-1 i, E m

i

$ SAFETY-RELATED HYDRAULIC SNUBBERS * .

O i SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT I

l 4

h g

NO. ON, LOCATION AND ELEVATION INACCESS IBLE ZONE ** TO REMOVE -

u Core Spray System 2E21-2SS 32 Reactor Building 66' A No No

.25SS 91 -6' A No No l 2SS I6 0' A No No 2SS 17 13' A No No

. 1SSS19 -3' A No No ,

15SS 20 -3' A No No 28SS 23 -4' A No No 25SS 96 -6' A No No !

, g 40SS 106 -12' A No No ,

j g 40SS 107 -12' A No No

y 39SS 108 -12' A No No L

39SS 109 -12' A No No

2SS 31 68' A No No 6SS41 70' A No No 4

6SS 42 69' A No No

2SS 18 14' A No No

! 2E21-3SS 46 DryweII 63' I No No l

3SS 47 63' I No No 1 3SS 48 65' I No No l

3SS49 66' I No No l'

7SS 53 63' I No No i

7SS 54 63' I No No ,

g 7SS 55 65' I No No g 7SS 56 66' I No No i

,8,

. i I

l i

E TABLE 3.7.5-1 (Continued)

E D SAFETY-RELATED HYDRAULIC SNUBBERS

M SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION SNUBBER ESPECIALLY DIFFICULT

' NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE N

M Reactor Water Cleanup System n 2G31-ISS3 Drywell 54' I No No l

Condensate Drains System 2B21-51SS103 Drywell 29' I No No

, 51SS105 26' I No No 51SS106 18' I No No 50SS109 31' I No No 50SS111 28' I No No 51SSil3 23' I No No SISS115 24' I No No t' SISS118 24' I No No n

Y High Pressure Coolant Injection System U

2E41-4SS44 Drywell 40' I No No 4SS45 35' I No No 4SS47 40' I No No 4SS49 37' I No No 4SS50 40' I No No .

4SSSI 30' I No No N

B n

tn TABLE 3.7.5-1 (Continued) a g SAFETY-RELATED liYDRAULIC SNUBBERS *

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT h NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE s

to Nuclear Steam Vent System 2B21-44SS129 Drywell 104' I No No 44SS131 93' I No No 44SS134 99' I No No 44SS136 97' I No No 44SS137 96' I No No 44SS138 95' I No No 44SS141 87' I No No 44SS142 87' I No No 44SS143 87' I No No R 44SS146 87' I No No o 46SS147 82' I No No y 44SS149 85' I No No g 44SS150 83' I No No 47SS155 75' I No No 47SS156 78' I No No 47SS157 75' I No No Standby Liquid Control System

^

! 2C41-9SS4 Drywell 63' I No No l 9SSS 47' I No No 9SS8 42' I No No 9SS10 38' I No No fR 9SS11 9SS12 39' 69' I I No No No No g 9SS13 52' I No No g 441-9SS26 Reactor Building 72' A No No 2 9SS27 72' A No No i

,o 6SS34 84' A No No l

l l

TABLE 3.7.5-1 (Continued)

SAFETY-RELATED llYDRAULIC SNUBBERS * .

s:

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

l TO REMOVE f

H Reactor Feedwater System l

w 2B21-2SS3 Drywell 38' I No No 2SS4 56' I No No 3SS6 41' I No No 3SS9 39' I No No 3SS11 41' I No No 3SS12 40' I No No 3SS13 61' I No No 5SS17 38' I No No SSS18 56' I No No 6SS20 41' I No No R#'

6SS23 39' I No No 6SS25 41' I No No Y 6SS26 40' I No No M 6SS27 63' I No No ISS227 34' I No No ISS228 38' I No No 2SS229 53' I No No 2SS230 62' I No No N

9 4

rt

.F

$ TABLE 3.7.5-1 (Continued) -

E S AFETY-RELATED llYDRAULIC SNUBBERS * ,

M SYSTEM SNUBBER INSTALLED SNUBBER ACCESSIBLE OR HIGil RADIATION ESPECIALLY DIFFICULT

' NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE E

y Residual lleat Removal System (Continued) u 2 Ell-60SS 440 Reactor Bldg. (Cont'd) 13' A No No 65SS441 3' A No No 65SS 442 3' A No No 60SS443 11' A No No 73SS 444 2' A No No 21SS445 5' A No No 68SS 448 13' A No No 75SS 449 2' A No No u 61SS 450 7' A No No 2 60SS451 13' A No No y 60SS 452 13' A No No 4

60SS 453 10' A No No 60SS 454 10' A No No 89SS459 11' A No No 89SS 460 10' A No No 89SS 461 6' A No No 53SS 462 15' A No No 53SS 463 14' A No No 53SS 464 14' A No No -

53SS 465 14' A No No 53SS 466 14' A No No 50SS 467 14' A No No l

50SS 468 17' A No No 56SS504 14' A No No 8, 56SS 505 7' A No No g 56SS 506 3' ,

A No No S 56SS 507 3' A No No z 56SS 508 4' A No No

,o 46SS 509 8' A No No 46SS 510 11' A No No

e h TABLE 3.7.5-1 (Continued) s g

SAFETY-RELATED llYDRAULIC SNUBBERS

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGli RADIATION ESPECIALLY DIFFICULT NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE o Service Water System 2SW-133SS22 Reactor Building -6' A No No 110SS35 -5' A No No 173SS72 14' A No No 142SS164 8' A No No

, 142SS165 8' A No No 133SS176 14' A No No 133SS177 -5' A No No 142SS74 40' A No No u 142SS75 40' A No No 5 140SS86 45' A No No l a 153SS102 44' A No No b 153SS109 44' A No No 173SS110 48' A No No 153SS115 44' A No No 103SS121 38' A No No 140SS167 42' A No No 173SS175 30' A- No No 142SS82 70' A No No .

173SSI14 70' A No No 103SS126 60' A No No N

8 B,

e

g TABLE 3.7.5-1 (Continued) x g S AFETY-RELATED HYDRAULIC SNUBBERS * ,

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR llIGli RADIATION ESPECIALLY DIFFICULT 8

NO. ON, LOCATION AND ELEVATION INACCESS IBLE ZONE ** TO REMOVE E

y Service Water System (Continued) u 2SW-103SS 127 Reactor Bldg. (Cont'd) 57' A No No 106SS 131 60' A No No 100SS 145 59' A No No 100SS 149 60' A No No 106SS 151 59' A No No 174SS70 42' A No No 106SS 156 60' A No No 142SS 168 58' A No No 142SS 169 71' A No No t' 100SS 193 62' A No No !

106SS 211 60' A No No Y 106SS 212 60' A No No b' 106SS 213 59' A No No 106SS 214 60' A No No 103SS 117 41' A No No 142SS 163 O A No No '

175SS 166A 42' A No No 175SS 166B 42' A No No 109SS 173 14' A No No -

174SS 174 43' A No No N

9 B -

n

en TABLE 3.7.5-1 (Continued) b 0; SAFETY-RELATED llYDRAULIC SNUBBERS * '

d Q SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGli RADIATION ESPECIALLY DIFFICULT i NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE Steam Relief Discharge System (Continued)

H 2B21-33SS251 DryweII (Cont'd) 32' I No No 20SS256 44' I No No l 20SS257 37' I No No 20SS258 36' I No No 20SS260 18' I No No 20SS261 20' I No No 20SS262 15' I No No 27SS264 10' I No No 27SS266 17' I No No w 27SS267 18' I No No 2 27SS269 36' I No No y 27SS270 44' I No No d, 27SS272 42' I No No

" 58SS275 18' I No No 58SS276 30' I No No 58SS277 31' I No No 58SS279 36' I No No 58SS280 37' I No No 58SS281 38' I No No -

12SS286 39' I No No 12SS287 39' I No No 12SS288 44' I No No 12SS289 39' I No No

@ 12SS290 43' I No No y 12SS292 42' I No No y 34SS296 35' ,

1 No No y 34SS297 35' I No No

" 34SS298 44' I Sio No

@ 34SS299 39' I No No 34SS300 44' I No No

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

a ATTACHMENT 7 1

SUBJECT:

Primary Containment Integrity APPLICABILITY: Unit I and Unit 2 DESCRIPTION OF TECHNICAL SPECIFICATION (TS) CHANGE: TS 4.6.1.1: Revise the Surveillance Requirement and associated footnote relative to the 31-day interval primary containment integrity demonstration.

l BASIS FOR CHANGE: The existing TS 4.6.1.1 requires that primary containment integrity be demonstrated on a 31-day basis by the verification of (1) containment penetrations being closed and (2) equipment hatches being sealed. Such verification is not required for penetrations which are locked or secured inside primary containment or for penetrations verified as closed during the previous 92 days. The current GE/BWR-4 Standard Technical Specifications (STS) now specify surveillance requirements for penetrations rather than providing separate surveillance requirements for containment equipment hatches. Therefore, the deletion of TS 4.6.1.1.a.2 is consistent with this GE/STS guidance. The proposed change concerning high radiation areas provides an exception to the verification requirements of TS 4.6.1.1 for containment penetrations which are inaccessible due to high radiation. The proposed change concerning " COLD SHUTDOWNS greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months " in the f ootnote for TS 4.6.1.1 is being provided to establish a criteria for allowing sufficient time for plant personnel to perform the specified surveillance. In addition, the footnote and associated reference to Special Test Exception 3.10.1 are being deleted because this test exception is associated with intial startup testing and is no longer required. Carolina Power & Light Company plans to request the deletion of Special Test Exception 3.10.1 in a license amendment submittal to be made in the near future.

o 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY LIMITING CORDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: CONDITIONS 1, 2, and 3.

l ACTION:

l Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY

! within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in

. COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l SURVEILLANCE REOUIREMEhTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a. At least oneg per 31 days by verifying that all primary containment penetrations not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except as provided in ,

Table 3.6.3-1 or Specification 3.6.3.

b. By verifying each primary containment air lock OPERABLE per l Specification 3.6.1.3. +

c. By verifying the suppression pool OPERABLE per Specification 3.6.2.2.

l l l l

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment or high radiation areas, and are locked, l sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months l except such verification need not be performed more of ten than once per 92 days.

BRUNSWICK - UNIT 1 3/4 6-1 Amendment No.

9 o

3/4.6 CONTAIhMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAIhMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAIhMENT INTEGRITY shall be maintained.

APPLICABILITY: CONDITIONS 1, 2, and 3. l ACTION:

Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAIhMENT INTEGRITY within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a. At least oneg per 31 days by verifying that all primary containment penetrations not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except as provided in Table 3.6.3-1 or Specification 3.6.3.

b. By verifying each primary containment air lock OPERABLE per j Specification 3.6.1.3. ,

c. By verifying the suppression pool OPERABLE per Specification 3.6.2.2.

s I

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment or high radiation areas, and are locked. l sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months l except such verification need not be performed more of ten than once per 92 days.

BRUNSWICK - UNIT 2 3/4 6-1 Amendment No.

?-

e ATTACHMENT 8

SUBJECT:

Isolation Actuation - Condenser Vacuum APPLICABILITY: Unit 1 and Unit 2 DESCRIPTION OF TECHNICAL SPECIFICATION (TS) CHANGE: TS Table 3.3.2-1 and TS Table 4.3.2-1: Revise the applicable operation condition footnote for Item 1.e, Condenser Vacuum-Low, to allow the isolation function logic to be bypassed if all turbine stop valves are closed.

BASIS FOR CHANGE: The proposed TS change reflects actual plant logic which allows the low condenser vacuum isolation function to be bypassed if the trubine stop valves are closed. The current Brunswick TS require the low condenser vacuum isolation of the main steam isolation valves to be operable in both the Run and Startup Condition if vessel pressure is greater than or equal to 500 psig.

This request is based upon documented test results indicating that instability problems encountered at older General Electric BWRs were not encountered during the Startup Condition of the BWR/4 design (refer to General Electric SIL No. 107 dated October 31, 1974). This change is also based on guidance as setforth in the current GE/BWR-4 Standard Technical Specifications.

t

e e

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTIONS ACTION 20 - Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

. ACTION 21 - Be in at least START-UP with the main steam line isolation valves closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 22 -

Be in at least START-UP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l ACTION 23 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within one hour.

ACTION 24 -

Isolate the reactor water cleanup system.

ACTION 25 -

Close the affected system isolation valves and declare the affected system inoperable.

ACTION 26 - Verify power availability to the bus at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 27 - Deactivate the shutdown cooling supply and reactor vessel head spray isolation valves in the closed position until the reactor steam dome pressure is within the specified limits.

NOTES l

When handling irradiated fuel in the secondary containment.,

a. See Specification 3.6.3.1, Tchle 3.6.3.1-1 for valves in each valve group.

b. A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for

! required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

c. With only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In thcce cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.

d. Trips the mechanical vacuum pumps.

e. A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE.

f. May be bypassed with all turbine stop valves closed.

g. Closes only RWCU outlet isolation valve. l

h. Alarm only.

l BRUNSWICK - UNIT 1 3/4 3-16 Amendment No.

o e

9 TABLE 4.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS tn N CllANNEL OPERATIONAL 9 CilANNEL FUNCTIONAL CHANNEL CONDITIONS IN WFIICil i TRIP FUNCTION AND INSTRUMENT NUMBER C11ECK TEST CALIBRATION SURVEILLANCE REQUIRED E

y 1. PRIMARY CONTAINMENT ISOLATION (Continued)

d. Main Steam Line Tunnel Temperature - liigh NA M R 1, 2, 3 (B21-TS-N010A,B ,C,D; B21-TS-N0llA,B ,C,D; B21-TS-N012A,B,C,D and B21-TS-N013A,B,C, D)

e. Condenser Vacuum - Low (B21-PS-N056 A,B,C,D) NA M R 1, 2f

f. Turbine Building Area Temp - liigh NA M R 1, 2, 3 u, (B21-TS-3225A,B,C,D; B21-TS-3226A,B,C,D; B21-TS-3227A,B,C,D; B21-TS-3228A,B,C,D; 3: B21-TS-3229A,B,C,D; B21-TS-3230A,B,C,D; B21-TS-3231A,B,C,D and B21-TS-3232A,B,C D)

Y O

N 8

k # When any turbine stop valve is open. l

.is

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTIONS ACTION 20 - Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 21 - Be in at least START-UP with the main steam lime isolation valves closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 22 - Be in at least START-UP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 23 - Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within one hour.

ACTION 24 - Isolate the reactor water cleanup system.

ACTION 25 - Close the affected system isolation valves and declare the affected system inoperable.

ACTION 26 - Verify power availability to the bus at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 27 - Deactivate the shutdown cooling supply and reactor vessel head spray isolation valves in the closed position until the reactor steam dome pressure is within the specified limits.

NOTES

When handling irradiated fuel in the secondary containment'.

a. See Specification 3.6.3.1, Table 3.6.3.1-1 for valves in each valve group.

b. A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

c. With only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.

d. Trips the mechanical vacuum pumps.

e. A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE.

f. May be bypassed with all turbine stop valves closed. l

g. Closed only RWCU outlet isolation valve.

h. Alarm only.

BRUNSWICK - UNIT 2 3/4 3-16 Amendment No.

ML20063M308

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