ML19312D689

Amends 26 & 50 to Licenses DPR-71 & DPR-62,respectively, Correcting Table of safety-related Hydraulic Snubbers, Providing for Systematic Implementation of Instrumentation of Mods & Elimination Shorting Links Requirement

ML19312D689
Person / Time
Site:	Brunswick
Issue date:	03/14/1980
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML19312D690	List: ML19312D689 ML19312D702
References
NUDOCS 8003250197
Download: ML19312D689 (58)

Text

e 4

M UNITED STATES

(

[

g NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555

'5 ej A'4*** /

CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 26 License No. DPR-71

l.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Carolina Power Light Company dated February 20, 1979, as supplemented January 14, 1980, and applications dated November 19, 1979 and January 24, 1980, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's' rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the applfeations, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public, and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

.2.

Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment

- and paragraph 2.C.(2) of Facility Operating License No. DPR-71 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B as revised through Amendment No. 26, are hereby incorporated in the license.

The licensee shall operate the facility in at ordance with the Technical Specifications.

8003250 53 7

3.

Th.is license;angndment -1s effecti:ye at of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION M

Thomas

. Ippolito, Chief

- Operating Reactors Branch #3 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications

-Date of Issuance:

March 14, 1980 t

i I

l c

ATTACHMENT TO LICENSE AMENDMENT NO. 26 FACILITY OPERATING LICENSE NO DPR-71 DOCKET NO 50-325 Remove the following p. ages and replace with identically numbered pages,

~3/4 3-1/3/4 3-2 3/4 3-5/3/4 3-6

.3/4 7-11/3/4 7-12 3/4 7-13/3/4-7-14 3/4'7-15/3/4 7-16 3/47-17/3/47-18 3/4 7-19/3/4 7-20 3/4 7-21/3/4 7-22 3/4 7-23/3/4 7-24 3/4 7-25/3/4 7-26 3/47-27/3/47-28 3/4 7-29/3/4 7-30 3/4 9-3/3/4 9-4 The underlined pages are overleaf pages and are' provided for convenience.

i

. ~., - -,

+=w 3/4.3 -INSTRUMENTATION 3/4. 3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

Set points and interlocks are given in Table 2.2.1-1.

APP'LICABILITY_: As shown Lin Table 3.3.1-1.

ACTION:

With the requirements for the minimum number of OPERABLE channels a.-

not satisfied for one trip system, place that trip system in the tripped condition within one hour or take the ACTION required by Table 3.3.1-1.

With the requirements for the minimum number of OPERABLE channels b.

not satisfied for both trip systems, take the ACTION required by

~

23.;

Table 3.3.1-1.

The provisions of Specification 3.0.3 are not applicable in c.

OPERATIONAL CONDITION 5.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system.

-The REACTOR PROTECTION SYSTEM RESPONSE tit 1E of each reactor trip 4.3.1.3 function of Table 3.3.1-2 shall be demonstrated to be within its limit at Each test.shall include at least one logic

-least once per 18 months.

train such that both logic trains aro-tested at least once per 36 months and one channel per function such that all -channels are tested at least once every N times 18 months where.N is the total number.of redundant

~

channels in a specific reactor trip function.

~

- 8RUNSWICK-UNIT l' 3/4 3-1

i.

e TABLE 3.3.1-1_

jg REACTOR PROTECTION-SYSTEM INSTRUMENTATION w'

'C APPLICABLE MINIMUM NUMBER-i k'o OPERATIONAL OPERABLE CHANNELS X

FUNCTIONAL UNIT AND INSTRUMENT NUMBER CONDITIONS PER TRIP SYSTEM (a)

ACTION l 8

g 1.

Intemediate Range Monitors:

p (C51-IRM-K601 A.B.C D,E,F,G,H) 3 1

l

a.. Neutron Flux - High 2',5(b)

^

3, 4 2

2 b.

Inoperative 2, 5 3

1 3, 4 2

2

2..

Average Power Range Monitor:

(C51-APRM-CH.A,8,C,0,E,F) a.

Neutron Flux - High,15%

2,5(b) 2 3

l

~

~.

'f b.

Flow Biased Neutron Flux -

'4 High 1

2 N

c.

Fixed Neutron Flux-High, 120%

1 2

4 d.-

Inoperative 1,2,5 2

5 e.

Downscale 1

2 4

f.

LPRM.

1, 2, 5

- (c)

NA 3.

Reactor Vessel Steam Dome Pressure -

High (821-PS-N023 A,B.C,D)

-1,2(d) 2 6

4.

Reactor Vessel Water Level -

'2 Low, Level #1 (B21-LIS-N017 A,B,C,0) 1, 2 2

6 k

-5.

Main Steam Line Isolation Valve -

3 Closure (B21-F022 A,B,C,0 and B21-F028 A B.C,0) 1 4

4 6.

Main Steam Line Radiation - High 1,2(d) 2 7

(012-FM v603 A,B,C.D) cn 1: g k

.~

a

,=

TABLE 3.3.3-1 (Continued)'

MN REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION 10 - In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In OPERATIONAL CONDITION 3 or 4, lock the reactor mode switch in the Shutdown position within one hour.

In OPERATIONAL CONDITION 5 suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods within one hour.

TABLE NOTATIONS.

a.

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

The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn

• and shutdown margin demonstrations.

c.

An APRM channel is inoperable if there are less than 2 LPRM inp'uts

~

per level or less than eleven LPRM inputs to an APRM channel.

d.

These functions are not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed.

e.

This function is not required to be OPERABLE when PRIMARY CONTAIN-MENT INTEGRITY is not required, f.

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2..

g '.

These functions are bypassed when THERMAL POWER is less than 30%

of RATED THERMAL POWER.

• Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

l lE

(-

- BRUNSWICK - UNIT 1,

3/4 3-5 Amendment No.

26

' ' ~ ~

l 1

1 TABLE 3,3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES I uptl10hAI. IINIT AND INSTRUMENT NUH0ER RESPONSE TIME

'~] Seconds )"-

4

• i '

l.

Intermediate Range Monitors (C51-lRM-K601 A,B,C,D.E.F,G,ll):

a.

Neutron Flux - liigh*

NA b.

Inopera tive.

NA 2.

Average Power Range Monitor * (C51-APRM-Cll.A,B,C,D,E F):

{

a.

neutron Flux - liigh,15%

.< 0.09

1..

I' low Blased Heutron Flux - High

~

flh c.

Neutron Flux - High,120%

< 0.09 it.

Inoperatlye Rh c.

Downscale NA I.

LPRM NA 3.

Reactor Vessel Steam Dome Pressure - High (B21-PS-N023 A.B.C.D) 1 55 0

o, Y-4.

Itcactor Vessel Water Level - Level #1 (B21-LIS;H017 A.B.C D) 1 1

05

t.. Main Steam Line isolatior, Valve-Cicsure (B21-F022 A,B,C,0 and D21-F028 A,B,C,0) < 0.06 i

(,. H.iin Steam Line Radiation - liigh (012-RM-K603 A,B,C,0)

NA 7.

lirywell Pressure - liigh (C71-PS-H002 A,B,C,0)

HA l

n. scram Discharge Volume Water Level - High (C11-LSil-N013 A,B,C,0)

HA l

'l. ' lin hine Stop Valve - Closure (EllC-SVOS-1X 2X,3X,4X)

< 0.06

'i 10 lin hino Control Valve Fast Closure.

g'{

control 011 Pressure - Low (EllC-PSL-1756,1757,1758,1759)

< 0.08 l

11. Itcactor Mode Switch in Shutdown Position (C71A-S1)

NA l

a

12. Manual Scram (C71A-53 A,B)

NA l

59 iiiiiilrnn'iletectors are exempt from response time testing. Response time shall be measured from i

delet. Lor output or input of first electronic component in channel.

~

j cA o

q-

=~

Ww oJ

g TABLE 3.7.5-1 Cj SAFETY RELATED HYDRAULIC SNUDDERS*

R 7t i

SNUBBER SYSTEM SNUBBER ~ INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT.

c-NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE-(A or I)

(Yes or No) '

(Yes or No)

~

Core Spray System 1E21.-2SSl6 Reactor Buildino O'

A-No No~

2SS17 13' A

No No 2SS18

.14' A

No No 15SS19

-3'

'A No No 15SS20

-3' A

No No 1

28SS23

-4' A

No No l

2SS31 68' A-No No 2SS32-66' A

No No w2 6SS41 70' A

.No No 6SS42 69' A

No No y

.25SS91-

-6' A

No No 25SS96

-6' A

No No 40SS106

-12' A

No No i

40SS107

-12' A

No No 39SS108

-12' A

No No 30SS109

-12' A

No No 3SS46 Drywell 63' I

No No.

l 3SS47 63' I

No No 3SS48 65' I

No No i

3SS49 66' I

No No 7SS53 63' I

No No i

7SS54 63' I

No No 75S55 65' I

No No

[

75S56 66' I

No No Reactor Water Cleanup System 1G31-lSS3 Drywell 54' I

No No Y

Es t

j.

TABLE 3.7.5-1-'(Continued)

,:o SAFETY RELATED HYDRAULIC SNUBBERS

• E SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION

'ESPECIALLY DIFFICULT-7 NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE e

(A or I)

(Yes or No)

(Yes or No),

2 Condensate Drain System '

.4 1B21-5155103 Drywell -

29' '

I No No l.

5155105 29' I

No

~No 51SS106 26' I

No-no

-5055109 18' I

No No 505S111 31' I

No No 515S113 28' I

No No-i 515S11S 23' I

No No 515S118 24' I

No No Control Rod Drive Systen R

1C11-165S1 Drywell 69' I

No No i

16SS6 63' I

. No No j

?

16S57 69'-

I No No 16S58

.70' I

No.

No 165S10 72' I

No No 165S11 72' I

No-No j

16SS12 72' I

No No High Pressure Coolant Injection Systen 1E41-4SS44

-Drywell 40' I

No No 4SS45 35' I

No No 4SS47 40' I

No No 45549 37' I

No No 45550 50' I

No No 45551 30' I

No No k

60559 Reactor Building

.4' A

No No i

R 65527

-5 A

No No 2

65S28 l'

A No-No s

E 6S530

-l' A

No No 65532

-5' A

No No

=

65533 l'

A No No 65535

-l' A

No No m

65S36

-5 A

No No 1

i g

TABLE 3.7.5-1 (Continued)

SAFETY RELATED HYDRAULIC SNUBBERS

• A SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT

- NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE

E

'(A or 1) -

(Yes or N0)

(Yes or No)

G High Pressure Coolant Injection System-(Cont'd) m 1E41-6SS37 Reactor Building O'

A No-No j

6SS38 (Cont'd)

.l' A

No No l 4

6SS40 2'

A No No 6SS4.2

-4' A

No No 60SS52 A

No No 6SS64

-1 A

No No 61SS71

-4 A

No No 61SS72

-1 A

No No w1 61SS73

-2 A

No No 61SS76 37' A

No No y

L GlSS77 37' A

No No l

44SS84 14' A

No No 44SS86 14' A

No No 44SS98-12'.

A No No 61SS99 22' A

No.

No 61SS100 17' A

No No

-1 60SS101 42' A

No No 60SS102 42' A

No No 19SS103

-3' A

No No i

2SS104 12' A

No No 25S105

-17' A

No No 2SS106

-12' A

No No 22SS178

-11' A

No No i

k 20SS195 4'

A No No 20SS196 4'

A No-No 22SS197 13' A

No No I

{

5 Standby Gas Treatment System ISGT-8SS17 Reactor Building 69' A

No' No j

Bs f

a

=_

c.

p TABLE 3.'/.5-l' (Continued)

SAFETY RELATED HYCRAULIC SNUBBERS

• lk

~!

SNUBBER-SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION.

ESPECIALLY DIFFICULT 3

NO.

ON, LOCATION AND ELEVATION INACCESSIBLE 20::D -

TO REMOVE (A or I)

,(Yes or No)

(Yes or.No)

.y Instrument Sensing System 1B21-7015S164 Drywell 104' I

No

-No

-5 701SS167 104' I

No No

-4 701SS169 100' I

No No 70lSS170',

103' I

No No d

7015S171 99' I

No No 70lSS172 101' I

No No 7015S175 100' I

No No 70lSS177 94' I

No No 70lSS178 97' I

No No

!i 7015S179 96' I

No No 70lSS184 88' I

No No l'

I R

Reactor Closed Cooling Water System y

1RCC-32SS30 Reactor 55' A

No No 32SS45 Building 60' A

No No 36SS78 54' A

No No 37SS79 54' A

No

'No 39SS80 59' A

No No 38SS81 54

A No

.No 75S112-

'57' A

No No 47SS167 59' A

No No 47SS168 58' A

No No 48SS169-60' A

No No 50SS272 4'

A No No l

60SS121 Drywell 17' I

No t;o.

60SS122 16' I

No No l

65SS128 7'

I No No 65S5129 9'

I No No 71S5139 9'

I No No 7355145 5'

I No No 195S157 21' I

No No 195S160 29 I

No No

TABLE 3.7.5-1 (Continued)

L SAFETY RELATED HYDRAULIC SNUBBERS

• E SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGil RADIATION ESPECIALLY DIFFICULT NO.

Off, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE 5

(A or I)

(Yes or No)

(Yes or No)

.Q Primary Steam System IPSN-A2SS30' Drywe11 65' I

No No-c A2SS31 64' I

No N'o 5

A3SS32 40' I.

No No A3SS33 35'-

I No No 7

A3SS34 35' I

No-

'No A3SS35 41' I

No No ASSS38 22' I

No No 825S40 63' I

No No B2SS41 64' I

No No B3SS42 40' I

No No 83SS43 35' I

No No B3SS44 40' I

No No B3SS46 46' I

No No B3SS47 35' I

No No B3SS48 40' I

No No w

B5SS50 18' I

No No 1

BSSS51 22' I

No No w

C2SS54 63' I

No No

.C2SS55 64' I-No No C3SSS6 40' I

No No C3SS57 35' I

No No C3SS58 40' I

No No C3SS60 38' I

No No C3SS61 35' I

No No C;

52 39' I

No No CSSS64 18' I

No No C5SS65 22' I

No No D2SS68 65' I

No No E

t D2SS69 65' I

No No

./

03SS70 40' I

No No S

D3SS71 35' I

No No E

D3SS72 35' I

NO NO D3SS73 41' I

No No i

E D5SS76 22' I

No No

~

B3SS190 42' I

No No i

~

C3SS272 42' I

No No A3SS292 42' I

No No

I~

I'

' TABLE 3.7.5-1,(Continued) 3 SAFETY REL'!ED HYDRAULIC SNUBBERS

• l Ey SNUBBER -

SYSTEM SNUBBER INSTALLED

. ACCESSIBLE OR-HIGH RADIATION

'ESPECIALLY DIFFICULT

's NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE (A or 1)

(Yes or No).

(Yes or No)

^

'~

Reactor Core Isolation Cooling System

j 1E51-4SS45 Drywell 31' I

No No 3SS46 39' I'

No No 3SS47 39' I

No No 4SS66 39' I

No No 4SS68 40' I

No No l~

.45569

.40'

.I No

-No i

45S70 39' I

No No 1

4SS71 36' I

No No

i-4SS72 31' I

No No-4SS73 30' I

No No I

41SS51 Reactor 40' A

No No w1 42SS74 Building

.20' A

No No 42SS75 20' A

No No y

42SS76 18' A

No No 42SS77 5'

A No-No 42SS78 0'

A No No 42SS79 4'

A No No 42SS80-

-13' A

No No 42SS81

-16 '

A No No I

42SS82

-9' A

No No 405583

-9' A

No No 405S84

-9' A

No-No 40SS35

-12' A

No No i

405586

-9 A

No No

!2 40SS87

-15' A

No No l g

40SS88

-13' A

No No 41S589 41' A

No No

i n

h 41SS95

-41' A

No No i

m 19SS113

-17' A

No No i

5 19SS114

-16' A

No Ho g

195S129 0'

A No No

{

l I

TABLE 3.7.5-1 (Continued) 3 SAFETY RELATED HYDRAULIC SNUBBERS *.

.E

q! SNUBBER' SYSTEM SNUBBER INSTALLED ACCESSIBLE OR.

HIGH RADlATION ESPECIALLY DIFFICULT-3 NO.

ON, LOCATION AND ELEVATICN INACCESSIBLE ZONE **

TO REMOVE-(A or I)

(Yes or No)

(Yes or No)

I E

l q.

Nuclear Steam Vent System a

1821-44SS129 Drywell 1 04' I

No No 44SS131 93' I

No No 4

I

'44SS134 99' I

No No l

445S136 97' I

No No 44SS137 -

96' I

No No 44SS138 95' I

_ No No 44SS141' 87'-

I No No 44SS142 87' I

No No w

44SS143-87' I

No No

~ :

2 44SS146 87' I

No No u

44SS147 82' I

No No 44SS149 85' I

No No i

U^

44SS150 83' I

No

• No l

47SS155 75' I

No No 47SS156 78'

-I No No j7SS157 75 I

No No Standby Liquid Control System 1C41-9554 Drywell 63' I

No No 9S55 47' I

No No l

l 9SS8 42' I

No No 95S10 38' I

No No y

95Sil 39' I

No No

=

9SS12 69' I

No No y

95S13 52' I

No No 1

9S526 Reactor 72' A

No No a5 9SS27 Building 72' A

No No g

6SS34 84 A'

No No j

)

t

TABLE 3.7.5-1 (Continued)

.. y l

3 SAFETY RELATED HYDRAULIC SNUBBERS *

(f 5

Q SNUBBER-SYSTEM SNUBBER INSTALLED ACCESSIBLE OR.

HIGH RADIATION ESPECIALLY DIFFICULT ZONE **

TO REMOVE NO.

ON, LOCATION AND ELEVATION INACCESSIBLE

1j (A or I)

(Yes or No)

(Yes.or.No) c

'z 1[

U Fuel Pool Cooling System 1G41-1SS22 Reactor 12' A

No

.No

~ ',

ISS24 Building 38' A

No No 1SS30 38' A

No No 125532 9'

A No No 12SS33 9'

A No No

'l 155S37 111' A

No No

i 20SS76 10B' A

No No

,l 115S79 89' A

No No R

22SS85 108' A

No No 12S598 88' A

No No

?

6SS111 88'~

-A No No 4.

g 75S121 87 A

No No 55S152 82 A

No No Reactor Recirculation System.

i 1832-SSA1 Drywell 8'

I No No I

SSB1-81' I

No No l

SSA2 11' I

No No SSB2 11' I

No No SSA3 11' I

No No k

SSB3 11' I

No No 8.

SSA4

'21' I

No No 2

SSB4 21' I

No No a

SSAS 21' I

No No SSB5 21' I

No No 2

P SSA6 27' I

No No 5586 27' I

No No g

SS89A 30' I-No No SSB9B-30' I

No No ssA10 24' I

No No t

MLL 3.7.5-1 (Continued).

SAFETY RELATED HYDRAULIC SNUBBERS

• y SNUBBER' SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT' NO.

-ON LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE

'5' lAor1)

(Yes or No)

(Yes or 'No).

9 Reactor Recirculation System -(Continued) e E

- l

'SSAll ll' I

, No No Drywell (Cont'd)11'.

I ho No SSBil SSAl2A

.30' I

No No' 4

SSAl2B 30' I

No No-

.SSB12A 30' I

No No SSB12B 30' I

No No l

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 w2 3570 32' I

No No

)

~4 3613 32' I

No No No 3617A 32' I

No 3617B 32' I

No No 3751 34' I

No No 3752 34' I

No No Off Gas System j

IPS-3417 Nitrogen and 31' A

No No 3418A Off Gas Bldo 33' A

No No j

i 3418B 33' A

No No 3419A 33' A

No No E

3419B 33' A

No No

[

3819 37' A

No No a

Reactor Feedwater System E

1821-2553 Drywell 38' I

No No

{

2SS4 56' I

No No 3556 41' I

fid No l

3SS9 39' I

No

' No i

3SSil 41' I

No.

No i

3sS12 40' I

No No 3S513 61' I

fio No SSS17 38' I

No No SSS18 56' I

No No

b i

TABLE 3.7.5-1 (Continued).

E

~

SAFETY RELATED HYDRAULIC SNUBBERSA 2

ESPECIALLY DIFFICULT-E SNUBBER-SYSTEM SNUBBER INSTALLED ACCESSIBLE'0R HIGH RADIATION E-NO.

ON. LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE 7

.(A or 1)

(Yes.or No)

(Yes or No)

~

E Reactor Feedwater System (Continued)

ESS20 Drywell(Cont'd)41' I

No No 1

6SS23 39' I

Ne No 6SS25 41' I

No No 6SS26:

40' I

No No 6SS27 63' I

No No 1SS227-34' I

No No ISS228 38' I

No No I

1SS229 53'-

I No No

~l 255230 62' I

No No i

2SS231 40' I

No No 2SS232 36' I

No No 3SS233-40' I

No No

't 3SS234 48' I

No No I

3SS235 63' I

No -

No I

?

4SS236 34' I

No No I

E' 4SS237 38' I

No No f

SSS238 53' I

No No

.5SS239 61' I

No No 6SS240

'41' I

No No 6SS241 36' I

J No No 6SS242 39' I

No No

-655243 48' I

No No 6SS244 61' I

No No

'l Residual Heat Removal System l1 1 Ell-90SS267 Drywell 79' I

No No l P

905S268 86' I

No No 2

90SS271 86' I

No No g

90SS274 93' I

No No i'

i 905S275 93' I

No No 90SS277 96' I

No No i1 9055278 96' I

No No 90SS25G 101' I

No No Ili

.90SS281 93' I

No No

's 90SS282 101' I

No No

1 ja SAFETY RELATED HYDRAULIC SNUBBERS *.

5 12 SNUBBER-SYSTEM SNUBBER INSTALLED ACCESSIBLE OR

'HIGH RADIATION ESPECIALLY-DIFFICULT

~

.R NO.

ON,-LOCATION AND ELEVATION INACCESSIBLE ZONE **'

TO, REMOVE f-

.(A or I)

(Yes or No)

(Yes or No) lE Residual Heat Removal System (Continued).

.Z lEll-90SS283 Atyxell(Cont'd)93' I'

No

'No 90SS284 101' I

No No 9055285 100' I~

No No ISS302 35' I

No

-No-1SS303 34' I

No No-ISS305 20' I -~

No No ISS306 19' I

No No 84SS309 33' I

. No No-84SS311 35' I

No

-. No 84SS312 35' I

No No

-w 87SS315 33' I

No No 2

87SS317 35' I

No No u

87SS318

'5' I

No No A>

905S388 79' I

No No 90SS389' 96' I

No No 90SS390 96' I

No No 90SS391 93' I

No No 90SS392 99' I

No No lEll-17SS3 Reactor 3'

A No No l 17SS4 Building

-1' A

No No 46SS7 12' A

No No i !-

46SS9-12' A

No No i l 56SS13 5'

A No No p

56SS15 4'

A No No a

20SS19-

-3' A

No No 95S520

-3' A

No No g

95SS22 9'

A No No j

i 95SS23 9'

A No No

-3' A

No No

=

20SS28 o

58SS32 3'

A No No 58SS33

-4' A

No No.

m 58SS35 8'

A No No 59S535 8'

A No No

r -

l:

I-1 l

g.

TABLE 3.7.5-1 -(Centinued) 3 SAFETY:RELATED HYDRAULIC SNUBBERS *

.g:

Y 1.

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR!

HIGH RADIATION ESPECIALLY DIFFICULT.

NO.

ON, LOCATION AND ELEVATION INACCESSIBLE _-

ZONE **

TO REMOVE I-

.E (A or 1)

(Yes or No)

(Yes oT ifo)

'Q Residual Heat Removal System (Continued) 1 Ell-1BSS40 Reactor Bldg.

8

A' No No 18SS47 (Cont'd) 12' A

No No.

18SS48 13' A

No No

(;

68SS59 15' A

No No 21SS63 8'

A No No 21S570 13' A

No No 21SS71 10' A

No ho 37SS95

-3' A

No

.No 61SS110 6'

A No No No 116SS143

-11' A

No No 113SS157

-11' A

No l

R 37SS184

~9' A

No No

/

535S192 78' A

No -

Ho'

^

Y 53SS195 14' A

No No 53SS197 14' A

No No 53SS200 14' A

No No

.50SS201 14' A

No No l

895S208' 5'

A No No 46SS216 28' A

No No 46SS217 31

A No No I

46SS218 30' A

No No No 3il 47SS223 33' A

No

. No j

F 47SS224 36' A

No 47SS225 36' A

No No i

475S227 39' A

No No s

47SS228 39' A

No No 95SS233 20' A

No No i ;

l 95SS234 24' A

No No 95SS235 31' A

No No Dj 131SS255 42' A

No No No No 131SS257 22' A

132SS263 30' A

No No l

i

i TABLE 3.7.5-1'(Continued)

S' E

SAFETY RELATED HYDRAULIC SNUBBERS

• E

' SNUBBER SYSTEM SNllBBER INSTALLED ACCESSIBLE OR

-HIGH RADIATION' ESPECIALLY 0IFFICULT.

?'

NO.

'ON, LCCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE-(A or I)

(Yes or No)

-(YesorNo)

-zy Residual Heat R.moval System (Continued) 1E11-132SS264 Reactor Bldg.

31'-

A No No 21SS296 (Cont'il)

~

39'.

A No No 21SS297 39'

-A No No 47SS323 42' A

No-No 47SS326 42' A

No

'No 47SS328 42' A

No No 49SS330 42' A

No No 49SS331

.42' A

No No 49SS333 42' A

No No 4

49SS334 43' A

No No 49SS336 40' A

NC No y

128SS355 42' A

No No y

.49SS359 42

.A No No 127SS376 59'

.A No No 12855387 43' A

No No 2SS396 5'

A No No 2SS397 3'

A No No SSS398

-41' A

No No

-2SS399

-3' A

No No SSS400-

-3' A

No No i

4SS401

-12' A

No No 5S5402

-12' A

No No F

3SS403

-11' A

No No 6SS404

-12' A

No No

'I 8SS405

-14' A

No No i

E 6SS406

-14' A

No No 85S407

-15' A

No

.No E

12SS408 14' A

No No 16SS409

-9' A

No No

'113S5410

-9' A

No No 955411

-14' A

No No 1095S412

-14' A

No No 2SS413 0'

A No No 132SS414 43' A-No No

TABLE 3.7.5-1 (Continued)

E-SAFETY RELATED HYDRAULIC SNUBBERS

• 5-3 SNUBBER-

' SYSTEM SNUBBER INSTALLE, ACCESSIBLE'~0R-HIGH RADIATION ESPECIALLY DIFFICULT-

.E-NO.

'ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **'

TO REMOVE (A or-I)

(Yes or No)

(Yesorlo)

~

e Residual Heat Removal System (Continued) lEll-20SS415 Reactor Blda.

-4' A

No No l.

-20SS416

'(Conttq) 4' A

No No 9555417 0'

A No No 9555418

-4' A

No No-l 45SS422

-2' A

No No 60SS423

-4' A

No No 60SS425

-2' A

No.

No 127SS426 43' A

No No 128SS427 21' A

No No i

128SS428 28' A

No No R

128SS429 39 '.

A No No

, 128SS430 31'~

A No No Y

128SS431 23' A

No -

No E'

127SS433 14' A

No No 127sS434 37' A

No No 127S5435 15' A

No No 60SS437 12' A

No No

-l 60SS438.

12' A

No No 60SS440 13' A

'No No l[

65SS441 3'

A No No.

65SS442 3'

A No Ne 60SS443 11' A

No No 73S5444 21' A

No No F

21SS445 5'

A No No I

E 83SS446 10' A

No No 68SS448 13' A

No No 7555449 7'

A No No I

6155450 2'

A No No e

I 60SS451 13' A

No No 60SS452 13'

_A No No 6055453

.10' A

No No 605S454 10' A

No No co 8955459 11' A

No No

?.

TABLE 3.7.5-1 (Ccntinued)

SAFETY RELATED HYDRAULIC SNUBBERS

• x p;

SNUBBER SYSTEM SNUBBER INSTALLED-ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT.

X-NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **~

TO REMOVE (A or I)

(Yes or No)

(Yes or.No)

. Residual Heat Eemoval System (Continued) lEll-89SS460 Reactor Bldg.

10' A

No No

~

895S461 (Cont'd)

-6' A

No No 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

[

t' 89SS480 67' A

No No

+

89SS487 67' A

No No

?

8955489 67' A

No No M

895S491 67' A

No No 91SS499 69' A

No No 91SS500 57' A

No No

)

56SS504 14' A

No Nol V

56SS505 7'

A No No 56SS506 3'

A

.No No 3

56SS507 3'

A No No 56SS508 4'

A No No 46SS509 8'

A No No 46SS510 11' A

No No l

F 46SS511 10' A

'No No g.

46SS512

-l' A

No No 58SS514 14' A

No No

-)

g 49SS515 37' A

No No 49SS516 37' A

No No i

4955517

-5' A

No No ij 51SS546 32' A

No No 5155547 28' A

No No i,

Ii 115SS549 31' A

No No N*

i$-

<t

=

TABLE 3.7.5-1 (Continu d)

SAFETY RELATED' HYDRAULIC' SNUBBERS

• Y.

SNUBBER.

.-SYSTEM SNUBBER INSTALLED

' ACCESSIBLE OR HIGH' RADIATION ESPECIALLY DIFFICULT-

.R-

~NO.

_0N, LOCATION AND ELEVATION INACCESSIBLE

' ZONE **

'T0; REMOVE (A or I)

(Yes or No).

(fes or No) i-

' Residual Heat Removal System (Continued) 1E11-54SSE51 Reactor Bldg.

31' A

N6 No 54SS552 (Cont'd) 28

A No No 98SS554 32' A

No No-5855563 7'

A No No 585S565 13

A No No

.58SS566 6'

A No-No 1075S573 15' A

No No 695S57A 6'

A No No 91SS5/5 53' A

No No 68SS577 8'

A

.No No R

53SS596 14' A

No No 50SS597 26' A

No No E

Service Water System

.ISW-133SS22-Reactor

-6' A

No No 110SS35 Building

-5' A

No No 174SS70 42' A

No No 173SS72 14' A

No No 142SS74 40' A

No No 142SS75 40' A

No No 140SS80 40' A

No No 1405582 70' A

No

.No i

g 140SS86 45' A

No No j

u l

E 153SS102 44'

~A No No

.I 3

15355109 44' A

No No I

5 173SS110 48' A

No No z

17355114 70' A

No No

?

1535S115 44' A

No No I

103SS117 41' A

No

.No 103S5121

.38' A

No No m

'. i 103SS126 60' A'

No No 10355127 57' A

No No

b TABLE 3.7.5-1 (Continued) a,.

E SAFETY RELATED HYDRAULIC SNUBBEfj*

j k

SNUBBER' SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT ~

.NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE (A or I)

(Yes or.No)

(Yes or No)

.g.

g Service Water System (Continued)

L 1SW-106SS131 Reactor Bldg'.

60' A

No No i

100SS145 (Cont'd) 59' A

No No 100SS149 60' A

No No 106SS151 59 A

No No 106SS156 60' A

No No i

142SS163.

60' A

No No l-

-142SS164' 8'

A No No 142S5165 8'

A No No

'j 175SS166 42' A

No No l t

140SS167 42' A

No No i

142SS168 58' A

No No l

?

142SS169 71' A

No '

No 0

138SS170 55' A

No No 140SS171 58' A

No No 139SS172

-58'

'A No No

~1745S174 42' A

fio No 173SS175 30' A.

No No 133SS177

-5' A

No No 100SS193 62' A

No

.No 139SS209 56' A

No No 139SS210 56' A

No No g

106SS211 60' A

No No E

106SS212 60' A

No N7 m

2 106SS213 59' A

No No

'5 106SS214 60 A

No No z

127SS215 17' A

No No l

123SS216 17' A

No No g

i I

TLUC I

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IVf DOM r YE o

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5 NNNNNNNNNNNNNNNNNNNNNNNNNNNNNNNN AOc IT Y

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NNNNNNNNNNNNNNNNNNNNNNNNNNNNNNN'.

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.O 801246790567901678906789023456759 566666667777788888899999000000000 222222222222222222222222333333333 SSSSSSSSSSSSSSSSSSSSSSSS5SSSSSSSS SSSSS55SSSSSSSSSSSSSSSS5SSSSSSSSS R

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E TABLE 3.7.5-1 (Centinued)

-E W

SAFETY RELATED HYDRAULIC SNUBBERS

• R*

SNUBBER

-SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT-3-

NO.

ON, LOCATION AND ELEVATION INACCESSIBLE' ZONE **

TO REMOVE 9

E (Aor1)

(Yes or No)-

(Yesorllo) b Q-Steam Relief Discharge System (Continued)

?

1B21-12SS310 Drywell_ ((ont'd)l7' I

No.

No 12SS311 11' I

No No 12SS312 11' I

No No

(

11SS313 16'

.I No No

]- ~

llSS314 11' I

No No 11S5315 7'

I No No l

19SS316 11' I

No No 195S317 17' I

No No 1

w2 18SS318 11' I

No No 20SS319 10' I

No-No u

'a 595S320 11' I

No No

's 59sS325 34' I

No,

No i

59sS326 35' I

No.

No 59SS327 36' I

No ho 5950329 38' I

No No 59SS330 44' I

No No i

33SS332 26' I

No No 33SS333 7'

I No No 33SS334 11' I

No No 3355335 29' I

No Nol 34SS336 27' I

No No 3455337 18' I

No No 34SS338 18' I

No No u

E 34SS339 11' I

No No j'

34SS340 7'

I No No

• Snubbers may be added to safety related systems without prior License Amendment to Table 3.7.5-1 provided I

that safety evaluations, documentation and reporting are provided in accordance with 10 CFR 50.59 and that a proposed revision to Table 3.7.5-1 is included with the next License Amendment request.

i

~ Modifications to this table due to changes in high radiation areas shall be submitted to the NRC as art of the next License Amendment request.

" T.,

M.X"'

REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least 2 source range monitor (SRM) channels

• shall be OPERABLE and inserted to the nonnal operation level with:

a.

A continuous visual indication in the control room, b.

One of the SRM detectors located in the quadrant where CORE.

ALTERATIONS are being performed and the other SRM detector located in an adjacent quadrant, and c.

The " shorting links" removed from the RPS circuitry prior to and during the time any control rod is withdrawn ** and shutdown margin demonstrations.

APPLICABILITY:. CONDITION 5.-

ACTION:

With the requirements of the above specification not satisfied, ime-diately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods.

The provisions of Specification 3.0.3 are not applicable.

SURYE8.ANCE REQUIREMENTS 4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:

a.

' At -least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; 1.

Performance of a CHANNEL CHECK,

• The use of special movable detectors during CORE ALTERATIONS in place of the nonnal-SRM nuclear detectors is. permissible as long as these special detectors are connected to the normal SRM circuits.

• • Not required for control rods removed per Specifications 3.9.10.1 or 33 3.9.10.2.

y=p BRUNSWICK - _ UNIT-1 3/4 9-3 Amendment No. 26 y

p

~

INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) 2.

-Verifying the detectors are inserted to the normal operat-

~

ing level. and 3.

During CORE ALTERATIONS, verifying that the detector of

,an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and one is

- located in the adjacent quadrant.

b.

Performance of a CHANNEL FUNCTIONAL TEST:

1.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of CORE ALTERATIONS, and 2.

At least once per 7 days, Verifying at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS c.

that the channel count rate is at least 3 cps.

1; J

4

/4 BRUNSWICK:- UNIT 1 me

=

43 W-,

1 M

g r g g

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[(p.ne.gjo,i UNITED STATES i

NUCLEAR REGULATORY COMMISSION t

% +...,r;'

wasumaron, o. c,:osss CAROLINA POWER & LIGHT COWANY DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 License No. DPR-62 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Carolina Power & Light Company dated February 20, 1979, as supplemented January 14, 1980, and applications dated November 19, 1979 and January 24, 1980, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Conmission's rules ar.1 regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confonnity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C..There is reasonable assurance (i) that the activities authoriied by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The -issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

j 4

2.

Accordingly,^ the license.is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph '2.C.(2) of Facility Operating License No. DPR-62 is hereby' amended to read as follows:

(2) -Technical Soectfications The Technical ' Specifications contained in Appendices A and B, as revised through Amendment No. 50, are hereby incorporated in the license. - The licensee shall operate the facility in accordance with the Technical Specifications.

k-

b

. 3.

This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION.

O Thomas

. Ippolito, Chief Operating Reactors-Branch #3 Division of Operating Reactors 4

Attachmenti Changes to the Technical Specifications

. Date of-Issuance: March 14,1980 O

s t

t

.i e =

ATTACHMENT TO LICENSE AMENDMENT NO. 50' FACILITY OPERATING LICENSE NO. DPR-62 DOCKET NO. 50-324 Remove the-following pages and replace with identically numbered pages.

3/4 3-1/3/4 3-2 3/4 3-5/3/4 3-6 3/4 7-11/3/4 7-12 3/47-13/3/47-14 3/4 7-15/3/4 7-16 3/4 7-17/3/4 7-18 3/4 7-19/3/4 7-20 3/4 7-21/3/4 7-22 3/4 7-23/3/4 7-24 3/4 7-25/3/4 7-26 3/4 7-27/3/4 7-28 3/4 7-29/3/4.7-30 3/4 9-3/3/4 9-4

-The underlined pages are overleaf pages and are provided for convenience.

b l

e l

l i

.4 i

-. ~. - -.,

.+.

~

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 - As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

Set points and interlocks a re given -in Table 2.2.11-1.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

With the requirements for the minimum number of OPERABL'E chan-a.

nels not satisfied. for one trip system, place that trip system in the tripped condition within one hour or take the ACTION required by Table 3.3.1-1.

b.

With the requirements for the minimum number of OPERABLE chan-

~..

nels not satisfied for both trip systems, take the ACTION

" ~ " "

required by Table 3.3.1-1.

c.

The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION S.

SURVEILLANCE REQUIREMENTS 4. 3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK. CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated autcmatic operation of all channels shall be performed at least once per 18 months and shall include calibration of ' time delay relays and timers necessary for proper functioning of.thei trip system.

4.3.1.3 The REACTOR-PROTECTION SYSTEM RESPONSE TIME of each reactor trip function of Table 3.3.1-2 shall be demonstrated to be within its limit at least once per.18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and 'one: channel per function such that all channels are tested at least L.

once' every N times 18 months where N is the total number of redundant channels in a specific. reactor trip function.

BRUNSWICK - UNIT 2 3/4 3-1 e

--we r

-w,

TABLE 3.3.1-l' E'

. REACTOR PROTECTION SYSTEM INSTRUMENTATION v,

.5 APPLICABLE MINIMUM NUMBER S

OPERATIONAL OPERABLE CHANNELS

~ FUNCTIONAL' UNIT AND INSTRUMENT NUMBER CONDITIONS PER TRIP SYSTEM (a) ACTION l

.c

• ~

1.

Intennediate Range Monitors:

[

'(C51-IRM K601 A,B,C,0,E,F,G H) a.

Neutron Flux - High 2, 5( )

3 1

l 3, 4 2

2 b.

Inoperative 2, 5 3

1 3, 4 2

2 2.

Average Power Ra-Monitor:

t (C51-APRM-CH.AA.C,D,E,F) y

- a'.

Neutron Flux - High,15%

2,5(b) 2 3

l-N b.

Flew Biased Neutron Flux -

High 1

2 4

c.

Fixed Neutron Flux-High,120%

1 2

4 Inoperative.

1,2,5 2

5 e.

Downscale 1

2 4

4 f.

LPRM 1,2,5 (c)

NA F

a 3.

Reactor Vessel Steam Dome Pressure -

5d) p High (B21-PS-N023 A,B,C,0) 1, 2 2

6 5

4.

. Reactor Vessel Water l'evel -

Low, Level #1 (B21-LIS-N017 A,B,C,D) 1, 2 -

2 6

=

5.

Main Steam Line Isolation Valve.

y, Closure (B21-F022A,B,C,Dand 4

4

~B21-F028.A,B,C,D) 1 6.

Main Steam Line Radiation - High 1,2(d-)

2 7

(012-RM-K603 A,B,C,0) l l

~

i

~

9

.c.

i, '

TABLE '3.3.1-1 (Continuc{l AL REACTOR PROTECTION SYSTEM INSTRUMENTATION

.;r ACTION 10 - In OPERATIONAL CONDITI0d 1 or 2, be in at least HOT SHUT 00WN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In OPERATIONAL CONDITION 3 or 4, lock the reactor mode switch in the Shutdown position within one hour.

In OPERATIONAL CONDITION 5, suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods within one hour.

TABLE NOTATIONS

..a.

A channel may be placed in an inoperable status for up to 2 hcurs for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

b.

.The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn

• and shutdown nargin denonstrations.

_[.g c.

An APRM channel is inoperable if there are less than 2 LPRM inputs- ~~ --- -- "- -

per level or less than eleven LPRM inputs to an APRM channel. '-

d.

These functions are not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed.

e.

This function is not required to be OPERABLE when PRIMARY CONTAIN-MENT INTEGRITY is not required.

f.

• dith any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

g.

These functions are bypassed when THERMAL POWER is less than 30%

of RATED THERMAL POWER.

• Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

I

?_..

BRUNS'AICK UNIT 2 -

3/4 3-5 Amendment No. 50 4

-e,..-

...-a

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

.-.....-w..,

n.

e.

TABLE 3.3.1-2 REACTOR PROTECTION SYSTEM RESIWISE TlHES E!

FuntilotlAL UNIT AND !NSTRUMENT NUMBER j

~

~

RESPONSE TIME (Seconds) l.

Intermediate Range Monitors (C51-IRM-K601 A.B.C.D.E.F, Goll):

E.

a.

Neutron Flux - High*

• I h.

Inoperative NA IIA

?.

' Average Power Range Monitor * (CSI-APRM-Cll.A,B,C.D.E.F):

a.

Neutron Flux - High, 15%

< 0.09 c.

Neutron Flux - High, 120%

~

FA b.

Flow Blased Neutron Flux - liigh

< 0.09 d.

Inoperative RA NA e.

Downscale I.

LPRM MA 3.

Reactor Vessel Steam Dome Pressure - High (B21-PS-N023 A,B,C.D)-

1 0.55

{

4.

Reactor Vessel Water Level - Level #1 (B21-LIS-N017 A,B,C,0) 1 05 1

5.

Hain Steam Line Isol'ation Valve-Closure (B21-F022 A,B,C,0 and B21-F028 A.B.C.D) 1 0.06 9

l 6.

Main Steam Line Radiatloa - liigh (012-RH-K603 A.B.C.D)

HA 7.

Drywell Pressure - liigh (C72-PS-N002 A,B,C,0)

NA l

n.

Scram Discharge Volume Water Level - liigh (C12-LSil-N013 A.B.C,0)

NA

.l I

4 lurbine Stop Valve - Closure (EllC-SV05-lX.2X,3X.4X) 1 0.06 10.

Turbine Control Valve Fast Closure.

y control Olt Pressure - Low (EHC-PSL-1756,1757,1758,1759) 1 0.08 l

5H 11.

Reactor Mode Switch in Shutdown Position (C72A-SI)

NA l

!5

17. ~ManualScram(C72A-S3A,B)

NA l

1 os3-Silentfron detectors are exempt from response time testing.

Response time shall be measured from j

g, ilniertor output or input of first electronic component in channel.

D*~

D D

FU^ 3 A i

.b.S b

e u.

s

TABLE 3.7.5-1 "g

SAFETY RELATED HYDRAULIC SNUBBERS

• j

~

rA 3

SNUBBER SYSTEM SNUBBER INSTALLE0 ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT p

NO.

~0N, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE

)

' Core Spray System za H

2E21-2SS32 Reactor Buildinc 66' A

No No

-26SS91

-6' A

No No N

' 2SS16 0'

A No No 2SS17 13' A

'No

'No 15SS19

-3' A

No No 15SS20

-3' A

No No 28SS23

-4' A

No No 25SS96

-6' A

No No 40SS106

-12' A

No No '

40SS107

-12' A

No No 393S108

-12' A

No No 395S109

-12' A

No No e3 2SS31 68' A

No No 6SS41 70' A

No No u

.L 6SS42 69' A

No No 2SS18 14' A

No No 2E21-3SS46 Drywell 63' I

No No 3SS47

'63' I

No No 3SS48 65' I

No No 3SS49 66' I

No No 7SS53 63' I

No No 7SS54 63' I

No No 7SS55 65' I

No No j

7SS56 66' I

No No llt -

ii j

~

t l

a e

O l

l

' TABLE 3.7.5-1 (Continued)

'co

?

SAFETY RELATED HYDRAULIC SNUBBERS

• E SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGil RADIATION ESPECIALLY DIFFICULT R

SNUBBER NO.

ON, LOCATION AND ELEVATION INACCESSIBLE 20 tie **

TO-REMOVE

{

'E-Reactor Water Cleanup System G

-2G31-lSS3 Drywell 54' A-No No ro Condensate Drains System 2B21-51SS103 Drywell 29' I

No No R

SISS105'

-26' I'

No No F

SlSS106 18' I

No No 51SS109 31' I

No No 3-51SSl11 28' I

No No SlSSll3 23' I

No No 1-w2 515S115 24' I

No No

~

51SS118 24' I

No No

-y e

ControT Rod Drive System 2Cl2-16SS1 Drywell 69' I

No No 165S6 68'

'I No No 16SS7 69' I

No No 16SS8 70' I

No No

' 16SS10 72' I

No No 16S511 72' I

No No 16SS12 72' I

'No No E

High Pressure Coolant Injection System 2E41-4SS44 Orywell 40' I

No No 3

4SS45 35' I

No No 5

4SS47 40' I

No No

~

4SS49 37' I

No No 4SS50 40' I

No No 4SS51 30' I

No No O

e 9

TABLE 3.7.5-1 (Continued)

SAFETY RELATED HYDRAULIC SNUBBERS

• k

SNUBBER.

SYSTEM SNUBBER INSTALLED ACCESSIBLE OR

HIGH RADIATION ESPECIALLY DIFFICULT 9

NO.

-ON, LOCATION AND ELEVATION-INACCESSIBLE ZONE **

TO REMOVE High Pressure Coolant Injection System (C'ontinued) 2E41-2SS5 Reactor Buildin

-7' A

No No 615S73

-2' A

No No 60SS9 4'

A

.No No l20SS15

-5' A

.No No

~ 44SS84 14' A

No No' 44SS86 14' A-No No

'44SS98 12' A

No No 19SS103

-3' A

No No 2SS104 12' A

No No 2SS105

-17' A

No No m2 2SS106

-12' A

No No 60SS52 41' A

No No y

1-21SS76 -

37' A

No No w

21SS77 37' A

No No 61SS99.

22' A

No No 61SS100 17' A

No No

'60SS101 42' A

No No 60SS102 42' A

No No 22SS178

-11' A

No No 6SS27 1'

A No No l 6SS28 l'

A No No 6SS30

-l' A

No No y

16SS32

-5' A

No No l'

6SS35

-l' A

No No

'g 6SS33 A

No No g

6SS36

-5' A

No No A.

No No r+

65537 0

7 6S538 l'

A No No 6SS40 2'

A No No 6SS42

-4' A

No No E

6SS64

-l' A

No No 61SS71

-4' A

No No A

No No 615572

-l' j

l r.

TABLE 3.7.5-1 (Continued)

SAFETY RELATED HYDRAULIC SNUBBERS

• j SNUBBER' SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT-X NO.~

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

~TO REMOVE e

g Standby Gas Treatment System

[

2-SGT8SSl7 Reactor Building 69' A

No No Instrument sensing System 2B21-70lSS164 Drywell 104' I

No No 7015S167 104' I

No No 70lSS169 ^

100' I

No No 701SS170 103' I

No No 70lSS171 99' I

No No 701SS172-101' I

No No R

701SS175 100' I

No No 70lSS177 94' I

No No

?

70lSS178 97' I

No No g

70lSS179 96' I

No No -

701SS184 88' I

No No Reactor Building Closed Cooling Water System 2RCC-58SS11 Reactor Building 39' A

No No 50SS33 48' A

No No 50SS34 48' A

No No 10SSS6 48' A

No

~

No 17SS74 38' A

No No 17SS75

" 48' A

No No 175S76 39' A

No No 3"

32SS30 55' A

No No

~

.32SS52 64' A

No No 36SS78 54' A

No No 37SS79 54' A

No.

No c.

F 8

TABLE 3.7.5-1 (Continued)

I SAFETY RELATED HYORAULIC SNUBBERS

• U.-

'i

~ SNUBBER-SYSTEM SNUBBER INSTALLED ACCESSIBLE OR.

HIGH RADIATION ESPECIALLY DIFFICULT S

NO.

ON, LOCATION AND ELEVATION INACCESSIBLE' ZONE **

TO REMOVE g

Reactor Building Closed Cooling Wcter System (Continued).

[

2RCC-39SS80 ReactorBldg(Cont'd) 59'

-A No No 38SS81 54' A

No No 34SS82-57' A

No No 35SS83 57' A

'No No 6SS107 60' A

No No 7SS112 57' A

No No

'47SS168 ~

58' A

No No 48SS169 60' A

No No 32SS45 60' A

No No

.51SS58 88' A

No No t

47SSl67 59' A

No No-l 2RCC-60SS121 Drywell 17' I

No' No

'60SS122 16' I

No-No

-d h

65SS128 7'

I No No 65SS129 7'

I No No 71SSl39 9'

I No No 73SS145 5'

I No -

No 19SSl57 21' I

No No.

19SS160 29' I

No No Primary Steam System 2PSN-A2SS30 Drywell 65' I

No No A2SS31 64' I.

No No A3SS32 40' I

No No k

A3SS33 35' l

I No No l

R A3SS34 35' l

I No No-i 2

'ASSS38-22' I

No No

[-

3 B2SS40 63' I

No No

'B2SS41 64' I

No No x

P B3SS43' 35' I

No No

]

I 1

TABLE 3.7.5-1 (Continued) j SAFETY RELATED HYDRAULIC SNUBBERS *_

3 p;:

SNUBBER

' SYSTEM SNUBBER INSTALLED ~

ACCESSIBLE OR-HIGH RADIATION ESPECIALLY DIFFICULT.

M' N0.

-ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **-

TO REMOVE-

-Primary Steam System (Continued) q g

2PSN-B3SS44

Drywell (Cont'd) 40' I

No No B3SS46 38' I

No No B3SS47 35' I

No No

.B3SS48 40' I

'No No B5SS50 18' I

No No B5SS51 22' I

No No

'C2SS54 63' I

No No C2SS55 64' I

No No R

C3SS56 40' I

No No No' C3SS57 35' I-No

~

No y

C3SS58 40' I

No '-

g C3SS60 38' I

No No C3SS61 35' I

No No C3SS62 39' I

No No CSSS64 18' I

No No CSSS65-22' I

No No D2SS68 65' I

No No D2SS69 65' I

No No D3SS70 40' I

No No 03SS71 35' I

No No D3SS72 35' I

No No 03SS73 41' I

No No D5SS76-22' I_

No No A3SS35 41' I

No No B3SS42 40' I

No.

No g

3 i

i n

rt

.F i

l s

TABLE 3.7.5-1 (Continued)

-SAFETY RELATED llYDRAULIC SNUBBERS

• M'

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

'R NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE.

Reactor Core Isolation Cooling System

"~

2E51-4SS45 Drywell 31' I

No-No 3SS46 39' I

No.

No 3SS47 39' I

No No 4SS66 39' I

No No 4SS100 40' I-No No 4SS101 40' I

Na No 4SS102 39' I

No No 4SS103.

36' I

No No w

4SS104 31' I

No No 1

4SS105 30' I

No No u

42SS76 18' A

No No 1

42SS77 5'

A No No 42SS79 4'

A No No 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 40SS87

-15' A

No No 40SS88

-13' A

No No 415S51 40' A

No No 42SS74 20' A

No No 42SS75 20' A

No No 3

0 i

g i

l g

i g

Y g-TABLE 3.7.5-1 (Continued)

SAFETY RELATED HYDRAULIC SNUBBERS

• il

' n

-SNUBBER

. SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT.

^'

x NO.

ON. LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE i

E-Q.

-Reactor Core Isolation Cooling System (Continued) 2E51-41SS89 Drywell (Cont'd) 41' A

No-No 41SS9S-41' A

No

- No 195S113

-17' A

No.

No 19SS114

-16' A

No No 16SS91

-6' A

No No 17SS94

-6' A

No

'No 42SS78 0

A No No 49SS129 0

A No No M-y

i e

't e

LIil y

i, g-1, g

.P' n

O ii 8

r O

03'.

TABLE 3.7.5-1 (Continued) 3.

SAFETY RELATED llYDRAULIC SNUBBERS

• r/i A

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECTALLY DIFFICULT-X NO.

ON, LOCATION AND ELEVATION INACCESSIBLE 20NE**

!0 RE."0VE

~

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 C7' I

No No 44SS143 87' I

No No 44SS146 87' I

No No R

44SS147 82' I

No No 44SS149 85' I

No No 7

44SS150 83' I

No No 44SS155 75' I

No No-44SS156 78' I

No No 44SS157 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 9SS11 39' I

No No 95512 69' I

No No p

i 9SS13 52' I

No No n,.

1 2C41-9SS26 Reactor Building 72' l

A No No 2

9SS27 72' j

A No No A

No No R

9SS34 84'

.E a

I l

I

TABLE 3.7.5-1 -(Continued)

SAFETY RELATED HYDRAULIC SNUBBERS *

- e'.

SNUBBER-SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT-M NO.

'ON, LOCATION AND ELEVATIOK INACCESSIBLE ZONE **-

TO REMOVE I

g.

Fuel Pool Cooling System

]

2G41-lSS22 Reactor Building 12' A

-No No 12SS32' 9'

A

No No.

'125533 9'

A

.No No 1SS24 38' A

No No ISS30 38' A

No No No No 155537 111' A

11SS54

-88

A No No-10SS67 89' A

No No 20SS76 108' A

No No 11SS79 89' A-No No m.

1 22SS85 108'-

A No' No 12SS98 88 A

No No y

A, 6SS111 88 A

No No o-7SS121 87' A

No No 4

555152 82' A

No No 59SS159 84' A

No.

No i ' -

Reactor Recirculation System 2B32-SSA1 Drywell-8' I

No No j

55B1 8'

I No No SSA2 11' I

No No SSB2 11' I

No-No SSA3

. 11' I

No No SSB3 11' I

No No k

SSA4 21' I

No No E

SSB4 21' I

No No s

SSA5

~

21' l

I No No l

No l

I "No a

SSB5 21' f

8

c l'...

+

ry g

. TABLE 3.7.5-1 (Continued)-

'l h

SAFETY RELATED HYDRAULIC SNUBBERS

• I 5-0:

SNUBBER SYSTEM SNUBBER' INSTALLED ACCE.,61BLE OR-HIGH RADIATION ESPECIALLY' DIFFICULT-i NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **-

TO REMOV_E Reactor'RecirculationSystem'(Continued)

~2B32-SSA6 Drywell-(Cont'd) 27' I

No

.No SSB6 27' I

No No SSB9A 30' I

No No SSB98 30' I

No' No SSA10 24' I

No No SSAll 11' I

No No SSBil 4'

I No No.

SSA12A 30' I

No No

'SSAl2B

~30' I

No.

No U

SSB12A 30' I

No No

[

SSB128 30' I

No No j

Reactor Vessel Instrumentation

I 2PS-3554 Drywell 63' I

No No 3558 68' I

No No l

3561 60' I

No No

1 3562 60' I

No.

No 3567 65' I

No No i

3570 65' I

No No 3613 90' I

No No i

F 3705A 32' I

No No 1

.3705B 32' I

No No j

3706 32' I

No No i

~@

3707 32' I

No No 3708 32' I

No No i

E 3709 32' I

No No 3710 32' I

No No 3722A 82' I

No No m

37228 8?'

I No No

jg.

TABLE 3.7.5-1 (Continued)

Cj_

SAFETY RELATED !!YDRAULIC SNUBEERS*.

M 3

. SNUBBER

SYSTEM SNUBBER INSTALLED ACCESSIBLE. 0ft HIGH RADIATION ESPECIALLY DIFFICULT-E.

NO.-

ON, LOCATION AND ELEVATION INACCESSIBLE,,

ZONE **

TO REMOVE-y Off Gas System.

31' A

No No 2PS-3417'

' Nitrogen and 3418A.

0ff Gas Bldg.

33' A

No No 34188 33' A

No.

No.

-3419A-33' A

No No 34198 33' A'

No

'No 3423 37' A

No No Reactor Feedwater System Y

2B21-2SS3 Drywell 38' I

No No

~2SS4 56' I

No

.No y

k 3SS6 41' I

No No N

3559 39' I

• No No

~

3SSl1 41' I

No No 35512 40' I

No No 35513 61 '

I No No SSS17 38' I

No.

No 55S18 56' I

No No 6S520 41' I

No No 65S23 39' I

No No 6SS25 41' I

No No 4

1 F

65S26-40' I

No No i

(

6SS27 63' I

No No

j s

.15S227 34' I

No No

' i 1SS228 3'8 '

I No No n

2SS229 53' I

No No E

2SS230 62' I

No No 8

e e

TABLE 3.7.5-1 (Centinued)

p 1

j g

SAFETY RELATED HYDRAULIC SNUBBERS

• 4 C

5 3

' SNUBBER --

~ SYSTEM SNUBBER IHSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT

'I N0.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE m

Reactor Feedwater System (Continued) e c'

2B21-3SS231 Drywell-(Cont'd)'

40' I

No No

-3SS232 36' I

No No m

355233 40' I

No No 3SS2.M 48' I

No No 3SS235 63' I

No,

No

~

4SS236 34' I

No No 455237-38'-

I No No SSS238 53

I No_

No e

SSS239 61' I

No No 6SS240' 41' I

No No 6SS241 36' I

No No w

i 1

6SS242 39' I

No No 6SS243 48' I

No

-No y.

A, 6SS244 61' I

No No Residual Heat Removal System 2 Ell-90SS267 Drywell 79' I

No No 4

90SS268 86' I

No No 90SS271 86' I

No No 90SS274 93' I

No No 90SS275 93' I

No-No 90SS277 96' I

No No 7

90SS278:

9'a '

I No No I-9055280 101' I

No No l'

90SS281 93' I

No No

'90SS282 101' I

No No l

9055283 93' I

No.

No E

90SS284 101' I

No No l.

90SS285 100' I

No No

i 8

ISS302 35' I

No No i'

15S303 34' I

No No ISS305 201 I

No No 4

TABLE 3.7.5-1 (Continued)

SAFETY RELATED HYDRAULIC SNUBBERS

• SNUBBER

' SYSTEM SNUBBER INSTALLED.

ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT Q

NO.

ON, LOCATION AND ELEVATION INACCESSIBLE

' ZONE **

TO REMOVE 4

Residual Heat Removal Systen (Continued)

U 2 Ell-1SS306 Dnvell (Cont'd) 19' I

No No 84SS309 33' I

No No 84SS311' 35' I

No No 84SS312 35' I

No No

.8755315 33' I

No No 87SS317.

35' I

No No

-87SS318 35' I

No No n

90SS388.

79' I

No No 90SS389 96' I

No No

.90SS390 96' I

No No k

90SS391~

93' I

No No

'90SS392 99 I

No No

~7 2 Ell-131SS255 Reactor Building 42' A

No No 131SS257 22' A

No No-132SS263 30' A

No No l

l132SS264 31' A

No No 128SS355 42' A

No No 128SS387 43' A

No No 1325S414

-14' A

No No l 127SS426 43' A

No No No 128SS427 21' A

No

'128SS428 28' A

No No' 12855429 39' A

No No y

31' A

No No 128SS430

<=

A No No a

128SS431 23' 5

12755376 59' A

No No 3

8 I

E f

j 8

l I

I I

TABLE 3.7.5-1 (Centinued)-

{

SAFETY RELATED HYDRAULIC SNUBBERS *~

b SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY-DIFFICULT k-NO.

ON, LOCATION AND ELEVATION INACCESSIBLE-ZONE **

TO REMOVE Residual Heat Removal System (Continued) 2 M

2 Ell-107SS573 Reactor Bldg (Cont'd) 15' A

No No

-N' 127SS433 14' A

No No 127SS435 15' A

No No 17SS3

-3' A

No

.No 17SS4

-l' A

No No 20SS19

-3' A

No No 95SS20

-3' A

No' No 95SS22 9'

A No No 95SS23 9'

A No No w-3

-20SS28

-3' A

No No 37SS95

-3' A

No No y

A 35SS360

-8' A

No No m

35SS361

-16'-

A No.

No 2SS397

-3' A

No No SSS398.

-4' 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 k

8SS407

-15' A

No No

=

12SS408

-14' A

No No k

9SS411

-14' i

A No No a

109SS412

-14' A

No 66 20SS415

-4' A

No No 20SS416

-4' A

No No x

95SS417 0'

A No No E

95SS418

-4' A

No No 9

i

TABLE 3.7.5-1 (Continued)-

SAFETY RELATED HYDRAULIC SNUBBERS

• NO.

ON, LOCATION AND ELEVATION

. INACCESSIBLE ZONE **

TO REMOVE SNUBBER SYSTEM. SNUBBER INSTALLED.

ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT'

~Q-Residual Heat Removal System (Continued)'

2 U

'2E11-45SS422 Reactor Bldg (Cont'd)

-2' A

No No N

56SS517

-5' A

No No 91SS575 53' A

No No 91SS500 57' A

No No 91SS499 69' A

No No 89SS491-67' A

No No

~ 89S5489 67' A

No No-l

.89SS487 67' A

No No w

89SS480 67' A'

No No 2

18SS469 53' A

No No~

46SS216 28' A

No No y

A, 46SS217 31' A

No No 46SS218 30' A

No No 47SS223 33' A

No No 47SS224 36' A

No No 47SS225 '

36' A

No No 4755227 39' A

No No 47:6228 39' A

No No 95SS233 20' A

No No 95SS234 24' A

h' No No 95SS235 31' A

No

' No 21SS296 39' A

No 21SS297 39' A

No No g

473S326 42' A-No No

.a 475S328 42' A

No No A

No No R'

49SS330 42' 5

495S331 42' A

No No 49SS333-42' A

No No z

49S5334 43' A

No No i

TABLE 3.7.5-1 (Continued) co E

SAFETY RELATED llYDRAULIC SNUBBERS

• O R

SNUBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT' x

~N0.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **-

TO REMOVE-I' Residual Heat Removal System (Continued)

[

2 Ell-49SS336 ReactorBldg(Cont'd) 40'

_A No No 49SS359 42' A

No No 18SS470 43' A

No No 58SS514 14' A

No No-49SS515 37' A'

No No 4955516.

37' A

No No 50SS597 20' A.

No No 46SS7 12' A

No No R

46SS9 12' A

No

.No 56SSl3 5'

A No No

.?

56SSl5

-4' A

No No 0

58SS32 3'

A No No 58SS33

-4' A

No No 58SS35-8' A-No No 58SS36 8'

A No No

-18SS40 8'

A No No 18SS48 13' A

No No 68SS59

~

15' A

No No 21SS63 8'

A No No 21SS70 13' A

No No 21SS71 10' A

No No 61SS110 6'

A No No 53SS192 18' A

No No 53SS195

, 14' A

No No 53SS197 14' i

A No No

'k 53SS200 14' A

No No a

50SS201 14' A

No No R

89SS208-5' l

A No No A

do No E-60SS438 12' 5

s e

+g.

TABLE 3.7.5-1 (C*ntinued) k-SAFETY RELATED HYDRAULIC SNUBBERS

• z

- E SNUBBER SYSTEM SNUBBER INSTALLED' ACCESSIBLE OR HIGH RADIATION ESPECIALLY' DIFFICULT j

-NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE h-Residual Heat Removal' System (Continued) z U

2 Ell-60SS440 Reactor Bldg. (Cont'd) 13' A

No No 6555441:

3' A

No No

~

65SS442 3'

A-No No 60SS443 11' A

No No 73SS444 2'

A No No 21SS445 5'

A No No 68SS448 13' A

No No 75SS449 2'

A No No 61SS450 7'

A No No R

60SS451 13' A

No No 60SS452 13' A

No No

?

60SS453 10' A

No No 60SS454 10' A

No No 8955459 11' A

No No 89SS460 10' A

No-No 89SS461 6'

A No No 53SS462 15' A

No No 53SS463 14' A

No No 53SS464 14' A

No Ne 53SS465 14' A

No No c

.53SS466 14' A

No No 53SS467 14' A

No No 50SS468 17' A

No No k

56SS504 14' A

No No R

56SS505 7

A No No 3

56SS506 3'

A No No R

56SS507 3'

A No No 56S5508 4'

A No No x

46SS509' 8'

A No No 46SS510.

11' j

A No No o

.i 1

TABLE 3.7.5-1 (Continued)

E i

SE-

. SAFETY'RELATED HYDRAULIC SNUBBERS *'

5

.~

~ SNUBBER 1 SYSTEM SNUBBER-INSTALLED.

ACCESSIBLE OR HIGH RADIATION ESPECIALLY: DIFFICULT M.

' NO..

.ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **-

TO REMOVE'-

y Residual Heat Removal System (Continued)

[

2 Ell-46SS511 Reactor -Bldg. '(Cont'd)

-10' A

No-No.

'46SS512

-l' A

'No No?

~

No No

'68SS577:

8' A

53SS596 14'.

A'

-No No' 116SS143

-11' A-No-No

-113SS157

-11' A

No-No 37SS184 9'

'A No No '-

2SS396' 5'

A No.

No 116SS409

-9' A'

No

'No-113SS410

-9' A

No No

_w1 2SS413

' 0' A

No No t

'60SS423

-4' A

No No y.

k 60SS425

-2 '.

A No No.

47SS323 42' A-No:

No-71SS393 9'

A No No 127SS434 37' A

No' No 60SS437 13' A

No No 83SS446 10' A

No No 51SS546 32' A

No No 51SS547 28' A-No No 115S5549 31' A

ho No 4

185S47-12 '.

A Ne No' 71SS174

-17' A

No No l 9'

A No No 71SS176 54SS551 31' j

A-No No g

54SS552-28' A

No No E

98S5554 32' A

No No 2

58SS563 7'

A No No i

5 58SS565 13' A

No No 58SS566 6'

A No No 2

P 69SS574 6'

A No No E-i

TABLE 3.7.5-1 (Continued)

-SAFETY RELATED HYDRAULIC SNUBBERS

• Ep;-

SNUBBER

_ SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICUL1-E NO.

ON, LOCATION AND ELEVATION INACCESSIBLE ZONE **

TO REMOVE' i

E Service Water System

'M 2SW-133SS22 Reactor Building

.-6' A

No No2 m

110SS35

-5'-

A No

'No

-173SS72 14' A

- No No'

'1425S164 8'

A No

'No 142SSI.65 8'

.A No No 133SS176.

14' A

No No 133SS177

-5' A

No No 142SS74 40' A

No No 142SS75 40' A

No No w3 1405580 40' A

No No

-140SS86 45' A

No

.No u

L, 153SS102 44

A No No.

'o 153SS109 44'

'A No No 173SS110 48' A

No No

-153SSil5 44' A

No-No 103SS121 38' A

No No 140SS167 42' A

No No 173SS175 30' A

No No 142SS82 70'

.A No No 173SSil4 70' A

No No 103SS126 60' A

No No l

8' 8

i

'j

.E

d REFUELING OPERATIONS 3/4.9.2. INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least 2 source ~ ange monitor (SRM) channels

• shall be OPERABLE r

and inserted to the normal operation level with:

a.

A continuous visual indication in the control room, b.

One of the SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other SRM detector located in an adjacent quadrant, and The " shorting links" renoved from the RPS circuitry prior c.

to and during the time any control rod is withdrawn ** and shutdown margin demonstrations.

APPLICABILITY: CONDITION 5.

ACTION:

.C:

With the requirements of the above specification not satisfied, imme ~ ~

diately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

4. 9. 2 Each of the above required SRM channels shall be demonstrated OPERABLE by:

l a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; l

1.

Performance of a CHANNEL CHECK, l

• The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is-permissible as long as these j

special-detectors.are connected to the normal SRM circuits.

}{

• • Not required for control' rods removed per Specification 3.9.10.1 or

==

3.9.10.2.

4 BRUNSWICK - UNIT 2 3/4 9-3 Amendment No. 50

~

4 s

ets t

INSTRUMENTATION' SURVEILLANCE REQUIREMENTS (Continued) t

2. :- -Verifying the detectors are inserted to.the normal operat-ing level, and 3.

During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant

~

where CORE ALTERATIONS are being performed and one is

~1ocated -in the adjacent quadrant.

b.-

Performance of a CHANNEL FUNCTIONAL TEST:

1.

-Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of CORE ALTERATIONS, and-2.

At least once per 7 days.

Verifying at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.during CORE ALTERATIONS c.

that the. channel. count rate is at least-3 cps, wg E

T xy 9:

~~

. BRUNSWICK - UNIT ~1

' H I-W

-v A

I 3

k.:i; 1_.