ML20043A891

Proposed Tech Specs Re Limiting Conditions for Operation & Surveillance Requirements Re Rcs,Eccs & Containment Sys

ML20043A891
Person / Time
Site:	Brunswick
Issue date:	01/24/1990
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20043A868	List: ML20043A867 ML20043A891
References
NUDOCS 9005230245
Download: ML20043A891 (28)

Text

_ _ . . _ -- _ . _ _ ._. _ .

p. .  :

7: ._, . .

-4

~ 4 INDEX I 1

I

. LIMITING CONDITIOWS FOR OPERATION AND SURVEILLANCE REQUIREMENTS -  ;

i

, SECTION PACE l

3/4.4 REACTOR COOLANT SYSTEM (Continued) L 3/4,4.4 CHEMISTRY............................................... 3/4 4-7  !

3/4.4.5 SPECIFIC' ACTIVITY....................................... 3/4'4-10 3/4.4.6 PRESSURE / TEMPERATURE LIMITS  :

Reactor Coolant System................................. . 3/4 4-13 Reactor Steam Dome...................................... 3/4'4-21  :

L 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-22' l

3/4.4.8 STRUCTURAL INTECRITY.................................... 3/4 4-23  :

i

, 3/4.$ K_MERCENCY CORE COOLING SYSTEMS

i. 3/4.5.1 li!CW PRESSURE COOLANT INJECTION SYSTEM.................. 3/4 5-1  !

3/4.5.2 AUTOMATI C DEPRES SURIZATION S YSTEM. . . . . . . . . . . . . . . . . . . . . . . 3/4 5-3 3/4.5.3 thW FRESSURE COOLING SYSTEMS Core Spray System....................................... 3/4 5-4 $

Low Pressure Coolant Injection System.'.................. 3/4 5-7 3/4.5.4 SUPPBESSION P00L........................................ 3/4 5-9  !

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Inte$rity........................... 3/4 6-1 Primary Containment Leakage............................. 3/4 6-2 Pr imar y Co nt a i nmen t Ai r Lo ck . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-4 l

Primarf Containment Structural Integrity................ 3/4 6-6 Primary Containment Interns 1 Pressure...................- 3/4 6-7 Primary Containment Average Air Temperature............. 3/4 6-8 BRUNSWICK - UNIT 1 VI Amendment No.

l 9005230245 900124 PDR ADOCK 05000324 P PDC .

REACTOR COOLANT SYSTEM, 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITINC COWDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on (1) Figure 3.4.6.1-1 for heatup by non-nuclear means, cooldown following a nuclear shutdown, and low power -

PHYSICS TESTSI (2) Figure 3.4.6.1-2 for operations with a critical core other than low power PHYSICS TESTS or when the reactor vessel is vented; and (3) Figures 3.4.6.1-3a, 3.4.6.1-3b, or 3.4.6.1-3c, as applicable for inservice hydrostatir or leak testing, with

a. A maximum heatup of 100'F in any one-hour period, except for inservice hydrostatic or leak testing at which time the maximum heatup shall not exceed 30'F in any one-hour period.

b. A maximum cooldown of 100'F in any one-hour period except for inservice hydrostatic or leak testing at which time maximum cooldown shall not exceed 30*F in any one-hour period.

c. A maximum temperature change limited to 10*F in any one-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and- ,

d. The reactor vesp'1 flange and head flange temperatures greater than or equal to 70*F when reactor vessel head bolting studs are under tension.

APPLICABILITY: At all times.

ACTIOWs With any of the above limits exceeded, restore the tempera _ture and/or pressure to within the limits within 30 minutest perform an engineering evaluation to determine the ef fects of the out-of-limit condition on the f racture toughness properties of the reactor coolant systeal deterrine that the system remains j acceptable for continued operations, or 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 /> and in COLD SHUTDOWN within the r. ext 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1 SURVEILLANCE REQUIREMENTS '

4.4.6 1.1 The reactor coolant system temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system l heatup, cooldown, and inservice leak and hydrostatic testing operations.

BRUNSWICK - UNIT 1 3/4 4-13 Amendment No.

~

REACTOR C001. ANT SYSTEM SURVEILI.ANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-2 within 15 minutti, prior to the withdrawal of eantrol rods to bring the reactor to criticality.

4.4.6.1.3 The reactor material irradiation surveillance specimens shall be removed and examined to determine changes in material properties at the-intervals shown in Table 4.4.6.1.3-1. The results of these examinations shall be used to update Figures 3.4.6.1-1,-3.4.6.1-2, 3.4.6.1-3a, 3.4.6.1-3b, and 3.4.6.1-3c, as appilcable. The cumulative effective full power years shall be determined at least once per 18 months.

1 1

BRUNSWICK - UNIT 1 3/4 4-14 Amendment No.

q 8

. ~i l,'

' FIGURE 3.4.6.1-1  !

PRESSURE-TEMPERATURE LIMITS  ;

., REACTOR VESSEL NORMAL OPERATION WITH CORE NOT CRITICAL 1200 ,

i i E

1100 '

)

j w' i i I

=

1 '

j .

900 e

l '

H C  :

L

[

e00 r

y ,

l l -

a N '

I

• l

!, 700 "

O, ,

r O l r

l 600 j l l 1 ,

( m I  !

l

{ 500 --

j }

e r 400 'N '

l l I 300 l8 f  !

-$v z ,

200 & [ -

a' l .r 100 (u ,

r

._J 0

l 100- l 200 300 400 500 6bd -? i (70) (175)

TEMPERATUPE (* F) l l

BART&1

1. FUE4 IN REACTOR 2.

3. 116EFF{9 7.1 E 10 M > 1 MY 4 RT = $1.4Nf(1/4T)

5. 15MIINSTRUENTLOCATIONCORRECTIONINCLUDED

6. REO. QUIDE 1.99 REY. 2 E21T&1 1.

OPERATE TO RIGBT AND/OR BELOW LIMITING LINES

2.

• INDICATES BOTH EEATUP AND C00LDOWN RATE

3. j FRISSURE AND TEMPERATURE INTERSECTION 8 NOTED BT PARENTEE8E8 l BRUNSWICK - UNIT 1 3/4 4 15 Amendment No, l'

-_ . _ . . _ . _ . _ . , . _ _ _ . - - - - l

1 i

FIGURE 3.4.6.1 2 j PRESSURE. TEMPERATURE LIMITS i REACTOR VESSEL i

NOPJiAL OPERATION WITH CORE CRITICAL t .)

l]

1200 ,

ol l r  ;

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1000 ' I j

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j 900 '

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~

// / -V J / '

//1 fl/ FJ' / r 0 100 200 300 400 500 GOD  ;

(70) (181) l .

(210) TEMPERATVRE (* F)

&&&EL

- 1. FUEL IN RF.ACTC5t l

2. < 16 EFP

3. I.1X10{I NfCM > 1 MEV s l

4 RT = 81.4 (1/4 T)

S.

15MIINSTRt/ENTLOCAT10NCORRECTIONINCLUDED

6. REG. GUIDE 1.99 REY. 2 E2Ifa; i

1. '

OPERATE TO RIGHT /.ND/0R BELOW LIMITING LINES

2.

• INDICATES BOTH HEATUP AND C00LDOWN RATE

3. .

4 PRESSURE AND TIMPERATURE INTERSECTIONS NOTED BY PARINTHESES OPERATION IN CROSS-RATOHED AREA PERM"T*D ONLY WHEN WATER LEVEL I$ WITHIN NORMAL RANGE FOR POWER OPT *.'JON.

BRUNSWICK - UNIT 1 '4 4-16 Amendment No, e

s

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

L -* i

. t i - -

' ' FIGURE 3.4.6.li3a  !

• • PRESSURE TEMPERATURE LIMITS

• REACTOR VESSEL

\

HYDROSTATIC AND LEAK TESTS 1200 , , , , . , , , , . ,

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! 6 . , , I ' ,' i e i $ i . t i .

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t ii 1 . i 2

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(299) ' '

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. ll

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LJ ! i t  !

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i i e i I i 8 8 70 80 90 100 110 120 130' 140- 150 160 170 180 }

l TEMPERATURE t'F) I e

f

1. FUEL IN REACTOR -l;

2. REACTOR NOT CRITICAL

3. l!

4.

REG. QUIDE 1.99 REV. 2

< 8 EFFY .

il '

S. 3.S Y 10" N/ CHI > 1 MEV

6. RT = 86* (1/4 T)

7. ISMI IN8TRUDENT LOCATIDH CORRECTION INCLUDED l, M ,

1, CPERATE TO RIGHT AND/OR BELOW LIMITING LINES .

2.

• INDICATES BOTH HEATUP AND C00LDOWN RATE i t 4

i.

PRESSUR2 AND TEMPERATURE INTERSECTIONS NOTED EY PARENTHESES lI ,

OPERATING LIMIT INDICAIJS TEMPERATURE REQUIRED IF TEST PRESSURE WAS EXCEEDED. '

I t

BRUNSWICK - UNIT 1 . 3/4 4 17 Amendment No.- I' r

1

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e

-r - ,# + , w ~ ,v - - . , . -- , , - , ,~we,a, w ,~< +w -,*

a --~w--+ -,--sv---- -,-- ,v,

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PRESSURE TEMPERATURE L2MITS REACTOR VESSEL .

I in'DRO*,TATIC AND 1.EAK TESTS l i

1200 , , . , , , , ,,

i i i . , ., ,

, , , ,, i , , , , j i i4 , i i ... , , ,7 .

. , i . , i ,, , , , ii . ,

i >

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'a 1100

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// yi,.m, i

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I I i i  ! i g/ A e f i i

ij 1000 i

d ucs

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w .

s > ,

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p .

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.(450) ;l r

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6 6 i , , , 4 + i .

70 90 90 100 110 120 130 140 150 160 170 180 190  !

TEMPERATURE (*F) I 1

IAIEI.i I

1. FUEL IN REACTCk  !

2. i

'3. <10Ng'N/DI 4.4 X 10 > 1 MEV I 4 RT = 73' (1/4 T)  !

3. 14SIINSTRUMENTLOCATIONCORRICTIONINCLUDED -

6. REG. GUIDE 1,99 REV. 2 I 7 REACTOR NOT CRITICAL '

M 1, $

OPERATE 70 RIGHT AND/OR !ELOW LIMITING LINES '

2

• INDICATES BOTH EEATUP AND C00LDOWN RATE l 3. PRESSURE AND TDiPERATURE INTERSECTIONS NOTED BY PARENTHESES

, 4 .ti

!. OPERATING LIMIT INDICATES TEMPERATURE REQUIRED IF TEST PRESSURE WAS EXCEEDED.  ;

I BRUNSWICK - UNIT 1 3/4 4 18 Amendment No.

B

-, , . . + - - ,- ,, - - .c... , . . . . - , . . , , ~ . , - . ,

FZGURE 3.4.6.1-3c  !

PRESSURE TEMPERATURE LIM 2TS

• REACTOR VESSEL I i

l HYDROSTATIC AND LEAK TESTS  !

1200 , . , , . , , , . ,, , , , , , , , , , , ,

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m ___.

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, i E100 . ,

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(442) f e

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, i i y . . . i 200 -

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i . . . , , i . . ! i i , 6 i i t . t i ,

i . , , , , , i i i ri .

,,e i t i e 4 i + . . .t 'i 70 80 90 100 110 120 130 140 150 160 170 180 190 TEMPERATURE ('F1 I 16 EEL.

1. rUEL IN REACTOR I

2. 2,12EFF{7

3. l 3.3 X 10 N > 1 MEV '

4 RT = 76*(1/4 /Q1 f)  ?

5. 15NIINSTRUMENTLOCATIONCORRECTIONINCLUDED

6. REO. GUIDE 1.90 REV. 2 f

7. REACTOR NOT CRITICAL i

.  !!2IEL.

1. OFERATE TO RIGHT AND/OR BELOW LIMITING LINES i

2.

• INDICATES BOTH REATUI AND C00LDOWN R/!E 3.

FRESSE31 AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES 4

OPERATING LIMIT INDICATES TEMPERATURE REQUIF.ED IF TEST PRESSURE WAS EXrtE %

BRUNSVICK - UNIT 1 3/4 4 19 Amendment No. t

+

TABLE 4.4.6.1.3-1 l REACTOR VESSEL MATERI AL SURVEILLANCE PROGRAM CAPSULE-WITHDRAWAL SCHEDULE .

CAPSULE VESSEL WITHDRAWAL TIME (a)

_ NUMBER LOCATION (EFPY) 3 300' 8 2 120 (b) f 1 30 (b) l l

I*} The specimen thall'be withdrawn during refueling outage immediately f preceeding or following the specified withdrawal time. i (b) The schedule for removal of the second and third capsule shall be proposed l af ter the results of the first capsule have been evaluated.

8 b l

r

. 9

+

1  :

l.

1 P

I r

BRUNSWICK - UhiT 1 3/4 4-20 Amendment No. l 1

_, +..

i i

. 1 REACTOR COOLANT SYSTEM i

REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1020 psig.

APPLICABILITY CONDITION 1* and 2*.

l ACTIOW1 With the reactor steam dome pressure exceeding 1020 psis, reduce the pressure to less than 1020 psig within 13 minutes or 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 />. ,

l SURVEILLANCE REQUIREMENTS 4.4.6.2 The resctor 1,tearn dome pressure shall be verified to be less than  :

1020 psi; at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.  :

i t

.i

• Not applicable during anticipated transients, reactor isolation, er reactor t ri p.

BRUNSWICK - UNIT 1 3/4 4-21 Amendment No.

s REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION f

3.4.7 - Two Main Steam Line Isolation Vit- as (MSIV) per main steam line shall i be OPERABLE with closing times 3 3 and $ 5 seconds.  !

APPLICABILITY: CONDITIONS-1, 2, and 3.

ACTION:

With one or moro MSIVs inoperable, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that at least one MSIV is maintained OPERABLE in each af fected main steam line that is open and either

1. The inoperable valve (s) is restored to OPERABLE status within 8 '

hours, or 2.- The affected main steam line(s) is isolated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> by use of a deactivated MSIV in the closed position.

' 3 Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in '

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

• I SURVIELLANCE REQ'JIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated CPERABLE by verifying full closure between 3 and 5 seconds when tested pursuant to  ;

specification 4.0.5.

l l

l 1

i r

BRUNSWICK - UNIT 1 3/4 4-22 Amendment No. [

P

- , e q - , ,

REACTOR COOLANT SYSTEM 3/4.4.8 STRUCTURAL INTECRITY LIMITING CONDITION FOR OPERATION 3.4.8 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.8.

APPLICABILITY: CONDITIONS 1, 2, 3, 4, and 5.

ACTIONt

a. With the structural integrity of any ASME Code Class 1 components not conforming to the above requirements, restors the structural integrity of the affected component to within its limit or isolate; the affected component p'rior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by-NDT considerationr.

b. With the structural integrity of any ASME Code Class 2 croponent(s) not confirming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the aff ected component (s) prior to increasing the Reactor Coolant System temperature above 2120F.

c. With the structural integrity of any ASME Code Class 3, component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) within its limit or isolate the affected component (s) from service.

d. The provisions of Specificat.on 3.0.4 are not applicable.

e. The provisions of Specification 3.0.3 are not applicable in CONDITION 5.

SURVEILLANCE REQUIREMENTS 4.4.8 The structural integrity of ASME. Code Class 1, 2, and 3 components shall be demonstr ated per the requirements of specification 4.0.5.

BRUNSWICK - UNIT 1 3/4 4-23 Amendment No.

l u -

REACTOR COOLANT SYSTEM BASES

• The surveillance requirements provide adequate assurance that concentrations take in excess of the limits will be detected in suf ficient time to corrective action.

3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body duses.resulting from a main steam line failure outside the containment during steady state operrtion will not eseeed small fractions of the dose guidelines of 10 CFR 100. Permitting operation to continue for limited time periods with higher specific activity levels accommodates short-term iodine spikes which may be essociated with power level changes, and is based on the fact that a steam line failure during these short time periods is considerably less likely. Operation at the higher activity levels, therefore, is restricted to a small fraction of the unit's total operating time. The upper limit of coolant iodine concentration during short-term lodine spikes ensures that the thyroid dose from a steam line failure will not exceed 10 CFR Part 100 dose guidelines.

Information obtained on iodine spiking will be used to assess the parameters associated with spikirg phenomena. A reduction in freque. icy of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolation valves prevents the release of activity to the environs should the steam line rupture occur. The l surveillance requirements provide adequate assurance that ezcessive specific activity levels in the reactor coolant will be detected in sufficient time to ,

take cerrective action. ,

i 3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant-System are designed to withetand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introdeced by normal load transients, reactor trips, and start-up and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.2 of the FSAR. During start up and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produco_

thernal stresses which vary from cempressive at the inner wall to tensile at ,

the outer wall. Thermal-induced compressive stresses tend to-alleviate the tensile stresses induced by the internal-pressure. During cooldown, thermal gradients to be accounted for are tensile at the inner wall and compressive at the outer wall.

BRUNSWICX - UNIT 1 B-3/4 4-3 Amendment No.

l

• I REACTOR COOLANT SYSTEM ~!

BASES - '

PRESSURE /TEMPERATf1RE LIMITS (Continued) >

i RT The reactor vessel materials have been tested to determine their initial f

{ NDT. The results of these tests are shown in CE NEDO 24161. Reactor operation and resultant fast neutron, E>1 Mev, fluence will cause an increase l in the RTNDT. Therefore, an adjusted reference temperature, bared upon the fluence can be predicted using the proper revision of Regulatory Guide 1,19.- l l- The pressure / temperature limit curves Figures 3 4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3a through 3.4.6.1-3c include predicted adjustments for this shif t in '

l RTNDT at the end of indicated EFPY, as well as adjustments to account for-the  !

location of the pressure-sensing instruments.

I The actual shift in RTFDT of the vessel material will be checked l periodically during operatton by removing and evaluating, in accordance with  !

I ASTM E185-82, reactor vessel material irradiation surveillance specimens i

installed near the inside wall of the reactor vessel in the core area. Since- '

the neutron spectra at the irradiation samples and vessel inside radius vary ,

little, the measured transition shift'for a sample can be adjusted with l confidence to the adjacent section of the reactor vessel. j t

The pressure / temperature limit lines shown in Figures 3.4.6.1-1, 3.4.6.1-2, '

and 3.4.6.1-3a through 3.4.6.1-3c have been provided to assure compliance with the minimum temperature requirements of the 1983 revision to Appendix C of ,

i 10CFR$0. The conservative method of the Standard Review Plan has been used for heatup and cooldown. i The number of reactor vessel irradiation surveillance specimens and the  ;

frequencies for removing and testing these specimens are provided in  :

Ta bl e 4. 4. 6.1. 3-1 to assure compliance with the requirements of ASTM E185-82.  :

i i

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'I

. r BRUNSWICK - UNIT 1 B 3/4 4-4 Amendment No. (

i r

)

,A - - , - -

. . - . . ~

r i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PACE 3/4.4 REACTOR COOLANT SYSTEM (Continued) '

3/4.4.4 CHEMISTRY................................................ 3/4 4-7 3/4.4.5 S PE CI FI C ACTI VI TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-10 l l i l 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-13 l

Reactor Steam Dome.............-......................... '3/4 4-21 ;

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES......................... 3/4 4-22 I

i 3/4.4.8 S TR UCTU RA L I NTE GR I T Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-23 3/4.5 EMERCENCY CORE COOLING SYSTEMS i

3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM................... 3/4 5-1 ,

3/4.5.2- AUTOMATIC DEPRESSURIZATION SYSTEM........................ 3/4 5-3 3/4.5.3 LOW PRESSURE COOLING SYSTEMS Core Spray System........................................ 3/4 5-4 1

Low Pressure Coolant Injectian System.................... 3/45-7 1

3/4.5.4 SUPPRESSION P00L......................................... 3/4 5-9' 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT t t

P ri ma ry Con ta inme nt Int eg ri t y. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-1 P ri ma ry Con t a inment Lea ka ge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-2 Primary Containment Air Lock............................. 3/4 6-4 Primary Containment Structural' Integrity................. 3/4 6-6 Prima ry Conta inment Int ernal Pre s su re . . . . . . . . . . . . . . . . . . . . 3/4 6-7 Primary Containment Average Air Tempera ture . . . . . . . . . . . .'. . 3/4 6-8 BRUNSWICK - UNIT 2 VI Amendment No.

b

[ : i

., )

REACTOR COOLANT SYSTEM i 1.

3/4.4.6 PRESSURE / TEMPERATURE LIMITS l REACTOR COOLANT SYSTEM l  !

LIMITINC CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited '

in accordance with the limit lines shown on (1) Figure 3.4.6.1-1 for heatup by  !

l non-nuclear means, cooldown following a nuclear shutdown, and low power j PHYSICS TESTSI (2) Figure 3.4.6.1-2 for operations with a, critical core other l than low power PHYSICS TESTS or when the reactor . vessel. is vented; and l (3) Figures 3.4.6.1-3a, 3.4.6.1-3b, or 3.4.6.1-3c, as applicable f or inservice .

hydrostatic or leak testing, with:  ;

a. A maximum heatup of 100*F in any one-hour period, except for inservice hydrostatic or leak testing at which time the maximum e

l heatup shall not ;xceed 30*F in any one-hour period.

b. A maximum cooldown of 100'F in any one-hour period.except for l inservice hydrostatic or leak testing ti which time maximum cooldown i shall not exceed 'O'F in any one-hesr period.

c. A maximum temper 6 ture charge limited to 10*F in any one-hour period during inservice hydrostatic and leak testing operations above the i heatup and cooldown limit curves, and i

d. The reactor vessel flange and head flange temperatures greater than or equal to 70'F when reactor vessel head bolting-studs are under tension.

[

APPLICABILITY: At all times.

ACTION: -'

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutest perform an engineering evaluation to r determine the effects of the out-of-limit condition on the fracture toughness:

properties of the reactor coolant systeel determine-that the system remains acceptable for continued operations, or 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 /> and in COLD SHUTDOW!s within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS - 4.4.6.1.1 The reactor coolant system temperature c.d pressure shall be-i determined to be within the limits at least once per 30 minutes during system l '

heatup, cooldown, and inservice leak and hydrostatic testing operations.

I 1

1-BRUNSWICK - UNIT 2 ' 3/4 4-13 Amendment No.

l F 1

i i

+

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be l determined to be to the right of the criticality limit line of  !

Figure 3.4.6.1-2 within 15 minutes prior to the withdrawal of control rods to '

bring the reactor to criticality. .  ;

4.4.6.1.3 The reactor material irradiation surveillance specimens shall be removed and examined to determine changes in material properties at the  !

intervals shown in Table 4. 4.6.1.3-1. The results of these examinations shall i be used to update Figures 3. 4.6.1-1, 3.4.6.1-2, 3. 4. 6.1-3a , 3.4. 6.1-3b , and  :

3.4.6.1-3c,_ as applicable. The cumulative effective full power years shall be-determined at least once per 18 months. +

l i

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BRUNSWICK - UNIT 2 3/4 4-14 Amendment No.

4 h

?

IICURE 3.4.6.1 1

' PRESSURE.-TEMPERATURE L2MITS REACTOR VESSEL NORMAL OPERATION WITH CORE NOT CRITICAL 1200 _ , , , i., ,,. . , , .

, , , , , , , , -, . i

. . . r , , . i i . . ,

._ i 4 i , . -

6 *a i , I 6 l i 6 i 1 ,

i .

i.. ,

t $ t e i i , , 1 i i .

i

. i i 1100 , ,,,

i i r

. . i t ,i 4 i .

r

! 4i e +

! i 41

, ,6 4 1 ,

.) i ii i , .

1000 , , ,

! , I v

, ..ir , . . -

ii X e 1 i j ) ,

. I I

,  ! 6 j i 900 ,

i e i 8 4 4 i i i

i

.r .

a

' ' i 800 ' '

s' -g ,

, , ,i c  :

_ i + ri f 700 , ',, ,

• l l

I h-k ~

i 600 F L i . ,a d i ,

m (550)- ' i m { i

{ 500

, t

. 11 400 to j i i ,

F .

i J 300 , l q.' / ,

Q r 200 .p., ,

s v

.c

/\

t00 [w s

1 s

+

0 1 200 l 100 300- 400 500 600 (70) (170)

TEMPERATURE (* F)

BELT

1. TUEL IN REACTOR

2. 116EFF{y -

3. 7.1 X 10 gN/Of2 > 1 HEV

4. RT = 93 (1/4 f) 5.

6.

15$1 INSTRtMNT LOCATION CORRECTION INCLUDEDl E2I E REG. GUIDE 1.99 REV. 2

1.  !

OPERATE TO RIGHT AND/OR BELOW LIMITING LINES

2.

• INDICATES BOTH REATUF AND COOLDOWN RATE 3.

PRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES 3 BRUNSWICK UNIT 2 3/4 4 15 Amendment No,

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

F8GURE 3.4.6.1 2 PRESSUREoTEMPERATURE LIM 7TS

, . REACTOR VESSEL NORut OPERATION WITH CORE CRITICAL 1200 i i

1100 e-1 1000 l k'

~ ' ~ ~~

' ;[

.4_.,

1 r

g l 900 )

r, 1 o r O J

800 ,

I

[ n C

2 100 f y i Of I i W

I ~

l GOD j f

i l l ,

500 l J

400 g 4 -, ( --.

300 '

g 7 w ."

6 -

r /

J ,V i

/ 0l 200 )

)/ ) f .e s'r ) / .F 100 ,, , , _r ,

fs' ) / / /

d' ) / ,F a )

J' ) / / / J' .

A' ) / J a J L/

0 100 500 300. 400 500 600

. W3 (2 0) TEMPERN (' F')

.)

BonT&L

1. FUEL IN H ACTOR

2. .i 16 EFPl

3. 7.1 X l'/y N > 1 MEV '

,i 1

4. RT " 03' /CM (1/4 T)  !

5. 1$51INSTRlNENTLOCATIONCORRECTIONINCLUDED

6. l REO. fUIDE 1.99 REV. 2 f21L1.1  !

1. OPERATE To RIORT AND/0R BELOW LIMITING LINES i

2.

• INDICATES BOTE BEATUP AND C00LDOWN RATE ?l

3. PRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PC.fMTHESES

4. }

OPERATION IN CROSS RATCHED AREA PDMITTED ONLY WPIN WATER LEVEL IS WITEIN NORMAL RANGE FOR POWER OPERAT10h.

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

q

l.

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FIGURE 3.4.6 1-3a PRESSURE-TEMPERATURE I.IMITS c

REACTOR VESSEL

, HYDROSTATIC AND LEAK TESTS i 1200 t

l. 00 3

Y'M i is i, nan.

' i V.r  ! ,

r 4 ,

t000 '

{r ,

4. --. '

c , ,' .

w

.g ---- .s . , .

- - i . ,

goo

> 1 i 9 i i i

-+ , , ,

1. I M l f i m ._ - -,i .i + .

1 000 4 u w a >. ! 4

./ . i

-l f 4

. / I I I

m s .

t /a?C ) # t i~ 700 * ' '

, J, . ur , . , ,

9 g 3 am I 6 i . i fiog i . . .

, , e ,

i' 600 ' ipm '

i i ty 4

i i

m m'.g4' * ' '

s- 30 , .

rA 1 i . .

01 1 500 9' ScF i g '

I

{ 400 l i * -

s

' l

. 300 (298) i

_ aiom .1

-- % i j

200 ' I f

i. 100 L' '

+

i r ,

70 00 90 100 t10 120 130 140 150 160 170 t90 I

j TEMPERATURE ('FF 848141 i

1. FUIL IN REACTOR

2. REACTOR NOT CRITICAL

3. RIG QUIDE 1.99 REV. 2

4. $ 4 IFFY I

3. 3.3 X 10  !

l 2 > 1 HEV

6. AT = 77,N/CH (1/4 f)

(

7 L

szu_ 13MIINSTRLSGNTLOCATICNCCRRZCTICNINCLUDED 1.

CFERATI TO A!G8T AND/OR RELOW LIMITING LINE3 2.

• INDICATIS 3073 EEATUF AND C00LDOWN RATI

, 3. '

l 4 FRassunt AND TsMP RATUR INTERasCTIoNS NOTED BY FARENTIESIS

• l CPERATING LIMIT INDICATIS TIMPERATURE RICUIRED IT TEST FRKSSURE WAS IXCIEED. +

BRUNSWICK UNIT 2 . ,j 3/4 4-17 AmendmenC NE ' ' '

, . FIGURE 3.4.6.1 3b' '

' PRESSURE TEMPERATURE LIMITS  ;

REACTOR VESSEL l

HYDR 0 STATIC AND LEAK TESTS 1200 ,, ,  !

l

~~~

t100 y,8 , u- ,

tioas2 .r 1000 '

S

=  !

_n E *

~

" i' i---

900 .

] ,

= ,

.l. s 800 8

, r r O d'

-f -

4/d]) . i r I 700

'F

/ i e F

_dI j em JF h 600 '" e

p-,

i s00

9. Ip ,

7 g .

(440)- "'"~

,dC,\

f s 400 l fp i

1 '

L f 1

( ) -

P L AN3E I t

t 200 -j l

3r i 1

_k }

100 ',

1 t

t 70 00 90 100 110 120 .130 140 150 160; 170 180 190  ;

TID #ERATURE ' (' F) l nam >

1. PUEL IN REACTtst i

1. 510 Rg7 N 2 ,g ,gy

3. j 4.4 X 10 4 RT ' = 42' (1/4

/CM f)

5. 15 I INSTRUMENT LOCATION CORRECTION INCLUDED

6. {

RIG. GUIDE 1.30 REY. 2

7. REACTCat NOT CRITICAL i

)

E2ITL

1. OPERATE To RIGBT AND/OR BELOW LIMITING LINES

2.

• INDICATES BOTB HEATUP AND C00LDOWN RATE

3. PRES 5URE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTRESE8 i

i 4

OPERATING LIMIT INDICATES TEMPERATUkE REQUIRED IF TEST PRES $URE WA8 EXCEEDED. I

,t BRUNSUICK - UNIT 2 3/4 4 18 Amendment No.

[ 1 i

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

. __ ^ _ _ ..

r

- (

FZGURE 3.4.6.1 3c t

• PRESSURE TEMPERATURE LIMITS  :

REACTOR VESSEL i

HYDROSTATIC AND LEAK TESTS

• 1200 . , , ,

! .--=*

l l 8 r '

1100 '

ng;

&s#Nh d'

i 1 l][fT I

$ 2rF 1000

, s '

900 - "

r s -

u

\

800 "

1 m i

, r ,

. I e 'mmi fIU E 'I 3

? 700 ,,

/

~

i .

s- ,

600 ' '

s i ,r

• p+

a l 500 l

\ a.

s s i 40*

f8 "

i$

s .

f -- 0 9,b3 40 M ggC 1 7- .pg.C m

(EN) '

',_m i 200 $,

3 '

_d 100- ,

l i i

l l

. i 70 00 90 100 110 120 130 140 150 160 170- 180 190 TEMPERATURE ('F) '

R&Ilh ,

, 1. FUEL IN REACTOR I

2. I12EFF{y

3. 3.3 X 10 N > 1 MV 4 RT

• 86* (1/4

/DI f)  ;

$. 1$$1INSTRUMENTLOCATIONCORRECTIONINCLUDED  ;

6. REO. OUIDE 1.99 REY. 2 4 7 REACTCEL NOT CRITICAL M i

1. OPERATE To RIG 8T AND/OR BELOW LIMITING LINES l--

2.

• INDICATES BOTE BEATUP AND C00'J0WN RATE

3. i PRESSURI AND TDtPLRATURE INTERSECTIONS NOTED BY PARENTRE8E8 -

4 .

OPERATING LIMIT INDICATES TDfPERATURE REQUIRED IF TEST FRE88URE WAS EXCEEDED. s BRUNSWICK - UNIT 2 3/4 4 19 Amendment No, t

i

4 . 4 '

TABLE 4.4.6.1.3-1 i

REACTOR VESSEL MATERI AL SURVEILLANCE PROGRAM CAPSULE WITHDRAWAL SCHEDULE CAPSULE VESSEL- WITHDRAWAL TIME (a)

NUMBER LOCATION (EFPY) 3 300* 10 l

2 120 (b) r 1 30 (b)

(*) The specimen shall be withdrawn during refueling outage inmediately preceeding or following the specified withdrawal time.

(b) The schedule for removal of the second and third capsule shall be proposed af ter the results of the first capsule have been evaluated.

I b l

f 6

6

)

BRUNSWICK - UNIT 2 3/4 4-20 Amendment Wo.

l

.-.-c. - - .-w e -e . ,

. e

. ., j e

l REACTOR COOLANT SYSTEM '

! REACNR STEAM DONE ,

LIMITINC CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1020 psis.

l APPLICABILITY: CONDITION 1* and 2*.

l ACTION  !

With the reactor steam dome pressure exceeding 1020 psig, reduce the pressure l to less than 1020 peig within 15 minutes or be in at least HOT SHUTDOWN within '

i 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ,

SURVEILLANCE REQUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than l

1020 psig at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

t 4

i 4

• Not applicable during anticipated transients, reactor isolation, or reactor trip.

i BRUNSWICK - UNIT 2 3/4 4-21 Amendment No. l!

t

t

• i

• (

REACTOR COOLANT SYSTEM 1 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES ,

LIMITING CONDITION FOR OPERATION .

3.4.7 Two Main steam Line Isolation Valves (MSIV) per main steam line shall be OPERABLE with closing times > 3 and < 5 seconds. [

t APPLICABILITY: CONDITIONS 1, 2, and 3.

ACTION:

With one or more MSIVs inoperable, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that at least one MSIV is maintained OPERABLE in each affected main steam lins that is open and either

a. The inoperable valve (s) is restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or

2. The affected main steam line(s) is isolated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> by use of a deactivated MSIV in the closed position.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the-following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS .

1 4.4./ Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and 5 seconds when tested pursuant to Specification 4.0.5.

i BRUNSWICK - UNIT 2 3/4 4-22 Amendment No.

, ~.

.**y REACTOR COOLANT SYSTEM 3/4.4'8 STRUCTURAL INTECRITY l

LIMITING CONDITION FOR OPERATION 3.4.8 The structural integrity of~ASME Code Class 1, 2, and 3 components-shall' be maintained in accordance with Specification 4.4.8. 1 APPLICABILITY: CONDITIONS 1, 2, 3, 4,. and 5.

ACTION El

a. With the structural integrity of any ASNE. Code Class 1 components not- I conforming to the above requirements, restore the structural- ,

Integrity of the affected' component to within its limit, or isolate- '

i the affected component prior to increasing the' Reactor Coolant System '!

temperature more than 50*F above the minimum- temperature required. by i

. NDT considerations.

(

b. With thel structural int'egrity of any ASME Code Class 2' components (s) [

not comforming to the above requirements, restore the structueal1  ;

integrity.of the affected component to within its limit. or isolete >

the affected component (s) prior to Encreasing the Reactor Coolant System temperature above 212*F.

c. With the structural integrity of any 'ASME Code Class 3 components (s')'

not conforming to the 'above requirements, restore.the structural' 4

integrity of the affected component (s) within its limit, or isolate the affected component (s) from service.

d. The provisions lof Specification 3.0.4 are. not applicable. '

e. The provisions of Specification 3.0.3 .are not applicable in CONDITION 5.

l

[ SURVEILLANCE REQUIREMENTS l

4.4.8 The structural integrity of ASME Code Class 1, 2, and 3' components shall be' demonstrated per -the requirements of Specification 4.0.5.

l-

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I BRUNSWICK - UNIT 2' 3/4 4-23 Amendment No..

j C

i

, i v - - , - . - --- , +.Ls . - . + a .d. -

t I

RE/4 TOR COOLANT SYSTEM BASES The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.5 SPECIFIC ACTIVITY The limi:stior s on th* specific activity of the primary coolant

~ ensure that tne 2-hou. t hyroid and ahole body doses resulting f rom a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines in 10CFR 100. Permitting operation ta continue for limited time periods with higher specific activity levels

.eccommodates short-term iodine spikes which may be associated with power level changes, and is based on the fact that a steam line fa lure during these short i time periods is considerably less likely. O pe r& t i or. . . .ne higher activity levels, therefore, is restricted to a small fraction of the unit 's total operat ing time. The upper limit of coolant iodine concentration during short-term iodine spikts ensures that the thyriod dose from a steam line failure will not exceed 10 CPR Part lu0 dose guidelines.

Information obtained on iodine spiking will be used to assess the parameters ascociated with spiking phenomenao A reduction in frequency of isotopic analysis following power changes may be permissible, if justified by the data obtained.

Closing the main steam line isolation valves prevents t'he release of activity to tho environs should the steam line rupture occur. The surveillar,ce requirements provide adequate assurance that excessive specific activity levels in the reactor ccolant will be detected in sufficient t ine t o take corrective action.

3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant Syrtem are designed to wittstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor t rips, and start up and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.2 of the FSAR. During start up and shutdown, the rates of temperature and pressure changen are limised so that the maximum specified heatup and cooldown rateu ere consistent with tie design assumptions and satisiy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel well produce thermal stresses which vary from compressive at the inner wall to tensile at the outor wall. Thermally induced compressive stresses tend to alleviate the tensile streuses induced by the internal pressure. Juring cooldown, thermal gradients to be accounted for are tensile at the inner wall and compressive at the outer wall.

i BRUNSWICK - UNIT 2 B 3/4 4-3 Amendment No.

< ;a*  ;

.g d p*

es p  ;

ll I REACTOR COOLANT SYSTEM i

BASES I. t PRESSURE / TEMPERATURE LIMITS (Continued) 4 The reactor-vessel materials have been f.ested to determine. their initial l RTNDT.. The results of these tests are shown in CE NEDO 24161.- Reactor operation 'and resultant fast neutron, D1 Mev,- fluence will cause an increase in the-RTNDT. Therefore, an adjusted reference' temperature, based upon the ',

t fluence can be predicted using' the proper revision of Regulatory Guide l'.99.

The~ pressure / temperature limit curves Figures 3.4.6.1-1,'3.4.6.1-2,-and j 3.4.6.1-3a through 3.4.6.1-3c include predicted adjustments'for this shift in  :

RTNDT at the end of indicated EFPY, as well as adjustments to account for the location of the pressure-sensing instruments..

I The actual shift in RTE T of the vessel r.aterial will be checked periodically during operatton by removing and evaluating, -in accordance with <

AS1H E185-82, reactor vessel material irradiation surveillance' specimens '

installed near the inside. wall.of the reactor vessel'in the core area. Since the neutron spectra at'the irradiation samples and vessel inside radius vary.

Little,.the measured transition shift.for's sample can.be adjusted with  !

confidence to the adjacent section'of the reactor vessel.

The pressure /temperat'ure tirait lines shown in Figures 3.4.6.1-1, 3.4.6.1-2, _.

and 3.4.6.1-3a through 3.4.6.1-3c have been provided to assure compliance with- t '

the minimum temperature requirements of the 1983 revision ~to Appendix C'of 10CFR50. The conservative method of the Standard Review Plan has been used .

for heatup and cooldown.- .

The number'of reactor vessel irradiation surveillance' specimens and the  ;

frequencies for removing and testing .these, specimens are 'provided. in' Table 4.4.6.1.3-1 to assure compliance with the requirements-of ASTM E185-82.

t t

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

1 f

e

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I

._ . - . _ ,