Text

Proposed Tech Spec Section 3/4.4.6, Pressure/Temp Limits, Modifying Current Wording,Limiting Conditions for Operation & Pressure/Temp Limit Curves Consistent W/Current Regulatory Guidance
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ENCLOSURE 5 t

g.

BRUNSWICK STEAM ELECTRIC PIANT, UNITS 1 AND 2 NRC DOCKETS 50-325 & 50 324 OPERATINC LICENSES DPR-71 & DPR 62 REQUEST FOR LICENSE AMENDMENT REACTOR COOIANT SYSTEM PRESSURE / TEMPERATURE LIMITS TECHNICAL SPECIFICATION PAGES l

o 8811030364 881026 PDR ADOCK 05000324 P

PDC

(BSEP-1-129)

INDCX

^ -

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 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 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES........................

3/4 4-22 3/4.4.8 STRUCTURAL INTEGRITY....................................

3/4 4-23 3/4.5 EMERCENCY CORE COOLING SYSTEMS 3/4.5.1 HICH PRESSURE COOLANT INJECTION SYSTEM..................

3/4 5-1 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 Low Pressure Coolant Injection System...................

3/4 5-7 3/4.5.4 SUPPRESSION P00L........................................

3/4 5-9 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINHENT Primary Containment Integrity...........................

3/4 6-1 a

Primary Containment Leakage.............................

3/4 6-2

)

l Primary Containment Air Lock............................

3/4 6-4 Primary Containment Structural Integrity................

3/4 6-6 Primary Containment Internal Pressure...................

3/4 6-7

(

Primary Containment Average Air Temperature.............

3/4 6-8 4

1 l

}

i I

I BRUNSWICK - UNIT 1 VI Amendment No.

i 9

(BSEP-1-129)

, REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING 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 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 3.4.6.1-34, 3.4.6.1-3b, or 3.4.6.1-3c, as applicable for inservice hydrostatic or leak testing, with a.

A maximum heatup of 100'F in any one-hour parlod, 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 heacup and cooldown limit curves, and d.

The reactor vessel flange and head flange temperatures greater than or equal to 70*F when raaetor 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 determine the ef fects of the out-of-limit condition on the fracture toughness properties of the reactor coolant systems determine that the system remains j

acceptable for continued operations, or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

)

i SURVEILLANCE REQUIREMENTS 4.4.6.1.!

The reactor coolant system temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system i

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

1 4

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

1 I

(BSEP-1-129)

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 cf 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 be used to update Figures 3.4.6.1-1, 3.4.6.1-2, 3.4.6.1-34, 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.

i i

p 1

l BRUNSWICX - UNIT 1 3/4 4-14 Amendment No.

(BSEP 1 129)

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PRE *4WI AND TINTEJA!n! INTIR3ECTICNS NOTED SY PAAENTHI:13 BRL95'.'ICK UNIT 1 3/4 4 15 Amond. ment No.

b (BSEP-1 129)

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

(BSEP 1 129)

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BRUNSICK. UNIT 1 3/4 4 17 Amendment No.

(BSEP 1 129)

FICURE 3.4.6.1 L PRESSURE TEMPEP.ATUP,E LIMITS REACTOR VESSEL HYDROSTATIC AllD LEAK TESTS 1200

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4

}

(B3EP 1-129)

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ERUSSWICK UNIT 1 3/4 4-19 Amend:en: No.

(BSEP-1-129)

TABLE 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM CAPSULE WITHDRAWAL SCHEDULE CAPSULE VESSEL WITHDRAWAL TIME (a)

NUMBER LOCATION (EFPY) 3 300*

8 2

120' (b) 1 30' (b)

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

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

f BRUNSWICK - UNIT 1 3/4 4-20 Amendment No.

l

(35EP-1-129) s REACTOR COOLANT SYST']

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

APPLICASILITY:

CONDITION 1* and 2*.

ACTIONt With the reactor steam dome pressure exceeding 1020 psig, reduce the pressure to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1020 psig at least once par 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l

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

I 1

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

l

(3SEP-1-129) o REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two Main Steam Line Isolatior, Valves (MSIV) per main steam line shall be OPERABLE with closing times > 3 and < 5 seconds.

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 HSIV is maintained OPERABLE in each affected main steam line that is open and either 1.

The inoperable valve (s) is restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or 2.

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

Otherwise, be in at least HOT SHUTDOWN within the next 12' hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVIELLANCE REQUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPF.RABLE by verifying full closure between 3 and 5 seconds when tested pursuant to Specification 4.0.5.

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

l

(BSEP-1-129)

REACTOR COOLANT SYSTEM 3/4.4.8 STRUCTURAL INTEGRITY 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.

ACTION a.

With the structural integrity of any ASME Code Class 1 components not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component prior to increasing the Reactor Coolant System 0

temperature more than 50 F above the minimum temperature required by NDT considerations, b.

With the structural integrity of any ASME Code Class 2 component (s) not confirming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant 0

System temperature above 212 F.

c.

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

d.

The provisions of Specification 3.0.4 are not applicable.

e.

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

SURVEILLANCE REQUIREMENTS

\\

l 4.4.8 The structural integrity of ASME Code Class 1, 2, and 3 components l

shall be demonstrated 'per the requirements of specification 4.0.5.

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

(BSEP-1-129)

REACTOR 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 limitations on the specific activity of the primary ecol 4it ensure that the 2-hour thyroid and whole body doses resulting from a main steam line frilure outside the containment during steady state operation will not exceed 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 associated 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 iodine spikes ensures that the thyroid dose from a steam line failure will not exceed 10 CFR Part 100 dose guidelines.

t Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolai;on valves prevents t.ne release of activity to the environs should the steam line rupture occur. The surveillance requirements provide adequate assurance that excessive specific l

sativity levels in the reactor coolant will be detected in sufficient time to take corrective action.

3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced 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 i

design assumptions and satisfy the stress limits for cyclic operation.

i l

. During heatup, the thermal gradients in the reactor vessel wall produce t

thermal stresses which vary from compressive 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.

I r

i i

i BRUNSWICK - UNIT 1 B 3/4 4-3 Amendment No.

t

(BSEP-1-129)

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The reactor vessel materials have been tested to determine their initial RT The results of these tests are shown in CE NEDO 24161.

Reactor NDT.

operation and resultant fast neutron, E>l Mev, fluence will cause an increase in the RT Therefore, an adjusted reference temperature, based upon the NDT.

fluence can be predicted using the proper revision of Regulatory Guide 1.99.

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

The actual shift in RTtiDT of the vessel material will be checked periodically during operation by removing and evaluating. in accordance with ASTM 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 sapples and vessel inside radius vary little, the measured transition shift for a sample can be adjusted with confidence to the adjacent section of the reactor vessel.

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 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.

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

(BSEP-2-135)

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PACE 3/4.4.3 REACTOR COOLANT SYSTEM LEAKACE Leakage Detection System.................................

3/4 4-5 Operational Leakage......................................

3/4 4-6 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 3/4.4.7 KAIN STEAM LINE ISOLATION VALVES.........................

3/4 4-22 3/4.4.8 S TR U CTU RA L I NTEG R I TY.....................................

3/4 4-23 3/4.5 EMERCENCY CORE COOLING SYSTEMS 3/4.5.1 HICH PRESSURE COOLANT INJECTION SYSTEM...................

3/4 5-1 3/4.5.2 AUTOHATIC DEPRESSURIZATION SYSTEM........................

3/4 5-3 3/4.5.3 LOW PRESSURE COOLING SYSTEMS Core Spray System........................................

3/4 5-4 l

Low Pressure Coolant Injection System....................

3/4 5-7 l

l 3/4.5.4 SUPPRESSION P00L.........................................

3/4 5-9 1

l l

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity............................

3/4 6-1 Primary Containment Leakage..............................

3/4 6-2 Primary Containment Air Lock.............................

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

l l

7

(3SEP-2-135) c' 1

REACTOR C00_LANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING 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 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 ventedt and (3) Figures 3.4.6.1-3a, 3.4.6.1-3b, or 3.4.6.1-3c, as applicable for 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 j

heatup shall not exceed 30'F in any one-hour period.

r I

j 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.

i c..

A maximum temperature change limited to 10'F in any one-hour period l

during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d.

The reactor vessel flange and head flange temperatures greater than or equal to 70*F when reactor vessel head bolting studs are under l

tension.

l i

APPLICABILITY 1 At all times.

4 I

ACTION:

l With any of the above limits exceeded, restore the temperature and/or pressure f

to within the limits within 30 minutest perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the reactor coolant systeel determine that the system remains j

acceptable for continued operations, or be in at least HOT SHUTDOWN within 12 t

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

i i

SURVEILLANCE REQUIREMENTS i'

4.4.6.1.1 The reactor coolant system temperature and pressure shall be l

determined to be within the limits at least once per 30 minutes during system t

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

i J

.l l

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

a

(BSEP-2-135)

R2 ACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be l

determined tm 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 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.

i l

{

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

(BSEP 2 135)

FIGURE 3.4.6.1 1 PRESSURE TEMPERATURE LIMITS REACTOR VESSEL NOP&\\L OPERATION WITH CORE NOT CRITICAL 1200

. i p

i i

i I+

e

, i

l.

6 l

f.

6 i

i 99g

. i i

i i

t e

l i

i a

g, t

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6.

J i i,, i,

f 9

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e

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i

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i fi 4 6

i

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i

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pij t

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1 i,,, i p

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4

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(

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l t00 1 2N 300 400 500 600 (70)

(170)

TEMPERATURE

(* F) 1.iiLI.;.

1.

FUEL IN REACTCR 2.

116 ETPb 2 > 1 MIV 3.

7.1 X 10 O (CM N/

1/4 f) 4.

RTp. = 93 3.

1311! INSTMtCNT LOCATION CCM.ICTION INCLil0ED 6.

KIG. GtJIOE 1.99 REY. 2 1.

CTIAA!! TO 11011T A.0/CR BELOW LIMITING LINES 2.

ICICATES BCTH REAT'JP A.C COCL*XM KA!E 3.

TH 3'JRZ A C TEMFERA*31 INTERSIOTIONS NOTED BY FAKINTHISES l

l BRUNSWICK UNIT 2 3/4 4 15 Amend::ent No.

l

(BSEPi2-135)

FIGURE 3.4.6.1 2 PRESSURE TEMPERATURE LIMITS REACTOR VESSEL NORMAL OPE?ATION WITH CORE CRITICAL 1200

,,,.i..

l t._,

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se.

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ff f.f /1 f 6

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0 100 200 300 400 500 600 t2 53 gypeyg,RE

(' P'1 t

i L%.ll 1.

TUIL IN RIACTCR l

2.

< 16 EFrh f

3.

7.1 X 10 N/CNa > 1 MIV

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93' (1/4 T)

RT"$$1INSTRLHENTLCCAT:

4.

15 N CCFAECTION INCLtl0ED 1

5.

6.

REG. GUI E 1.99 RIV. 2 i

l 1

mLt 1

1.

CFERA!! TO RIGHT AN0/CR 3.1.CW LIMITING LINES I

2.

lNOICATES BCT!! EEATUP AhJ COCLD W N RATE l

3.

FRES3URE AND TIFFERATURE INTIRSECTIONS NOTED BY FAKINTHESES I

4 CFERATION IN CROS3-MATCMID ARIA FEMITTED CNLY WMEN WATER LIVEL l

rs wria!N NemAL RANcE rcR twtR CtERATION BRUNSWICK - UNIT 2 3/4 4 16 Amendment No.

7,_ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _

4 o

(BSEP 2 135)

FIGURE 3.6.6.1 3a PRESSURE-TEMPERATURE LIMITS REACTOR VESSEL HYDROSTATIC AND LEAK TESTS 1200

.. i i

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i t

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.vf

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1000 in 1

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,s I, ii 70 80 90 100 110 120 120 140 150 160 170 100 TEVPERATL'RE (' F)

IAILL.

t, tvE1. In azA TcR 1.

RI. ACTOR NOT CRITICAL 3.

RIO CUICE 1.99 KIV. 2 4

< s ETPY 3.

3.5 X 10 N/CM2 > 1 MIV 17 6.

RTp, = 77e (1/4 f) 7, 13 I$1 INSTKtt"TNT LOCATION COFJtECTION INCLV0ED EIIL 1.

OttRATE TO RIOHT AND/CR IE1.0W LIMITING LINES 2.

INDICATES SOTH t!IATUF AND COCL:O.M RATE 3.

FTISSURE A.ND TIMPERATVRI INTERSECT!CNS MCTIO SY FARINTHE E3 6

cFERATINc L:n:7 :.ctCATEs TerERAT;u KI;utRro tr TEsf FMSSURI WAS EXCIEOEO.

BRUNS'.IICK - UNIT 2 3/4 4 17 Amendment No.

~

FIGURE 3.6.6.1 3b PRESSURE-TEMPERATURE LIMITS 5

REACTOR VESSEL HYDROSTATIC AND LEAK TESTS 1200 i.

6 t

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70 80 90 100 tto 120 130 t40 150 1G0 170 180 190 TEMPERATURE

(' P"1 LtI.L1 1.

TVEL IN RIACTCR ItoErr{,N/CM'>1MIV 2.

3.

4.4 X 10 4.

RT

= 82e (1/4 f) 15NI INSTRLHINT LOCATION CCFJLECTICN INCL!,';ED 8.

RIO. G'J!tt 1.99 RIV 2 7.

RT. ACTOR NCT CRI!! CAL

50. ilk 1.

CFIRATI TO RIGHT MD/CR BELCW LIMITING LINES 2.

IN0! CATES 5073 KIAftlP AC CCCL.*WN RATE 3.

PRISSL'R.E MD TEMIERAIt'RI INTERSECTICNS NCTID BY FARINTHISES 4

CF!1ATING LIMIT 11;OICATES TIMFIRATL1LI RIOU! KID IT TEST F1tESSL*RI WAS EXCIIOED.

BRUNS'JICK - UNIT 2 3/4 4-18 Amendment No.

(BSEP 2 135)

FICURE 3.4.6.1 3c PRESSURE TEMPERATURE LIMITS REACTOR VESSEL l

[

1 HYDROSTATIC AND LEAK TESTS i

l i

t 1200 i_

6...

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t

. i.

i.

.i

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.,i 70 80 90 100 110 120 130 140 150 160 170 100 190 TEuPERATL'RE ('F1 l

LSIIla 1

1.

TVEL IN N J 2.

112EFr{y 3.

3.3 X 10 N/c# > 1 HIV 4

RT

= al' (1/4 f) 5.

15hIIINsTxMNTLOCATIONCORAZOTIONINCLUOID 6.

KEG. GUIOE 1.99 RIV 2 7.

RIA0 TOR NOT CRITICAL

!!2III.

1.

CTERATE TO RIG 87 M3/CR IILN LIMUNG LINES 2.

+ In:I:strs scia atArvr Aso C cLoob.y RAtt 3.

ratssrR.t ne itxrtxAr;tt InittsrcTI:ss netts er rAxrwrursts 4.

CrtxATI m LIMIT Iso!: Arts trartxAnu trou!R.to Ir its? FREssURI WAs IX;EIOED BRUNSWICK - UNIT 2 3/4 4-19 Amendment No.

(33EP-2-135) e O

TABLE 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROCRAM CAPSULE WITHDRAWAL SCHEDULE CAPSULE VESSEL WITHDRAWAL TIME (a)

NUMBER LOCATION (EFPY) 3 300' 10 2

120 (b) 1 30 (b)

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

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

P l

i i:

BRUNSWICK - UNIT 2 3/4 4-20 Amendment No.

l t

~ _ - -.

(33EP-2-135) e C

REACTOR COOLANT SYSTEM 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*.

ACTION!

With the reactor steam dome pressure exceeding 1020 psig, reduce the pressure to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1020 psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

BRUNSWICK - UNIT 2 3/4 4-21 Amendment No.

l

(3SEF-2-135)

O l

REACTOR COOLANT SYSTEM 3/4.4.7 HAIN STEAM LINE ISOLATION VALVES LIMITINC 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.

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 line that is open and eithert 4.

The inoperable valve (n) is restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or 2.

The affected main steam line(s) is isolated within 8 hcurs 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying fu11 closure between 3 and 5 seconds when tested pursuant to Specification 4.0.3.

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(BSIP-2-135) e o

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 4.4.3.

APPLICABILITY 1 CCNDITIONS 1, 2, 3, 4, and 5.

ACTIONI With the structural integrity of any ASME Code Class I components not a.

conforming to the above requirements, restore the structural integrity of the affected component to within its limit, or isolate the affected component prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum temperature required by NDT considerations.

With the structural integrity of any ASME Code Class 2 components (s) b.

not comforming to the above requirements, restore the structural integrity of the affected component to within its limit, or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 212*F.

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

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 Specification 3.0.4 are not applicable.

The provisions of Specification 3.0.3 are not applicable in e.

L CONDITION 5.

SURVE!LLANCE REQUIREMENTS 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.

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BRUNSWICK - UNIT 2 3/4 4 23 g,ng go, l

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(BSEP-2-135) o REACTOR 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 limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during ste:dy state operation will not exceed small fractions of the dose guidelines in 10CFR 100.

Permitting operation to continue for limited time periods with higher specitic activity levels accommodates short-term iodine spikes which may be associated 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 restric'.ed to a small fraction of the unit's total operating time. The upper limit of coolant iodine concentration during short-term iodine spikes ensures that the thyriod dose from a sesam 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 spiking phenomena. 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 the release of activity to the environs should the steam line rupture occur. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure chr.nges. These cyclic loads are introduced 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 produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. Thermally induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. During cooldown, thermal gradients tc be accounted for are tensile at the inner wall and compressive at the outer wall.

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(BSEP-2-135)

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE I.IMITS (Continued)

The reactor vessel materials have been tested to determine their initial RTNDT. The results of these tests are shown in GE NEDO 24161.

Reactor operation and resultant fast neutron, E)1 Mev, fluence will cause an increase in the RT Therefore, an adjusted reference temperature, based upon the NDT.

fluence can be predicted using the proper revision of Regulatory Culde 1.99.

The pressure / temperature limit curves Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-32 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.

The actual shift in RTNDT of the vessel material will be checked periodically during operation by removing and evaluating, in accordance with ASTM 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 little, the measured transition shift for a sample can be adjusted withthe ir confidence to the adjacent section of the reactor vessel.

The pressure /temperaturc 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 10CFa50.

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.

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l BRUNSWICK - 1) NIT 2 B 3/4 4-4 Amendment No.

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ML20205M941

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