ML20011F729

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Amends 140 & 172 to Licenses DPR-71 & DPR-62,respectively, Revising Tech Spec Section 3/4 4.6, Pressure/Temp Limits, to Modify Current Wording,Limiting Conditions for Operation & Pressure/Temp Limit Curves for Consistency
ML20011F729
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 02/15/1990
From: Tourigny E
Office of Nuclear Reactor Regulation
To:
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ML20011F730 List:
References
NUDOCS 9003070329
Download: ML20011F729 (34)
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{{#Wiki_filter:43 CIC 'o, UNITED STATES , _.f5 "1 NUCLEAR REGULATORY COMMISSION 3,,, s. I ,e WASHINGTON D. C. 20555 o \\,....,o# CAROLINA POWER & LIGHT COMPANY, et al. DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT l' AMENDMENT TO FACILITY OPERATING LICENSE i Amendment No.140 License No. DPR-71 1. The Nuclear Regulatory Commission (the Connission) has found that: A. The application for anendment filed by Carolina Power & Light Company (the licensee), dated October 26, 1988, as supplemented March 30, 1989, June 13, 1989 and August 4,1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),andtheCommission'srulesandregulationssetforthin10CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (1)thattheactivitiesauthorized 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 Consission'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. L 2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; i and saragraph 2.C.(2) of Facility Operating License No. DPR-71 is t hereay amended to read as follows: 1 i 9003070329 900215 3 l gDR ADOCKOSOOg4 g- .w.

./ i-L i ; (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.140. are hereby incorporated in the license-Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days of issuance. I FOR THE NUCLEAR REGULATORY COMMISSION Edmond G. Tourigry, Acting Director Project Directorate 11-1 Division of Reactor Projects - 1/II Office of Nuclear Reactor P,egulation

Attachment:

Changes to the Technical Specifications Date of Issuance: February 15, 1990 I i 0FC tLA:f ^h RPR:PM:PD21:DRPR: OGC

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r /TTACHMENT T0 LICENSE AMENDMENT NO. 140 FACILITY OPERATING LICENSE NO.' DPR-71 DOCKET 1 HO. 50-325 Replace the following ' ages of the Appendix A Technical Specifications with the enclosed pages. T1e revised areas are indicated by marginal lines. Remove Pages Insert Pages VI 3/4 4-13. VI 3/4 4-13 3/4 4-14 3/4 4-14 3/4 4-15 3/4 4-15 3/4 4-16 3/4 4-16 3/4 4-17 3/4 4-17 3/4 4-18 3/4 4-18 3/4 4-19 3/4 4-19 3/4 4-20 3/4 4-20 3/4 4-21 3/4 4 - 3/4 4-23 B3/4 4-3 B3/4 4-3 B3/4 4-4 B3/4 4-4

1 INDEX i LIMITINC CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS - I SECTION PACE 3/4.4 REACTOR COOLANT SYSTEM-(Continued) 3/4.4.4 C H EM I S TR Y............................................... 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 R e a c t o r S t e a m Do me...................................... 3/4 4-21 3/4.4.7 HAIN STEAM LINE ISOLATION VALVES........................ 3/4 4-22 3/4.4.8 ST RU CTUR AL I NT EG R I TY.................................... 3/4 4-23 3/4.5 - EMERCENCY CORE COOLINC SYSTEMS 3/4.5.1 HICH 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 I 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 CONTAINMENT t

Primary Containment Integrity...........................

3/4 6-1 Primary Containment Leakage............................. 3/4 6-2 Prinary 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 BRUNSWICK - UNIT 1 VI Amendment No.140

F' ? ^ 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) rigures 3.4.6.1-3a, 3.4.6.1-3b, or 3.4.6.1-3c, as applicable for inservice r hydrostatic or leak testing, with 4 t A maximum heatup of 100*F in any one-hour period, except for a. ' 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. l A maximum temperature change limited to 10'F in any one-hour period c. during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d. The reactor vessel flange and head flange temperattres 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 = determine. the ef fects of the out-of-limit condition on the f racture toughness properties of the reactor coolant system; determine that the system remains acceptable for continued operations, or be in at least HOT SHUTDOWN within 12 hours' and in COLD SHUTDOWN within the next 24 hours. SURVEILLANCE REQUIREMENTS 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 . heatup, cooldown, and inservice leak and hydrostatic testing operations. BRUNSWICK - UNIT 1 3/4 4-13 Amendment No. 77/. 140

a a v. REACTOR COOLANT SYSTEM SURVEILLANCE 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 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. BRUNSWICK - UNIT 1 3/4 4-14 Amendment No.140

e l FIGURE 3.4.6.1 1 PRESSURE TEMPERATURE LIMITS REACTOR VESSEL NORMAL OPERATION WITH CORE NOT CRITICAL-1200 i I ~ 1100-r j k l I 1000 I o 900 c N 900 cX r m J d 700 k e b~ 7 Ir 600 l~ I i us r -{ { 500 [ I (. r 400 f ) i f l 300 f .r, 200 ( [ b A [ 100 [ u E .J L 0 l 100 l 200 300 400 500 600-i (70) (175) TEMPERATUFE (* F) IMiLt 1. TVEL IN REACTOR 2. < 16 EFP I.1 X 10{ NfCH 3. > 1 MEV 4 RT = 81.4 (1/4 T) ISNIINSTRUMENTLOCATIONCORRECTIONINCLUDED 5. 5. RIO, GUIDE 1.99 REY. 2 E91Lh 1. OPERATE TO RIGHT AND/OR BELOW LIMITING LINES 2.

  • INDICATES BOTH HEATUP AND C00LDOWN RATE 3.

PRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES BRUNSWICK - UNIT 1 3/4 4-15 Amendment No. 140

4 FIGURE 3.4.6.1-2 PRESSURE TEMPERATURE LIMITS REACTOR VESSEL NORMAL OPERATION WITH CORE CRITICAL i 1200 I I i f I f i i 1100 1 I e I = I 1000 I k i { 900 r B00 g l C h I ~ ? 700 j j w k E (660) f I O 600 w 500 J l r r 400 N. ).--. 300 [ / ,,jf r s 200 / F / f 2 / r 2' 100 // / / / // / V J/ .f V / a J 1A/ / V ~ / l A/T / r i / r p a 0 100 200 300 400 500 GOO (70) (181) l (210) TEMPERATURE (*F) 161illL 1. FUEL IN REACTOR 2. <16EFPg 3 7.1 X 10 NfCM>1MEV 4 RT

  • 61.4 (1/4 T) 5.

15YkIINSTRUMENTLOCATIONCORRECTIONINCLUDED 6. RIG. GUIDE 1.99 REV. 2 M OPERATE TO RIGHT AND/OR BELOW LIMITING LINES 2.

  • INDICATES BOTH HEATUP AND COOLDOWN RATE 3

FRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES 4 CPERATION IN CROSS-MATCHID AREA PERMITTED ONLY WHEN WATER LEVEL IS WITHIN NORMAL RANGE FOR POWER OPERATION. BRUNSWICK - UNIT 1 3/4 4-16 Amendment No. 140

4 FIGURE 3.4.6.1-3m PRESSURE TEMPERATURE LIMITS REACTOR VESSEL HYDROSTATIC AND LEAK TESTS 1200 ii 1 i i > - i i.,

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  • INDICATES BOTH HEATUP AND COOLDOWN RATE 3.

PRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES 4 CPERATING LIMIT INDICATES TEMPERATURE REQUIRED IF TEST PRESS'JRE WAS EXCEEDIO BRUNSWICK - UNIT 1 3/4 6-17 Amendment No. 777,140 l,

,y FIGURE 3.4,6.1 3b PRESSURE TEMPERATURE LIMITS

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REACTOR VESSEL HYDROSTATIC AND LEAK TESTS 1200 -1 ii. , 4,. ,.i. 4 . i.. ... 4 , i, i i i i i i i. . i,< 4 i i i - i b 4 1. 4 i i I i i i e i.. 4 t i l . i ). i 4 i 1. 6 i i i i t. !. i / 4 i i. i. i.s , 6 3 , v / t, uw i..i i i iei i i i .. i i gi ) i, ,,i+ 1 i i .,.i + i i .. i # f = i... i i I i i 8 i i e if i i t 1000 . i u .ii i i i i. + e--7 e 4 ,i i+ i e c 900 l l l, l l 3 l' a i ,4 i a a i i> i ,i w , i 800 +.' . r' f it s n si r i l 3 i J i J ! i i i /f 3 ) i i it 700 i i + ...e i i i .r-l 6 jr i 4 a w ar i 600 l i S,gI.' l, f i s-i i i i i s I v r ! ' ) l .i w 500 f \\@ l 0 (450) gc8 r 400 j l i s i i i i i i u a i 1 i . i 300 (299) l r _h .i. i 4. 4 i i ' i i ',',l 200 i A, 100 t I i i i i i i . i 70 80 90 100 110 120 130-140 150 160 170 180 190 TEMPERATURE (' F) EARL 1. FUEL IN REACTOR 2. < 10 ETP{ N/CM 3. 4.4 X 10 > 1 MEV 73' (1/4 T) 4 RT = 5, 15$1INSTRUMENTLOCATIONCORRECTIONINCLUDED 6. RIG. GUIDE 1.99 REY. 2 7. REACTOR NOT CRITICAL I!9.UL 1. OPERATE TO RIGHT AND/OR BELOW LIMITING LINES 2.

  • INDICATTS BOTH HEATUP AND COOLDOWN RATE 3.

PRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES 4 OPERATING LIMIT INDICATES TDiPERATURE REQUIRED IF TEST PRIESURE WAS EXCEEDED. BRUNSWICK - UNIT 1 3/4 4-18 Amendment No. 77,140

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  • INDICATES BOTH HEATUP AND COCLDOWN RATE 2

PRESS' IRE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES CPERATING LIMIT INDICATES TEMPERATURE REQUIRED IF TEST PRESSURE WAS EXCEE0ED BRUNSWICK - UNIT 1 3/4 4 19 Amendment No.140

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d 3.- s t . TABLE 4.4.6.1.3-1 => REACTOR VESSEL MATERI AL SURVEILLANCE PROGRAM CAPSULE WITHDRAWAL SCHEDULE CAPSULE: ' VESSEL WITHDRAWAL TIME (a) -NUMBER LOCATION' (EFPY) 3 300' 8 4 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. 't Y I 1 0 BRUNSWICK - UNIT 1 3/4 4-20 Amendment No. 140 l

z,. 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 J* 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 hours. 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 hours.

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

BRUNSWICK - UNIT 1 3/4 4-21 Amendment No.140 f

. i. ^ REACTOR COOLANT SYSTEM -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 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 af fected main steam line that is open and eithers. 1. The inoperable valve (s) is restored to OPERABLE ' status within B hcurs, or 2.. The affected main steam line(s) is isolated within 8 hours 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 hours. SURVIELLANCE REQUIREMENTS

4. 4. 7.

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. BRUNSWICK - UNIT 1 3/4 4-22 Amendment No.140 l

_, i. 'i e' 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. l APPLICABILITY: CONDITIONS 1, 2, 3, 4, and 5. I ACTION With.the structural integrity of any ASME Code Class 1 components not. a. conforming to the above requirements, restore the structural integrity of the af fected 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, With the structural integrity of any ASME Code Class 3 component (s) c. i not conforming to the above requirement s, restore.the structural. integrity of the af fected component (s) within its limit or isolate the affected component (s) from service.. l d. The provisions of Specification 3.0.4 are not applicable. 1 The provisions of Specification 3.0.3 are not applicable in e. J CONDITION 5. J l-l SURVEILLANCE REQUIREMENTS l l 4.4.8 The structural integrity of ASME Code Class 1, 2, and 3 components shall be demonstusted per the requirements of specification 4.0.5. l I' BRUNSWICK - UNIT 1 3/4 4-23 Amendment No.140 l

C ,,[ m e ,b ( w..

REACTOR COOLANT SYSTEM-b~

BASES 1 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 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: accomodates short-term iodine spikes which may be associated with power level 4 changes, and is based on the f act 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 l operating r.ime. The upper limit of coolant iodine concentration during short-term iodine spikes ensures that the thyroid dose f rom 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 spiking phenomena. A reduction in frequency of l-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 l 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 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 design assumptions and satisfy the stress limits for cyclic operation. During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which varf 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 l-gradients to be accounted for are tensile at the inner wall and compressive at I the outer wall. i l I l: l-l' BRUNSWICK - UNIT 1 B 3/4 4-3 Amendment No. 777,140 t

7-r, C a ~ c LREACTOR 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 f 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 RTUDT 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 samples and vessel inside radius vary ' little, the measured transition shif t for a sample can be adjusted with ccnfidence 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 l the minimum temperature requirements of the 1983 revision to Appendix C of i -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 l-Table 4.4.6.1.3-1 to assure compliance with the requirements of ASTM E185-82. l 1- + l -1 1 + l BRUNSWICK - UNIT 1 B 3/4 4-4 Amendment No.140

p, 3 gf 3,,; . [...., UNITED STATES s - * - i ' 4 'i ij'; NUCLEAR REGULATORY COMMISSION T? .u E E., f'. WASHINGTCN, D C. 7055$ 1 e CAROLINA POWER & LIGHT COMPANY, et al. \\ DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE 4 Amendment 110. 1 72 License No. OPR-62 e 1. The Nuclear Regulatory Commission (the Comission) has found that: A. The application for amendment filed by Carolina Power & Light Company (thelicensee),datedOctober 26, 1988, as supplemented March 30,- 1989, June 13, 1989 and August 4,1989, complies with the standards and requirements of the Atomic Eneray Act of 1954, as amended (the - Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; ' 8. The facility will operate in conformity with the application, the _ provisions of-the Act, and the rules'and regulations of the Commission; C. There is reasonable assurance: (1) 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 Technica3 Specifications as indicated in the attachment to this license amendment; i; and paragraph 2.C.(2) of Facility Operating License No. DPR-62 is hereby amended to read as folicws:

y ' 5. c n 2 (2) Technical Specifications -The Technical Specifications contained in Appendices A and B, as revised through Amendment No.172, are hereby incorporated in the . license. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications. 3.- This license amendment is effective as of. the date of its issuance and shall be implemented within 60 days'of issuance. FOR THE NUCLEAR REGULATORY COMMISSION 1 Edmond G. Tourigny, Acting Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

c Changes to the Technical ,I Specifications Date of Issuance: February 15, 1990 i l i '0FC :LA i ((^PR:PH:PD21:DRPR: OGC

(A)D:PD21:DRPR

] ....:........,p,_:............:.. ,9WF' ny lNAME: P, n :NLe: bid 1 W: a l.....:............:............:.pJP._______:_7_.......__:............:............:........... r DATE : 7//f/90

n//1 /90
1-//L/90
/-/ //90 l

I 0FFICIAL RECORD COPY l

p ATTACHMENT TO LICENSE Af tENDMENT fl0.172 FACILITY OPERATING LICEllSE N0. OPR-62 DOCKET NO. 50-324 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.- The revised areas are indicated by marginal lines. Reniove Pages Insert Pages VI VI 3/4 4 3/4 4-13 3/4 4-14 3/4 4-14 3/4 4-15 3/4 4-15 3/4 4 3/4 4-16 3/4 4-17 3/4 4 3/4 4-18 3/4 4-18 3/4 4 3/4 4-19 3/4 4-20 3/4 4-20 3/4 4-21 3/4 4-22 3/4 4-23 B3/4 4-3 B3/4 4-3 B3/4,4-4 83/4 4-4 i L 4 a

t "e= INDEX ' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 4 SECTION PACE i-3/4.4 REACTOR COOLANT SYSTEM (Continued) 3/4.4.4 CHEMISTRY................................................ 3/4 4-7 1 ii f 3/4.4.5 SPECIFIC ACTIVITY........................................ 3/4 4-10 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-13 j Reactor Steam Dome....................................... '3/4 4-21 i 3/4.4.7 M AI N STEAM LI NE I SO LATION V ALVE S......................... 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 1 3/4.5.4 SUPPRESSION PO0L......................................... 3/4 5-9 3/4.6 CONTAINHENT SYSTEMS 3/4.6.1 PRIMARY CONTAINHENT P ri ma ry Con ta inment Int eg ri t y............................ 3/4 6-1 + P rima ry Conta inment Leaka ge.............................. 3/4-6-2 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 BRUNSWICK - UNIT 2 VI Amendment No. Jp),172

4 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 f or heatup by non-nuclear means, cooldewn following a nuclear shutdown, and low power PHYSICS TESTSt (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. hydrostatic or' leak testing, with A maximum heatup of 100*F in any one-hour period, except for a. 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 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 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the reactor coolant system; determine that the system remains acceptable for continued operations, or be in at-least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. SURVEILLANCE REQUIREMENTS 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 heetup, cooldown, and inservice leak and hydrostatic testing operations. BRUNSWICK - UNIT 2 3/4 4-13 Amendment No.Jf/,172

f! ~ -2.- f, h REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) C 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 L 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 ji 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 1 i f I i -l 1 ) 1 BRUNSWICK - UNIT 2 3/4 4-14 Amendment No. 172 )

's -FIGURE 3.4'.6.1 I PRESSURE TEMPERATURE LIMITS

  • i ~'

REACTOR VESSEL-NORMAL OPERATION WITH CORE NOT CRITICAL I? 1200 ! ii 6 i i . ! i. i .1 i ii3 i .. i 6 iii. J i i, ,,, j t i f ! I i i i ' ' i I., i J i i e . i

g, ;

1100 i i i i i t is ei i .1 1 1 t i 1 I i !I i i i i [ 1000 I i y. i i i. i i . iii ii 7 i i i i i i i i i i t I .t 'I 6 i. i 6 4 i f i ..i! i F 4 4 . i 4 . 6 6 e i i i i .i ..T f 900 l. i ir I r i i i J i 800 [ g .i I 4i i 4 g# r i 700 (n si i E f 8 l/ 600 [ h (550) r .g 500 4 e i i f I [ e 's i i 400 4f [ l r 300 f [ / 200 4 / i 4 r' f / 100 [ s s i i 0 J 100 l 200 300 400 500 600 (70) (170) TEMPERATURE (*F) FASES. 1. FUEL IN REACTOR 2. Y.16 EPP{ N/CM 3. 1 X 10 > 1 MEV 93O (1/4 T) 4. RT"N!INSTRUMENTLOCATIONCORRECTIONINCLUD = 5. 15 6. REO. GUIDE 1.99 RLV. 2 E11IL 1. OPERATE TO RIGHT AND/OR BELOW LIMITING LINES 2.

  • INDICATES BOTH HEATUP AND C00LDOWN RATE 3,

PRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES BRUNSWICK - UNIT 2 3/4 4 15 Amendment No.172

p: ..l-k -l.'i FIGURE-'3 e.6.b 2 PRESSURE TEMPERATURE LIMITS -REACTOR VESSEL ,1 NORMAL OPERATION WITH CORE CRITICAL 7 1200' 7 1 ti l '1100 i ! ) ~ u ,j 1000 [ 1 i r 900-l I o r O j 800 [ g i c V 2 700 M' g S r ~ 600 / f l' f j S00 l Q. r fr J 400 4 [ $q h^' v 300 7 7 w r .r /_ J ,F / [f 200 j r / } s' > / / /." 100 g ,, y rr i / / / A' > / .7 / 2 J' / / / F / s > f/. I 2 A' ) / f l 0 100 v0 300 400 500 -G00 l 2 0) TEMPERATURE (* F) i BASES. 1. FUEL IN REACTOR 2. 5,16EFPh 3. 7,1 X 10 N/CM > 1 MEV 4. RT'NIINSTRUMENTLOCATIONCORRECTIONINCLUD a 93 (1/4 T) 5. 15 6. REG, GUIDE 1,90 REV, 2 E!2II.L. 1. OPERATE TO RIGHT AND/OR BELOW LIMITING LINES .2,

  • INDICATES BOTH HEATUP AND COOLDOWN RAIE 3.

PRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES 4 OPERATION IN CRO 5'!!ATCICD AREA PERMITTED ONLY WHEN WATER LEVEL IS WITHIN NORMAL RANCE FOR POWER OPERATION. BRUNSWICK UNIT 2 3/4 4 16 Amendment No.172

-e - ,e; z.: FIGURE 3.4.6.1-3a PRESSURE TEMPERATURE LIMITS REACTOR VESSEL HYDROSTATIC AND LEAK TESTS 1200 1100 d\\r i n an i i i i v f i mr 6 f r 1000 '00* _O .q.--- 900 + =,j a g p ? ~ =. 800 UH' Cl i e + f / o. t 77L ) .F 700 e r e u s e e 4

0. j@ *

^ 600 W av .e 3 ^* Y h l$l 500 $C\\@c l iE / f <mo = 300 I2983 e- _me i + 200 100 70 80 90 100 110 120 130 140 150 160 170 180 TEMPERATURE ('F1 AMIL 1. FUEL IN REACTOR 2. REACTOR NOT CRITICAL 3. REG GUIDE I 99 REV. 2 4. < 8 ETPY 5. 3.5 X 10 N/CM > I MEV 6. RT a 77 (1/4 T) ISSIINSTRUMENTLOCATIONCORRECTIONINCLUDED 7. mut I. OPERATE TO RIGHT AND/OR BELOW LIMITING LINES 2,

  • INDICATES BOTH HEATUP AND COOLDOWN RATE 3.

PRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES 4 OPERATING LIMIT INDICATES TEMPERATURI REQUIRED IF TEST PRESSURE WAS EXCEEDED. BRUNSWICK - UNIT 2 3/4 4-17 Amendment No. 147, 172

4-e FICURE 3.4.6.1 3b PRESSURE TEMPERATURE LIMITS '+ :

  • REACTOR VESSEL a

HYDROSTATIC AND LEAK TESTS. 1200_ i i F 1100' s# ) ~~~ i i g 'e (iusc j tioos) f 1000 9 o i+ ) .___5 r ::: 900 g S O t 000 ' Boo' y 6/2UJ _ i r 700 f ? 6 e r 600 o Y f',, r B -s /p _.- i 500 ', i Y go '5 e w-r _mei I (440) Q' cgBC 400 j Ig98I oo n AN 3E 200 I$ Ifr i 100 70 80 90 100 110 120 130 140 150 160 170 180' 190 TEMPERATURE (*F) AMluL 1. FUEL IN REACTOR 2. I.4X10{# < 10 ETP 3. N/CM > 1 MEV 4 RT = 82 (1/4 T) 5. 15931INSTRUMENTLOCATIONCORRECTIONINCLUDED 6. REG. QUIDE 1.99 REV. 2 7. PIACTOR NOT CRITICAL NOTES: 1. OPERATE TO RIGHT AND/OR BELOW LIMITING LINES 2.

  • INDICATES BOTH HEATUP AND COOLDOWN RATE 3.

PRESSURE AND TEMPERATURE INTERSECTICNS NOTED BY PARENTHESES 4. OPERATING LIMIT INDICATES TEMPERATURE REQUIRED IF TEST PRESSURE WAS EXCEEDED. BRUNSWICK UNIT 2 3/4 4-18 Amendment No. Q,172

c. 4 FIGURE 3,4',6.1-3c = PRESSURE. TEMPERATURE LIMITS REACTOR VESSEL e-HYDROSTATIC AND LEAK TESTS 1200 i i i i i t i 1100 ,i 7-s+1 MTag) vf $s 1000 Y' h h = _-__ 900 a g_ i 800 1 a me J' l/MB4 f l0 5] 2F l ? 700 w t .e F a, w 600 .J' s s Y' 7 4 ..s i 500 se' g -o-s j 3Q* t .2w _mp-1 i

  1. cJfe (410)

.pg,c 400 2 (300)- P _ANEE 200 3 i 100 140 150 160 170 180 190 70 80 90 100 110 120 130 TEMPERATURE ('F1 16EL 1. FUEL IN REACTOR 2. < 12 ETP{ N/CM 3. 5.3 X 10 > 1 MEV 4 RT = 86* (1/4 T) 5. 1531INSTRUMENTLOCATIONCORRECTIONINCLUDED 6. REO. GUIDE 1.09 SEV. 2 7. REACTOR NOT CRITICAL F.21LL 1. OPERATE TO RIGHT AND/OR BELOW LIMITING LINES 2.

  • INDICATES BOTH HEATUP AND COOLDOWN RATE 3.

PRESSURE AND TEMPERATURE INTERSECTIONS NOTED BY PARENTHESES 4 OPERATING LIMIT INDICATES TEMPERATURE REQUIRED IF TEST PRESSURE WAS EXCEEDED. BRUNSWICK - UNIT 2 3/4 4-19 Amendment No.172

s. 11 a m, 3 ~ e x .1 TABLE.4.4.6.1.3-1 REACTOR VESSEL HATERI AL SURVEILLANCE PROCRAM CAPSULE WITHDRAWAL SCHEDULE t-CAPSULE VESSEL WITHDRAWAL TI'ME(a) i NUMBER LOCATION (EFPY) -3 300* 10- '2 120 (b)' .l 11 30 (b) h (a)-The. specimen shall be withdrawn during ref ueling 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. t 1. L i L 1 I-l. l BRUNSWICK - UNIT 2 3/4 4-20 Amendment No.172 -l 1, 6

'~ F q 4]' REACTOR COOLANT SYSTEM REACTOR STEAM' DOME LIMITING CONDIVION FOR OPERATION- ~i j ~3.4.6.2..The pressure in the reactor steam dome shall be less than 1020 psig. APPLIC ABI LITY1. CONDITION 1* and 2*. ~ ACTION i 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 i 12 hours. I SURVEILLANCE REQUIREMENTS -j 'l 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1020 psig at least once per 12 hours. } i i i r i s ! g H

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

j-i 1 l 1 BRUNSWICK - UNIT 2 3/4 4-21 Amendment No. 172

m REACTOR COOLANT SYSTEM 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. APPLICABILITY: CONDIT!CNS 1, 2, and 3. ACTIOWs Wiih 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 maintaitied OPERABLE in each affected main steam line that is open and eithert a. The inoperable valve (s) is restored to OPERABLE status within 8 hours, or 2. The affected main steam line(s) is isolated within 8 hours 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 hours. SURVEILLANCE REQUIREMENTS 4.4.7 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.$. BRUNSWICK - UNIT 2 3/4 4-22 Amendment No.172 l 3

. I, 7-f. [ SEACTOR COOLANT SYSTEM 3/4.t.8 STRUCTURAL INTECRITY LIMITIWC CONDITION FOR OPERATION 3.4.8 The structural integrity of ASME Code Class 1, 2, and 3 component s shall be maintained in accordance with Specification 4.4.8. APPLICABILITY: CONDITIONS 1, 2, 3, 4, and 5. ACTIOWs With the structural integrity of any ASME Code Class I components not r a. conforming to the above requirements, restore the structural integrity of the af fected 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 require' by WDT considerations. b. With the structural integrity of any ASME Code Class 2 components (s) 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. 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. CONDITION 5. SURVEILLANCE 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. I { l l BRUNSWICK - UNIT 2 3/4 4-23 Amendment No. 172 1

i 4-I g. REACTOR COOLANT SYSTEM BASES i l The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action. L 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 f rom a main steam line failure outside the containment during steady state operation will not exceed i small fractions of the dose guidelines in 10CFR 100. Permitting operation to continue for limited time periods with higher specific activity levels acconsnodates 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 bdine concentration during short-term iodine spikes ensures that the thyriod 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 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 i l activity levels in the reactor coolant will be detected in suf ficient time to take corrective action. l 3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the i L' effects of cyclic loads due to system temperature and pressure changes. These l cyclic loads are introduced by normal load transients, reactor trips, and i 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 I 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 to be accounted for are tensile at the inner wall and compressive at the outer wall. BRUNSWICK - UNIT 2 B 3/4 4-3 Amendment No. UJ,172

a 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 adjutted reference temperature, based upon the NDT. fluence can be predicted using the proper revision of Regulatory Guide 1.99. The pressure / temperature limiA curves Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3a through 3.4.6.1-3t inclode predicted adjustments for this shift in RTNDT at the end of indicated EFDY, te well as adjustments to account for the location of the pressure-sentir.g instrvnients. The actual shift in RigDT of the vencel material will be checked periodically during operatton by remaving and evaluating, in accordance with ASTM E185-82, reactor vessel ma:etial irr+4istion surveillance specimens installed near the inside vall of the reactor-vessel in the core area. Since the neutron spectra at the irradiation sarepics and vessel inside radius vary little, the measured transitien shift for a sample can be adjusted with confidence to the adjacent section of tire reactor vessel. The pressure / temperature limit lires shown $n Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3a through 3.4.4.1-3c have cuen provided to assure compliance with the minimum temperature req 9iremst.ts of ste 1983 revision to Appendix C of 10CFR50. The conservative 6cthod of the Standard Review Plan has been used for heatup and cooldown. The number of reactor vessel irradiattors surveillance specimens and the f requencies for removing and testing these specimens are provided in Table 4.4.6.1.3-1 to assure complinnce with the requirements of ASTM E185-82. i BRUNSWICK - UNIT 2 B 3/4 4-4 Amendment No. 172 .}}

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