Text

Amends 38 to Licenses NPF-37 & NPF-66 & Amends 25 to Licenses NPF-72 & NPF-77,respectively,approving Changes to Tech Specs Which Would Reduce RHR Min Flowrate During Refueling Operations
	ML20059F114
Person / Time
Site:	Byron, Braidwood
Issue date:	08/31/1990
From:	Barrett R; Office of Nuclear Reactor Regulation
To:	; Commonwealth Edison Co
Shared Package
ML20059F117	List: ML20059F114; ML20059F124;
References
	NPF-37-A-038, NPF-66-A-038, NPF-72-A-025, NPF-77-A-025 NUDOCS 9009110095
	Download: ML20059F114 (46)
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. m AtENDMENT TO' FACILITY'OPERAT1HG LICENSE c

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iAmendmen t No. 38 '

t.icense ho, NPF-37',

1.: -The Muclear Regulatory Constission (the Conaiission) has fou'nd that:-

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The application for -anendment by Cons'onwealth Edison Conipany!(the',

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licensee)'datedJanuary 31,;.1990, as.supplenented August-30 '1990,;

J complies with:the standards:andLre h

Act c.f;1954, as enended (the' Act) quirements of the Aton:ic Energy and the Consdssion's' rvles anC,

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regulatior,s set forth in 10 CFR' Chapter I; m

B.

Thel facility will operate in conforn.ity with the applicatien.

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the provisfor.s of: the Act, and the rules,and' regulatiorc of the Comission;'

R C.

Tiereisreasonableassurance-(i)thattheactivitiesauthorized

't by' this anendment car, be conducted without endangering the health and safety of the public, < A di) that such: activities'wi'.1-ber t

conducted in con:pliance witn t7e Conn.ission's regulations; D.

The issuance of Ms actndnent will' rict.,e inimical' to the. concion Jefense and security or to the health and sbfety.of the public; end

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E.

The issuance of this emendnent is in accordance with 10 CFR Part 51 of the Conmission's regulations and all applicable requiren.ents hbve been satisfied.

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Accordingly, the license is amended by changes to the Technical Specifi-caticr:s as indicated in the attachnent to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby an. ended to read as follows:

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.(2) ;Technica'1 Specifications' The-Technical Specifications contained'iri Appendix A'as revised;

.through Amendment No. -38 and the Environmental Protection Plan'

'y-contained in Appendix B', both of which are attached hereto, are

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,hereby incorporated into this license., The licensee shallioperate;

' the facility in accordance with the Technical-_ Specifications and the, W

Environmental Protection Plan.-

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This license. amendment is effective as of the date of.' itsi-issuance.,

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T FOR'THE NUCLEAR.REGULA'ORY COMMISSION-4 3._

i, Rich rd'J. Barrett, Director-T a

Project. Directorate III-2 l

Division.of ReactortProjects'- III, IV, Y and-Special: Projects' c

Office of Nuclear-: Reactor Regulation-

Attachment:

Changes to the Technical Specifications

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Dete of Issuance: August'31,~1990 t

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COPEONWEALTH EDISON COM0ANY DOCKET-NO. STN 50-455

'm BYRON STAT 10L UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE:

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~ Ainend$ent Ho' 78 License No. NPI'-60 y;

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The Nuclearz Regulatory Conn:ission (the Comission) has-found that:1

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The application for _ amendment by: Comonwealth Edison Conpany (the V'

licensee)datedJanuary 31 1990, as supplemented August-30,_1990,.

complies with the stalidards:end-requirements of the Atomic Energy

,c Act of 1954. as'an. ended :(ths Act). and the Comission's'rulesLand:

4 rz 1ations set for'h in 10 CFR Chapter 1;

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'B.

k;; fai:ility 111 olerate.in conforn+1ty with the: application, the provisior,s of-tie Act,~and the rules and regulations-of the-Comission; C.

Thereisreasonableassurance-(i)thattheactivitiesauthorized by this an.tndment can' be conducted without endangering'the-health;

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end safety of. the public, and:.(ii) that such activities will_bei conducted in compliance with the Comission's regulations;!

D.

Tkissuance of this enendnent will not be inimical to the comon defenst and security or to the health and safety of the public; and E.

The issuance of this en:endn.ent is in accordance with 10' CFR Part 51-of the Conrission's regulations and all applicable requirenents.

hbve been satisfied.

AccordinglyIndicated in the attachment to this license amendirent, and'the license is ame 2.

caticos as a

paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby antnded to read as follows:

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TheTechnicalSpecificationscontainedEinAppendixA/h0 REG-1113).

as revised through Anendment No.-.38 and revised by 1sttachment_2.-

to NPF-66, and the Environnental Protection Plan contained in 3.

Appendix B, both'of which were attached to License No. NPF-37, dated February 14,1985,; are.hereby incorporated into'this;1icense.

Attachnent 2 containsla revision to Appendix' A which is hereby-inccrporated.into this license. The licensee shell. operate.the' facility in'accordance with.the. Technical Specifications:and the-Environnental: Protection Plan.

3.

This license atendment is effective as of the date of its issuance.:

FOR THE NUCLEAR REGULATORY COMMISSION-

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Riche W Barret t, ' Director Project Directorate III Division of Reactor Projects' 4111, IV, Y and Special: Projects Office of Nuclear Reactor Regulation Attachn ent:'-

- Cbbr,ses to the Technical Specifications Date of. Issuance:. August 31, 1990 r..

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E ATTAC HME NT, TpJ l CEpSE, AM,E NpDME N,T, Npb}g,; ANpjg l

/ FACILITY OFERATING LICENSE NOSr NPF-37 AND NPF-66 1

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DOC KET NOS S.T.N. 50..454..A.N.D..S.T.N. 50..4 55,

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Revise _ the Appendix' A Technical-Specifications by renoving the pages identified

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below and inserting the attached pages.. The revised pages are identified by-

, a the captioned anendrent. nun;ber and contain u.arginalslines. indicating'the area q

- of change.

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Rer.ove Pag,e,s:

-Insert Pages e

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LIX JIX' X

.XL a

.XVI XVI J

XVIII XV111

~l 3/4 4-41 3/4:4-41'

.3/4 5-5 3/4 5-5; j

3/4 5-9 3/4'5-9:

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,o" 3/4 5-11" 3/4 9-9 13/4 9-9

-3/4 9-10 3/4 9-10.

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'B 3/4 4-16:

B 3/4~4-16 P

B 3/4 5-1 B 3/4 5-1 B 3/4 5-2 B 3/4 5-2 B 3/4 5-3 m

a:T B 3/4 5-4 B 3/4 9-2 B 3/4 9-2 M1 B 3/4 9-3 B 3/4 9-3 l

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. LIMITING CONDITIONS'FOR OPERATION AND SURVEILLANCE REQUIREMENTS-SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS

~ /4.5.1 ACClNULATORS.........c...................................-

3/4 5-1 3

3/4.5.2 ECCS '.4.dSTEMS

'T,yg g 350'F...........................

3/4 5-3 FIGURE 4.5-1 RESIDUAL' HEAT REM 0"AL NMP MINIMUM ACCEPTABLE

~ PERFORMANCE CURVE...................................

3/4 5-6a 3/4.5.3 ECCS SUBSYSTEMS - T,yg <

350'F............................

'94 5-7 2 '4. 5.

ECCS SUBSYSTEMS - T,yg LESS THAN OR EQUAL TO 200*F Pressurizer Level Greater Than 5 Percent (Level 409.5')..

3/4 5-9 Pressurizer Level Less Than or-Equal to 5 Percent (Level'409.5').....................................

3/4 5-10

~

3/4.b.5 REFUELING WATER STORAGE TANK.............................

3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4. 6.'1 PRIMARY CONTAINMENT Co ntai nme nt I nte gri ty..................................... -

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

3/4 6-2 Containment Air Locks....................................

3/4 6-4 Interna 1' Pressure..................

3/4 6-6 Air Temperature..........................................

3/4 6 Containment Vessel Structural: Integrity...-...............

~ /4 6-8 3

Containment Purge Ventilation System.....................

3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................

3/4 6-13 Spray Additive System....................................

3/.4 6-14 Containment Cooling System...............................

3/4 6-15 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................

3/4 6-16 TABLE 3.6-1 CONTAINMENT ISOLATION VALVES..........................

3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors........................................

3/4 6-25 Electric Hydrogen'Recombiners............................

3/4 6-26 BYRON - UNITS 1 & 2 IX AMENDMENT NO. 38 l

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-LIMITING' CONDITIONS:FOR OPERATION AND SURVEILLANCE REQUIREMENTS' c,

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'SECTION PAGE l

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b 3/4.7. PLANT SYSTEMS d

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~ 3/4.' 7.1 TURBINE CYCLE rJ n

' Safety Va1ves........,...................................

3/4174'

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~ TABLE 3.7 MAXIMUM' ALLOWABLE POWER RANGE NEUTRON i

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FLUX HIGH SETPOINT WITH INOPERABLE-

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STEAM LINE SAFETY VALVES DURING FOUR LOOP 1

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1-0PERATION.................................,.

3/4L7-2 4

y TAO'.E 3. 7-2 STEAM LINE SAFETY VALVES PER L00P.....................i 3/4.7-3' N*

' Auxi l i ary Feedwate r System.. '............................. :

3/4 7-4'

(

L Condensate Storage Tank..................................

3/4 7 ' Specific Activity........................................

3/4 7-7 1

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TABLE 4.7-1 SECONDARY: COOLANT SYSTEM SPECIFIC ACTIVITY

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SAMPLE AND ANALYSIS FR0 GRAM.........................

3/4 7-8 F

Main Steam Line Isolation Va1ves.........................

3/4'7-9 3/4.7.2 STEAM _ GENERATOR PRESSURE / TEMPERATURE LIMITATION..........

3/4'7-10 1

3/4.7.3

. COMPONENT COOLING WATER SYSTEM...........................

3/4-7-11 1

3/4.7.4 ESSENTIAL SERVICE WATER-SYSTEM...........................

3/4:7-12 3/4.7.5 ULTIMATE HEAT SINK.......................................

, 3/4 7.-13 '.

3 3/4.7.6

' CONTROL ROOM VENTILATION SYSTEM..........................

3/4 7-16

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l 3/4.7.7 <LNON-ACCESSIBLE AREA EXHAUST FILTER PLENUM

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o VENTILATION SYSTEM.......................................

3/4 7-19 L

-'3/4.7.8 SNUBBERS.................................................

3/4 7-22 L

' FIGURE 4.7-1 LSAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST...........

3/4;7-27 y

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-SEALED SOURCE CONTAMINATION..............................

3/4 7-28:

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BYRON!-' UNITS 1 & 2 X

AMENDMENT NO. 38 -

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-3/4.4.5. STEAM GENERATORS.........................................

B 3/4 4-3

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!3/4,4.6' REACTOR COOLANT SYSTEM LEAKAGE............................

B 3/4 4-4:

3/4.4;7 CHEMISTRY.................................................

B 3/4 4-5~-

3/4.4.8 ' SPECIFIC ACTIVITY.'.........................................

'B 3/4 4-5' l

L1/4.4.9. PRESSURE / TEMPERATURE LIMITS...............................

B 3/4 4 ;

TABLE.B_3/4.4-la REACTOR: VESSEL TOUGHNESS (UNIT 1)................

B 3/4'4-11 s

TABLE B 3/4.4-1b REACTOR VESSEL TOUGHNESS (UNTT 2)...............'.

B 3/4.4-12 FIGURE-B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV)'AS A FUNCTION OF

. FULL POWER SERVICE LIFE........................

B 3/4-4-13 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RT NOT l

FOR REACTOR VESSEL STEELS EXPOSED TO 1

IRRADIATION AT 550'F...........................-

B 3/4 4-14 3/4.4.10 STRUCTURAL INTEGRITY.....................................

B 3/4 4-It' '

L 3/4.4.11~ REACTOR-VESSEL HEAD VENTS................................

B 3/4 4-17 3/4.5 EMERGENCY CORE COOLING SYSTEMS

-1 ACCUMULATORS.........................'.....................

B 3/4 5-1 3/4.5.1 3/4.5.2, 3/4.5.3'AND 3/4.5.4 ECCS SUBSYSTEMS.......................

B 3/4.5-1 3/4.5.5 REFUELING WATER STORAGE TANK..............................

B-3/4 5-4 t

3/4.6 CONiAINMENT SYSTEMS 3/4.6.1 PRIMARY-CONTAINMENT.......................................

B 3/4 6-1

'I 3/4.6.2 DJPRESSURIZATION AND COOLING SYSTEMS......................

B 3/4 6-3 3/4.6.3 -CONTAINMENT' ISOLATION VALVES..............................

B 3/4-6-4 I

3/4.6.4 COMBUSTIBLE GAS CONTR0L...................................

B 3/4'6-4 i

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V i-BYRON - UNITS 1 & 2 XVI AMENDMENT H0. 38 l.

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BASES-

-SECTION:

PAGE-3/4.9.6. REFUELING MACHINE.........................................

B.3/4'9-2' 3/4. 9. 7. ' CRANE TRAVEL - SPENT FUEL STORAGE FACILITY................

B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.............

B 3/4 9-2 L3/4.9.9 CONTAIHMENT PURGE IS0LATION' SYSTEM........................

.B 3/4 9-3 J3/4.9.10.and 2/4.9.11 WATER LEVEL - REACTOR VESSEL and

. STORAGE P00L.............................................

B 3/4~9-3 3/4.9.12' FUEL HANDLING BUILDING EXHAUST FILTEP. PLENUM SYSTEM....... :B 3/4 S-3

.3/4.10 SPECIAL TEST EXCEPT70NS.

3/4.10.1 SHUTDOWN MARGIN..'.........................................

B 3/4 10-1

.3/4.10.2 GROUP. HEIGHT,-INSERTION, AND POWER DISTRIBUTION LIMITS....

B 3/4 10-1 3/4.10.3 PHYSICS TESTS.............................................- B 3/4 10-1:

3/4.10.4 REACTOR COOLANT L00PS.....................................

B 3/4 10-1 3/4.10.S POSITION INDICAi!ON SYSTEM - SHUT 00WN.....................

B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS,

-3/4.11.1 L I Q U I D E F F LU E KT S.........................................

B 3/4 11-1:

3/4.11.2 GASEOUS EFFLUENTS........................................

B 3/4'11-3' 3/4.11.3 -SOLID RADI0 ACTIVE WASTES.................................

B 3/4 11-7 3/4.11-4 TOTAL D0SE...............................................

B 3/4.11-7 3/4.12-RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM.......................................

B 3/4 12 3/4.12.2 LAND USE CENSUS..........................................

B 3/4 12-1

-3/4.12.3 INTERLABORATORY COMPARIS0N PR0 GRAM.......................

B 3/4 12-2 BYRON -' UNITS 1 & 2 XVIII AMENDMENT NO. 38 y

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REACTOR COOLANT SYSTEM SURVEILLANCE RECL'IREMENTS 4.4.9.3.1 Each PPRV shall be demonstrated OPERABLE by:

a.

Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; b.

Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and c.

Verifying tae PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:

a.

For RHR suction relief valve RH8708B verify at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that valves RH8702A and RH87028 ere open.

b.

For RHR suction relief valve RH8708A verify at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that valves RH8701A and RH8701B are open.

c.

Testing pursuant to Specification 4.0.C.

4.4.9.3.3 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vant(s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at

' east once per 31 days.

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BYRON - UNITS 1 & 2 3/4 4-41 AMENDMENT NO.

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EMERGENCY CORE COOLING SYSTEMS /

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$URVEILLANCE REQUIREMENTS (Continued)3

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l1)) For all accessible' areas of the containment prior to

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. establishing CONTAINMENT INTEGRITY,.and

-2), 10f-the areas affected'within containment at the completion of-'

each containment entry when CONTAINMENT INTEGRITY is established.(

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At11 east _once per.18 months by-H l'

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J 1)- : Verifying automatic'~ interlock action of the RHR System *from d

~the Reactor Coolant System by ensuring that'any simulated ort 4

(actual Reactor Coolant System pressure signal greater than or' equal to 360 psig prevents the valves from being'. opened.

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A visual inspection of the containmant sump andl verifying that-2) 7 the subsystem suction inlets are not restricted by debris:and-l-

that the sump components (trash racks, screens, etc.).show no' 4

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evidence of structural distress or abnormal corrosion.

4, e.

At least once per.18 months,.during shutdown, by:

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Verifying that each automatic valve lin'the flow path actuates; o;

to its correct' position on a-Safety Injection test signal and-on a RWST Level-Low-Low test signal, and

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Verifying that each of the following. pumps start automatically' fi upon receipt cf a Safety Injection actuation test signal I

a)

Centrifugal charging pump,-

O b)

Safety Injection pump, and c)

'RHR pump.-

a f.

By verifying that each of the following pumps develops the indicated:

differential pressure on recirculation flow when tested pursuant to

-Specification 4.0.5:

1)

TCentrifugal charging pump t 2396 psid, 2)

Safety Injection pump 1'1412 psid, and 3)

RHR pump In accordance with Figure 4.5-1

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BYRON - UNITS.1 & 2 3/4 5-5 AMENDMENT NO. 38

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EMERGENCY CORE COOLING SYSTEMS-

' 3/4.5.4 ECCS SUBSYSTEMS - T.._ LESS THAN OR EQUAL TO 200'F' s

PRESSURIZER LEVEL GREATER THAN 5 PERCENT (LEVEL-409.5'):

LIMITING CONDITION FOR OPERATION-t 3.5.4.1D All, Safety _ Injection pumps shall be inoperable.

APPLICABILITY:

MODE 5 with pressurizer level greater than 5 percent, and; H0DE 6 with pressurizer level greater than 5 percent and the reactor vessel' head resting on the reactor vessel g

flange.:

ACTION:-

' With a= Safety ~ Injection pump OPERABLE, restore all Safety Injer. tion pumps to inoperable status within <4 hours.

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SURVEILLANCE REQUIREMENTS:

4.5.4.1 All Safety Injection pumps shall be demonstrated inoperable

by

' ' verifying that the motor circuit breakers are secured'in the open position at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

An inoperable pump may be energizesi for testing or for filling accumulators provided the discharge of the pump is isolated from the RCS by a closed

. isolation valve with power _ removed from the valve operator, or by a manual isolation valve secured in the closed position.

BYRON - UNITS 1 & 2 3/4 5-9 AMENDMENT NO. 38

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q,e 3l EMERGENCY CORE COOLING SYSTEMS z+

3/4.5.4' ECCS SUBSYSTEMS - T LESS THAN OR EQUAL V 200*F

' PRESSURIZER LEVEL LESS THAN-OR EQUAL TO 5 PERCENT (LEVEL 409.5')

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LIMITING CONDITION FOR OPERATION 3.5.4.'2 At least one Safety ' Injection pump and flowpath shall be available, i or

' the hot side of~the RCS must be adequately vented and have valve alignments; to allow gravity feed'frem tSe RWST.

APPLICABILITY:

Either MODE

.or MODE 6 with pressurizer level less than or equal to 5 percent.;

ACTION:

If neither Safety Injection pump is available and the hot side of the RCS is not adequately vented then immediately initiate corrective action to restore either condition.or establish pressurizer level greater than 5 percent.

SURVEILLANCE REQUIREMENTS 4.5.4.2.1. At least one Safety Injection pump shall be demonstrated available, when required, by verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that 1) the motor circuit breakers are.rs:ked in'and open with the control swite.h in the pull out position, and 2) an OPERABLE flowpath exists from the RWST to the RCS, or-4.5.4.2.2' The RCS shall be demonstrated to be adequately vented, when required, by verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that:

a.

One of the following hot side vent paths is available:

1)

The reactor vessel head is removed, or 2)

The pressurizer upper manway is removed, it has been at least 140 hours0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months since shutdown and the RCS is 140*F or less, or 3)

Three pressurizer safety valves are removed, it has been at least 410 hours0.00475 days 0.114 hours 6.779101e-4 weeks 1.56005e-4 months since shutdown and the RCS is 140*F or less, or 4)

Two pressurizer safety valves are removed, it has been at least 850 hours0.00984 days 0.236 hours 0.00141 weeks 3.23425e-4 months since shutdown and the RCS is 140*F or less.

b.

An OPERABLE flowpath that will permit gravity feed from the RWST is available.

BYRON - UNIls 1 & 2 3/4 5-10 AMENDMENT NO. 38 4

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EMERGENCY CbRE CbOLING SYSTEMS :

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LIMITING CONDITION IOR OPERATION' I

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3.5.5 The refueling water storage. tank (RWST) and 'the heat traced portion-4

,4 of, the RWST vent path shall. be OPERABLE with:.

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A minimum contained borated water level.of 89%,

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'A minimum boron concen' ration of 2000 ppa,)

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, A minirJm Water temper 8ture of 35'F, and j

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A maximum water temperature of 100'F.

1 APPLICABILITY:. MODES 1, 2, 3, and 4.

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ACTION:

' With the RWST inoperable, restore the tank to OPERABLE status within ~1 hour or.

i

- be in at 1 east HOT-STANDBY lwithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the!

l following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

.v q

s SURVEILLANCE REQUIREMENTS i

4.5.5 The RWST shall'be demonstsated OPERABLE:

I a.

At least once per 7 days by:

1)

-Verifying the contained borated water levs1 in:the tank, and-2)

Verifying the boron concentration of the water.

e 1

b.'

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is either less than'35'F or greater than 100*F, and 1

c.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST vent path temperature to be greater than-or equal to 35'F when the outside air temperature is less than 35*F.

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4 -

BYRON - UNITS 1 & 2 3/4 5-11 AMENDMENT NO. 38

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REFUELING OPERATIONS' I

'.3/4.9.8 ' RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION i

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HIGH' WATER LEVEC T

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LIMITING CONDITION FOR OPERATION i

3.9'.8.1.';At least one residual heat removal (RHR): loop 1shall_be OPERABLE-and 1

t in operation."

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APPLICABILITY:

MODE 6, when the water level above the' top of the reactor 4

vessel flange is greater than or ' equal to 23' feet.

U

'ACT10Ni With no RHR' loop OPERABLE-and in operation,. suspend all operations involving s

aniincrease in the-reactor decay heat load or a reduction in boron concentra-m

, tion of the Reactor Coolant ~ System and immediately initiate corrective action;

[

to return'the required RHR loop to OPERABLE and operating status as soon as

.possible.

Close all containment penetrations providing direct access.from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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L SURVEILLANCE REQUIREMENTS l-4.9.8.1 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , one RHR loop shall'be verified in operation.

andcirculatingcoolantataflowrateofgreaterthanorequalto1000gpmwith l

RCS temperature less than or equal to 140 F.

4 1:

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l FThe RHR loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor

-vesse,1; hot legs.

-t BYRON - UNITS 1 & 2 3/4 9-9 AMENDMENT NO. 38 i

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,C REFUELING' OPERATIONS la LOW WATER LEVEL a

- LIMITING CONDITION FOR OPERATION

-I 3.9.8.2 Two1 residual heat removal (RHR) loops shall be OPERABLE, and at least:

, one RHR loop shall be in operation, j

l APPLICABILITY:

MODE 6, when the water level above the top of-the reactor vessel flange is:less than 23 feet.

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a ACTION:

~

With less than the required RHR loops OPERABLE, immediately initiate a.-

corrective action to return the-required RHR loops to OPERABLE:

status, or establish greater than or equal to 23 feetof water above:

q the reactor vessel flange, as.soon as possible, o

b.

With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of'the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop,to operation.

Close all containment penetrations providing' i

direct access from the containment atmosphere-to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.9.8.2 At' least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months one -RHR loop shall be verified in operation andcirculatingcoolantataflowrateofgreaterthanorequalto-1000gpmwith RCSitemperature _less than or equal to 140 F.

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BYRON - UNITS 1 & 2 3/4 9-10 AMENDMENT NO. 38 1

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PRESSURE / TEMPERATURE LIMITS (Continued)'

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The 'use' of the composite curve is necessary to set conservative 'heetup :

y' limitations because'it is possible for conditions to exist such that over:the.

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' course,of the heatup ramp the controlling condition switches from.the inside 4

M to the outside and the pressure limit must at all times be based on analysis.

j' y1 of the most critical criterion.

Finally, tne com o site curves for the.heatup rate data and the cooldown H

rate data are ?'HPt d for possible errors in the pressure and' temperature 1

i sensing instrum nts by the. values indicated on the respective curves.-

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Although'the pressurizer operates in temperature ranges above those for; L

4

, t sich there;is reason for concern of nonductile failure, operating limits

.O are provided to assure compatibility of operation with the fatigue analysis o

performed in accordance with the-ASME Code requirements.

3 The OPERABILITY of two PORVs, or two RHR suction valves, or an RCS ventJ

' opening of. at.least 2 square inches ensures that-the RCS will-be protected:from fpressure transients.which could exceed-the limits of Appendix G to 10 CFR Part 50 when one or mo:'e of. the RCS cold legs are less than~ or' equal-to. 350'F..

Either PORV.has adequate relieving capability to protect the RCS from6overpres-4

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surization when the transient is-limited to either: (1) the: start of an idle RCP with the secondary water temperature of the steam' generator less than or-

- equal to 50 F above the RCS. cold-leg temperatures, or (2) the start of a.

centrifugal-charging pump and its injection into a water solid RCS.

These two scenarios are analyzed to determine the resulting overshootr p

L assuming a single.PORV actuation with a stroke time of 2.0 seconds from full L

closed to full open.

Figure 3.4-4 is based upon this analysis-and represents L

the maximum allowable PORV variable setpoint such that, for the two overpres-t surization transients.noted, the resulting pressure will not exceed the i

Appendix G ' reactor vessel NDT limits (nominal 10. effective full power years L

for. Unit 1 oniy).

' 3/4.4;10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life.of the plant.

These programs are in accorcance with Section XI of the-ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the.

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L Commiss' ion ;,arsuant to 10.CFR 50.55a(g)(6)(1).

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p BYRON - UNITS 1 & 2 B 3/4 4-16 AMENDMENT NO. 38

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l 3/4.5 EMERGENCY CORE'C00 LING SYSTEMS-

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3/4.5!1 The OPERABILITY of each Reactor Coolant System (RCSj accumulator ensuresi that. a' sufficient solume of borated water will be immediately forced into the

' core through each of the cold legs in the event the RCS' pressure falls below J

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' the' pressure of the accumulators. LThis initial surge of. water into the core

- provides the initial cooling mechanism during large RCS pipe -ruptures.

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The ' limits on accumulator.. volume, boron concentration and pressure ensure' that the assumptions used for accumulator injection in.the safety analysis are q

met.- A contained borated water level between 31% and 63% ensures' a volume of:

i greater than or ' equal to 6995 gallons but less than or equal to 7217 gallons.'

n The accumulator power operated isolation valves are considered to be

~~

!' operating' bypt.sses!' in' the context of IEEE Std. 279-1971, which requires that.

H bypasses of a= protective function be removed automatically whenever-permissive-conditions.are not met.-

In addition, as these accumulator isolation valves O

a fail to meet single failure criteria, removal of: power totthe valves is reouired.

The limit's for' operation with an accumt.1stor inoperable for'any reasor.

except'toiisolation valve closed minimizes the time exposure of the. plant;to a LOCA event occurring concurrent.with failure:of an additional accumulator which may result in unacceptable peak cladding temperatures.

If a closed isolation valve cannot be immediately opened, the full capabii dy of one accumulator:is not available and prompt action is required to p uce the-reactor

.in:a mode where this capability is not required.

The' requirement-to verify accumulator' isolation valves shut with power removed from the valve operator when the pressurizer is solid ensures the accumulators will not inject water and cause a pressure transient when the Reactor Coolant System is on solid plant pressure control.

t 1/4.5.2, 3/4.5.3 AND 3/4.5.4'ECCS SUBSYSTEMS The OPERABILITY of two indepcadent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsysten through any single failure consideration.

Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures-within acceptable limits for'all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe-downward.

In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.

With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable

. reactivity condition of the t eactor and the limited core cooling requirements.

BYRON - UNITS 1 & 2 B 3/4 5-1 AMENDMENT NO.38 i

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EMERGENCY CORE COOLINGiSYSTEMS-

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ECCS'SUBSYSTEh5 (Continued)i

-The limitation for a maximum of one centrifugal charging pump to be

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OPERABLE and-the~ Surveillance Req'uirement to verify'all charging pumps except p

the: required OPERABLE Charging _ pump to be inoperable in MODE 4 with one or.

more of-the RCS cold;1egs less than or equal to 330*F, MODE _5, and MODE 6 with-y4

~ the reactor vessel head _on. provides assurance'that a mass addition pressure R

.. transient.can be relieved by the operation of a single PORV~or RHR suction c relief valve.

Similarly, the requirement to verify all Safety Injection pumps -

aretinoperable in MODE-4 with the temperature of one or more of the RCS Cold:

.]

. Legs-less than or equal to 330*F, in MODE 5 with pressurizer level greate than

_5' percent /(Level 409.5') and in MODE 6 with pressurizer level greater than--

o 5 percent and the reactor vessel. head resting on the reactor vessel flange, provides assurance thatLa mass addition pressure transiant can be relieved by a' single!PORV or RHR~ suction relief valve.

In MODE 5 and MODE 6 with pressurizer level less than or equal to 5 percent, at least one. Safety Injection pump or gravity feed from the RWST must be avail-d able~-to mitigate the effects of a loss of decay heat removal during partially

-drained conditions.

Surveillance. requirements assure availability, out prevent inadvertent actuation during these modes.

The desired flow path for the SI

-pump or. gravity. feed varies with RCS configuration and is, therefore, procedurally.

addressed.

The ' Surveillance Requirements define what constitutes an adequate hot side vent for various plant conditions.

It was determined that removing the reactor-ll vessel head was an adequate vent under all conditions.

Other venting alterna-tives have restrict' ions based on time from shutdown and RCS temperature.

The values.n the' surveillance were taken from the graph on the following page.

-l The Surveillance Requirements provided to ensure OPERABILITY-of. each L

. component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.

Surveillance Requirements for throttle vaTVe position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA.

Maintenance of proper flow resistance and pressure drop In the piping system to each injection point is necessary to:

(1)' prevent total pump flow from exceeding

. runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable

' level of total ECCS flow to all injection points equal to or above that ' assumed

~

in the ECCS-LOCA. analyses.

The Surveillance Requirements for leakage testing of ECCS check valves ensures that a failure of one valve will not cause an intersystem LOCA.

In Mode 3, with pressurizer pressure below 1000 psig, the accumulators will be available with their isolation valves either closed but energized, or open, whenever a SI8809 valve is closed to perform check valve

? leakage testing.

BYRON - UNITS 1 & 2 8 3/4 5-2 AMENDMENT NO. 38

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3/4.5.5-REFUELING WATER STORAGE TANK

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.The OPERABILITY'of the~refuelitq water storage tank _(RWST) as part of the i

'ECCS ensures that a sufficient supply of borated water is-available for injection 1

4 by the ECCS in the eventfof a LOCA.. The limits:on RWST minimum volume'and boron-concentration' ensure that: (1) sufficient water is available:within containment-to

,.y permit; recirculation cooling flow to the core, and (2) the reactor will remain:

'suberitical in the cold condition following mixing of-the RWST and the RCS water.

volumes with all control rods inserted except for the most reactive. control' m

-?

assembly. These; assumptions'are consistent with the LOCA analyses.;

i:' '

The contained water volume limit includes an allowance for water not:

i usable because of tank discharge line' location or other physical characteristics.

A minimum contained borated water level of 89% ensures a volume of greater than or equal;to 395,000 gallons.

The' limits on; contained, water volume and boron concentration.of,the RWST

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also ensure a pH value of between 8.5 and 11.0-for the solution. recirculated-within containment after a LOCA. ~This pH band minimizes the evolut. ion of iodine and minimizes the effect of chloride and caustic stress corrosion ont j

mechanical systems and components.

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I tBYRON,- UNITS 1 & 2 B 3/4 5-4' AMENDMENT NO. 38

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3/4.9.6 REFUELING MACHINE

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The 0PERABILITY requirements for the refueling machine and auxiliary hoist ensure that:- (1): refueling machines will be used for movement of drive rods ~ and

_ N.,

< fuel assemblies,.(2) each refueling machine has sufficient load capacity to liit a drive rod or fuel assembly, and (3) the core internals and reactor vessel

.are protected from excessive lifting force'in the event they are inadvertently

' engaged during-lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEt. STORAGE FACILITY.

h The. restriction on movement of loads in excess of the nominal weight of a feel and control-. rod assembly and associated handling tool over other fuel assemblies.in the. storage. pool areas ensures that in the event this load is dropped: (1) the. activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel.in the storage racks will 'not result in a critica1' array. This assumption-is consistent with.

the activity release' assumed in the safety analyses.

=3/4.9.8 RESIDUAL HEAT. REMOVAL AND COOLANT CIRCULATION L.

.R '

The requirement that 'at least one residual heat removal-(RHR) loop be in

- operation ' ensures =that: (1) sufficient cooling capacity is available to remove, decay heat.and maintain the water in.the reactor vessel below 140'F as required during the' REFUELING MODE, and-(2) sufficient coolant circulation is maintained through the core.to minimize the affeet of a boron dilution incident and prevent boron stratification.

=

k' The surveillance requirement verifics that the RHR loop is operating and

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. circulating reactori coolant to ensure the capability of the RHR system to main-tain compliance with plant design limits.

The required RHR loop reactorc coolant flowrate is determined by the flowrate necessary to:

(1) provide sufficient U

decay heat removal capability, (2) maintain the reactor coolant temperature rise

- through the core within design limits, for compliance with flowrates assumed in the boron dilution analysis,. (3) prevent thermal and boron stratification in the core,'(4) preclude cavitation of the reactor coo knt downstream of the RHR o

-"s Eflow controlivalve,.and (5) ensure that' inadvertent boron dilution events can be identified and' terminated by operator action prior to the reactor returning critical.

In addition, during operation of the RHR loop with the water level in the' ivicinity of the reactor, vessel nozzles, the RHR loop flowrate determination =must

~ L also consider the RHR pump suction requirements.

At this water level, thy RHR b

pump can experience vortexing or cavitation conditions which would cause the loss'of RHR pump. operation, if the flowrate demand is too high.

Operation with

$p reactor' coolant water at this level is often called mid-loop operation.

Care i

must be. taken in" determining the RHR loop flowrate, when operating with water

-d level in this region, to prevent loss of the RHR pump and subsequent loss of the RHR loop'for. decay heat removal.

,1 BYRON - UNITS 1 &'2 B 3/4 9-2 AMENDMENT NO. 38

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3/4.9.8-RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION (Continued)-

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.The requirement to have two. RHR loops-OPERABLE: when there is '.less than.

23 feet.of. water above the reactor vessel flange ensures that a single. failure l

of the operating _RHR' loop will not result in a complete loss of RHR capability.

.With the reactor vessel-head removed and at least 23 feet of' water above_the

1 reactor vessel flange, a11arge heat sink is available for core cooling._.'Thus,.

]

in the event of a failure of the operating RHR. loop, adequate time is provided M

to' initiate emergency procedures-to cool the core.

0?

.3/4.9.9 CONTAINKENT PURGE ISOLATION SYSTEM

._ The' OPERABILITY of this' system ensures that the containment-purge j

~

. penetrations will be automatically isolated'upon detection of high radiation 4

h levels ~within the containment.

The OPERABILITY.of this system is required to restrict the releasefof radioactive material from the containment atmosphere

to the environment.

~

3/4.9.10 and'3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL

(

The restrictions.on minimum water level ensure that sufficient water.

?

depth is'available:to remove 99% of the! assumed 10% iodine gap activity released J

from the. rupture of an-irradiated fuel assembly.~ The minimum water _ depth _is

.i consistent with'the assumptions'of_the safety analysis.

j

.d l3'/4.9112. FUEL HANDLING BUILDING EXHAUST FILTER PLENUM

.The limitations:on the Fuel. Handling Building Exhaust Filter Plenum y ensure' that'all radioactive material released from an irradiated fuel assembly ~

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will be filtered through the HEPA filters and charcoal adsorber prior to. dis-L charge to the atmosphere.

The OPERABILITY'of this system and-the resulting L

' iodine removal capacity are consistent with the assumptions of.the safety, a

L*

analyses. ! ANSI N510-1980 will be used as a procedural guide for' surveillance l

testing.'

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NUCLEAR REGULATORY COMMISSION -

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AMENDMENT TO FACILITY OPERATING LICENSE-1 Amendment No.:25-r License No. NPF-72 1

1. -

The Nuclear-Regulatory Con.n'ission (the Consiission) has found that:

A.-

The application for; amendment by Conronwealth Edison Conpany (the licensee)datedJanuary 31, 1990,'as supplemented August 30, 1990,-

4 f

- Act of.1954, as an4 ended (the Act) and the-Consiission's: rules and.

^

ccn: plies with. the' standards and requirements of the Aton:ic Energy j

Es regulations set forth'in 10 CFR Chapter I-e c o u

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'B.

The facility will operate in conformity with_the application,.

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the' provisions of the. Act,- and-the rules and regulations of.the i

-Conmission;,

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Thereisreasorableassurance-(1)thattheactivitiesauthorized by-this'antodnient can be conducted without endangering the health

j er.d safety of the public,'and (ii) that such activities will be.

J ccoducted inicompliar ce.with the Conmission's regulations; i

D.-

The issuance df this arendn.ent.will oct be inimical to the-connon

'i defense and security or to the health'and-safety of the.public ~

11

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.Theissuance:ofthisamendnientisiirEaccordancewith10CFRPart51 l

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'of the Conn;ission's regulations and all applicable requirements have been satisfied.

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, Accor.dingly, the license is amendeo by changes: to' the Technical' Specifi-a taticns as indicated in the attachment to.thistlicense amendment, and

. paragraph 2.C'.(2) of Facility Operating License No. NPF-72.is hereby:

1

, a enended to. read as follows:

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'(2)- Technical Specifications' The Technical Specifications contained.in. Appendix A as revised' f

through Amendment No. 25'~and the Environmental Protection Plan'

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-contained in Appendix B, both of-which are attached hereto, are 1

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-hereby incorporated,into this licensea The licensee shall operate 1

the' facility in.accordance with'the Technical Specifications.and the Environmental Protection Plan J

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H 3. -- This license amendment,is effective as of the date of its issuance and is'to be implemented by, December 15, 1990, j

.FOR THE' NUCLEAR REG LATORY COMMISSION-4 l,

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Richar J.uBarrett, Director L

-Project Directorate III-2 1

-Division of Reactor Projects - III, 1

m IV, Y and.Special Projects ~

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Office of Nuclear Reactor Regulation

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Attachment:

Changes to the. Technical L,. '

Specifications 4

Date of Issuance:. August 31, 1990 r

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NUCLEAR REGULATORY COMMISSION i

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COMMONWEAtTHEDLSpMpMLAM 6

DOCKET:NO. STN 50-457 1

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BRAIDWOOD,$1ATip!liUhlT NO.: 2 -

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. AMENDMENT TO FACILITY OPERATING LICENSE !

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. Amendment No. 25 R

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License No. NPF-77 1

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The Nuclear Regulatory Cons.ission1(the Consiission) has found that:

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- The application for amendn.ent bh Contonweilth Edison COipany (the_

j licensee) dated January 31, 1990,Las supplemented August 30,.1990, gm,

complies with the. standards and: requirements of the Atomic Energy '

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- Act of 1954, as an. ended (the Act) and the Concission's rules and d

i regulat.ionssetlforthin10CFR.ChapterL1, as,.,

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The facility will operate in' conformity with the:applicatiers the-provisiors.cfethe.Act, and the rules and, regulations ~of the' M4 Coramission; y

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sThere'is reasonable assurance (i)Jthat;the actiyities authorized 1

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by this aundt.ent can,be conducted without endangering the health 4

.and safety of the public, and1(ii)Jthat? such activities will!be 4

cor. ducted in compliance with the Consission's reg ~ulations;-

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.The issuance of.,this amendnent willfidt be inimical?to the conson n

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defense land security or to the health and~ safety of the public; 0

and QL,

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7 EL The-issuance of this enendn.ent is in actordance with 10 CFR Part 51 a

f f of the Cocmission's regulations and d applicable requiren.ents t

have been satisfied.

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Accordingly, the license is amended by> changes tol the' Technical Specifi-V cations as indicated in the attachment'to this license-amendment, and.

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paragraph 2.C.(2) of' Facility Operating" License No. NPF-77 is'hereby

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' amended to read as follows:

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-(2)L Technical Specifications EsSM

-The Tebhnical Specifications contained in Appendix A'as revised'

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through Amendment No. 25-and the Environmental Protection-Plan =

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N("I'm' contained in Appendix B, both of which were attached to Licenset 9

No,'NPF-72, dated July'2, 1987, are hereby incorporate.into this f

' license. The -licensee shall operate the facility in= atcordance -

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,with the Technical Specifications and.the Environmental Protection.

N-Plan.

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~This' license' amendment is effective as of the date ofiits issuance and.

o3%h is to be implemented t,y December 15, 1990.

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h-FOR THE NUCLEAR REGULATORYJCOMMISSION m

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Richard J. Barrett, Director-g

-:.e Project Directorate 111-2

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"m Division of Reactor Projects.- III, IV, Y and Special: Projects J

Office of Nuclear Reactor Regulation-3

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Attachment:

Changes to the-Technical J

Sslp Specifications im 1

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Date..of Issuance: ~ August 31, 1990 M-lj,(

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LATTACHMENT-TO LICENSE AMENDMENT NOS. 25 AND 25' i

FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 1

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. DOCKET NOS.- STN 50-456 AND STN: 50-457 Replace:the following pages of the Appendix "A"-Technical Specifications -

l with the attached pages. The revised pages are identified by amendment-W number and contain vertical linas indicating the area of change.

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Remove Pages insert Pages IX-IX-Xc X

XVII XVI.

I XVIII XVIII 3/4 4-41 3/4 4-41 3/4 5-5 3/4 5-5 3/4 5-9 3/4 5-9 3/4 5-10 3/4 5-11

'3/4.9-9 3/4 9 3/4:9-10 3/4 9-10 j

B 3/4 4-16

-B 3/4 4-16 B 3/4 5 B 3/4 5-1 B 3/4 5-2 B 3/4 5-2 B'3/4 5 -

B 3/4 5-4 B 3/4-9 B'3/4 9-2

.B'3/4 9-3 B 3/4 9 Et n

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y SECTION' PAGE<

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3/4.5 EMERGENCY CORE COOLING SYSTEMS 1

3/4.5.1 ACCUMULATORS..................-..........................

3/4 5-1 i

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3/4.5.2-ECCS SUBSYSTEMS - T,yg > 350_*F..c........................

3/4 5-3 1

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' FIGURE 4.5 1; RESIDUAL-HEA" REMOVAL PUMP MINIMUM 7

ACCEPTABLE PERFORMANCE CURVE.......................

3/4 5-6a; r

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'r 3/4.5.3 ECCS SUBSYSTEMS - Tavg <350'F.............................

3/4 5-7:

L" 3/4.5.4 ECCS SUBSYSTEMS - T LESS THAN OR EQUAL TO 200'F j

avg l;

Pressurizer Level Greater Than 5 Percent (Level 409.5')..

"3/4 5-9 I

Pressurizer Level Less Than or Equal.to 5; Percent (Level 409.5')............................

3/4 5-10

?

'3/4.5.5 REFUELING WATER STORAGE TANK.......

3/4 5-11 1

1?

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.3/4.6 CONTAINMENT SYSTEMS l-3/4.6.1 PRIMARY CONTAINMENT ~

1

_ Containment Integrity....................................

3/4-6-1 a

l

Containment Leakage......................................

3/4 6-2i E Containment Air Locks....................................

3/4 6-4 o

Internal Pressure........................................

3/4 6-6:

Air Temperature...........................................

3/4 6-7 Containment Vessel Structural Integrity..................

3/4 6-8' Containment Purge Ventilation System......................

3/4 6-11 3/4.6.2.

DEPRESSURIZATION AND COOLING SYSTEMS i

i<

Containment Spray System.................................

'3/4 6-13 Spray Additive System.....................................

3/4 6-14 i

Containment Cooling System...............................

3/4 6 l

<3/4.6.3 CONTAINMENT ISOLATION VALVES.............................

3/4 6-16 t

m TABLE 3.6-1 CONTAINMENT ISOLATION VALVES..........................

3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors........................................

3/4 6-25 Electric Hydrogen Recombiners............................

3/4 6-26 BRAIDWOOD - UNITS 1 & 2 IX AMENDMENT NO.25 p,'

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LIMITINGCONDITIONSFOROPERAkIONAND'SUR'EILLANCEREQUIREMENTS V

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SECTION PAGE 1

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~ /4.7' PLANT SYSTEMS-3

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_3 4.7.1. -TURBINE CYCLE:

1 i

Safety Valves....................

3/_4' 7-1l d

1 TABLE-3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON p

FLUX HIGH SETPOINT WITH INOPERABLE o

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t STEAM LINE SAFETY VALVES DURING FOUR LOOP 1

0PERATION...........................................

3/4 7-2_

q TABLE 3,7-2 -STEAM LINE SAFETY VALVES'PER L00P.....................

3/4 7-35

-t

' Auxiliary Feedwater System................................

3/4'7-4 6

. Condensate. Storage Tank..................................

3/4 7-6 Specific Activity........................................

3/4 7-7.

JTABLE 4.7-l SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY d

Vh'

_ SAMPLE AND~ ANALYSIS PR0 GRAM.........................-

3/4 7-8:

l Main Steam-Line Isolation Va1ves......;..................

3/4 7-9' M

73/4.7f2!

STEAMGENERdTORPRESSURE/TEMPERATURELIMITATION..........

3/4-7-10 3/4.7q3; -COMP'ONENT COOLING WATER SYSTEM............................

3/4 7-11 3

.4

(y "3/4~.7.4 LESSENTIAL SERVICE WATER SYSTEM...........................

3/4 7-12 H

3 /4 ' '7. 5 '

U LT MAT E H EAT S I N K.......................................

3/4 7-13 l

J3/4!7.6 CONTROL ROOM VENTILATION SYSTEM..........................

3/4 7-14' 4.~.,

3/4.7.7: ENON-ACCESSIBLE-AREA EXHAUST FILTER PLENUM J

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, -VENTILATION SYSTEM.......................................

3/4 7 j 9-3/4.7.8' SNUBBERS....'.............................................

3/4 7-20~

sk' FIG 0REi4.7-1 SAMFt E PLAN 2) FOR SNUBBER FUNCTIONAL TEST...........

3/4 7-25 3

!!b > g 3/4.7.9 SEALED: SOURCE CONTAMINATION..............................

3/4 7-26 1

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BRAIDWOOD' UNITS'l & 2 X

AMEN 0 MENT NO. 25 s%

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~SECTION:

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3/4.4.5' STEAM GENERATORS..........................................

B 3/4 4-3 F

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............-.....'..........

B 3/4 4-4

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-3/4.4;7. CHEMISTRY.....................

B 3/4 4-5 o

3/4.4.8 SPECIFIC ACTIVITY.........................................

B 3/4'4-5 I

oT 3/414.9 PRESSURE / TEMPERATURE LIMITS...............................

B-3/4 4-7,

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TABLE B 3/4. 4-la REACTOR VESSEL TOUGHNESS (UNIT 1)................

B 3/4 4-11:

TABLE B '3/4.4-lb REACTOR VESSEL TOUGHNESS (UNIT 2).....-...........

B 3/4 4-12 FIGURE B 3/4.4-1: FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE........................

B 3/4 4-13 FIGURE B 3/4.4-2s EFFECT OF FLUENCE AND COPPER ON SHIFT OF_RTNDT FOR REACTOR VESSEL STEELS EXPOSED TO 1

IRRADIATION AT 550*F...........................

B 3/4 4-14 3/4.4.10 STRUCTURAL INTEGRITY.....................................

B 3/4 4-16 3/4.4.11-REACTOR VESSEL HEAD VENTS................................

B 3/4 4-17

. 3 /4.' 5 EMERGENCY CORE COOLING SYSTEMS E

'3/4.5;1-ACCUMULATORS..............................................

B 3/4 5-1 3/4.5.2, 3/4.5.3 and 3/A.5.4 ECCS SUBSYSTEMS.......................

B 3/4 5-1 3/4. 5. 5 ' REFUE LING WATER LTORAGE TANK..............................

B 3/4 5-4 3/4.6 CONTAINMENT SYSTEMS g

'3/4.6.1 PRIMARY. CONTAINMENT..............................

B 3/4 6-l' 3/4.6.2.0EPRESSURIZATION AND COOLING SYSTEMS......................

D 3/4 6-3 Y

-3/4l6.3 CONT INMENT~ ISOLATION VALVES..............................

B 3/4 6-4 3/4;6.4 COMBUSTIBLE GAS CONTR0L...................................

B 3/4 6-4

.BRAIDWOOD - UNITS 1 & 2 XVI AMENDMENT NO.25

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- %. e L 3/4' 9. 6: REFUELING MACHINE..........................................

B 3/4.9-2 n ', '< ;3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY.................

5 J/4 9-2 o

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$I4;9.8 RESIDUAL HEAT. REMOVAL AND COOLANT CIRCULATION.............

B 3/4 9-2 s

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3/4'.9.9 CONTAIMMENT PURGE ISOLATION SYSTEM........................

B 3/4 9-3

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3/4.9.10:and-3/4.9.11 WATER LEVEL - REACTOR ^ VESSEL and

.yNI STO RAG E. P00 L.'............................................

B 3 /4 9-3 ;

' 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUM SYSTEM.......

B 3/4 9-3 t

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3/4.10 SPECIAL TEST-EXCEPTIONS

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a B 3/4'10-1 3/4.10.'1 SHUTDOWN MARGIN.................s.........................

13/4.10.2 GROUP. HEIGHT, INSERTION, AND-POWER DISTRIBUTION LIMITS...

B 3/4 10-1; 3/4.10.3 PHYSICS TESTS.............................................

B 3/4 10-1 ~

3/4.10.4 REACTOR C00LANT; LOOPS......................................

B.3/4 10 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.....................

B 3/4 10-1 f3/4.11' RADI0 ACTIVE EFFLUENTS

3/4.11.11 LIQUID EFFLUENTS.........................................

B 3/4 11-1;

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. 3/4.11.2 GASEOUS EFFLUENTS........................................

B 3/4 11-3 U

13/4.11.3 50 LID RADI0 ACTIVE WASTES.................................

B:3/4 11-7 3:

- E

,3/4.11.4' TOTAL D0SE...............................................

BL3/4.11k7.

J s.

dt 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING q

3/412.1 MONITO RING P R0G RAM.......................................

B - 3/4 12-1

=t.[

3/4.12.2 LAND USE CENSUS..........................................

B 3/4 12-1

< i3/4.12.3-INTERLABORATORY. COMPARISON PR0 GRAM.......................

B.3/4 12-2

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-N (BRAIDWOOD - UNITS l' & 2 XVIII AMENDMENT NO. 25

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EMERGENCY CORE' COOLING' SYSTEMS-

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l PRESSURIZER LEVEL GREATER THAN 5 PERCENT (LEVEL 409.5')-

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-. LIMITING CONDITION FOR OPERATION i!
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.3.5.4.1' All Safety Injection' pumps shall be inoperable.
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,1 APPLICABILITY: ' MODE 5 with pressurizer level greater than 5 percent, and.
5 3'
MODE 6 with pressurizer level greater than 5 percent and 3
the reactor vessel head resting on the reactor vessel;
>p
(, < fij flange..
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.m, ACTION:
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With a Safety Injection pump OPERABLE, restore all~ Safety Injection pumps to 7l x
-inoperable status.within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
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SURVEILLANCE REQUIREMENTS::
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'4.5.4.1 All Safety Injection pumps shall be demonstrated. inoperable
by.
d verifying that the motor circuit breakers are secured in'the open positi.on.
j at-least once-'per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
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q "An inoperable pump may be energized for testing or for filling accumulators
.provided the= discharge of the pump is isolated from the RCS by a closed isolation valve with power removed from the valve' operator, or by a manual-
-isolation valve secured in the. closed position.
4BRAIOWOOD -' UNITS 1 & 2 3/4 5-9 AMENDMENT NO.' 25 :
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3/4.5;4 ECCS SUBSYSTEMS - T_.._ LESS THAN OR EQUAL TO 200'F
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3 PRESSURIZER LEVEL LESS THAN OR EQUAL TO 5= PERCENT-(LEVEL 409.5'1 l
LIMITING CONDITION FOR OPERATION 3.5.4.2 At least one Safety Injection pump and flowpath shall be available,
)
or j
the-hot side of the RCS must be adequately vented'and have valve alignments to allow gravity feed from'the-RWST.
1 APPLICABILITY:
Either MODE 5 or MODE.6 with pressurizer level less than or 1
equal to 5-percent.
ACTION:'
=
If neither Safety Injection pump is available and the hot side of the RCS is
'1 not adequately vented then immediately initiate corrective action to restore
either condition or establish pressurizer level greater than 5 percent.
'l SURVEILLANCE REQUIREMENTS:
4.5.4.2.1 At. least one Safety Injection pump shall be demonstrated available, when required,'by verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that 1) the motor circuit breakers are racked in and open with the control switch in the pull out= position,>and 2) an OPERABLE flowpath exists from the RWST to the RCS, or
-4.i.4.2.2 The RCS shall be demonstrated to be adequately vented, when' required, 5
.by' verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that:-
.a.
One of the following hot side vent paths is available:-
1)
The reactor vessel head is removed, or 2)
The pressurizer upper manway is removed, it has been at least I
t 140 hours0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months since shutdown and the RCS is 140*F or-less, or 3)
Three pressurizer safety valves are removed, it has been at least 410 hours0.00475 days 0.114 hours 6.779101e-4 weeks 1.56005e-4 months since. shutdown and the RCS is 140'F or less, or 4)
Two pressurizer safety valves are removed, it has been at least' 850 hours0.00984 days 0.236 hours 0.00141 weeks 3.23425e-4 months since shutdown and the'RCS is 140'F or less, b.
An OPERABLE flowpath that will permit gravity feed from the RWST is available.
i BRAIDWOOD - UNITS 1 &'2 3/4 5-10 AMENDMENT NO.25
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' 0' iR EMERGENCY CORE COOLING SYSTEMS pv 3/4.5.5' REFUELING WATER STORAGE TANK fu 1 m
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>! LIMITING CONDITION FOR OPERATION
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I 3.5.55-The refueling water storage tank (RWST) and the heat traced portion of. the RWST vent path shall be _0PERABLE with:
"n a.
A minimum contained borated water level of 89%,
1...
a-b.
'A miaimum boron concentration of 2000 ppa, c.
A minimum water temperature of 35'F, and f
d.
A maximum water temperature of 100'F.
1 APPLICABILITY:- MODES 1, 2, 3, and 4.-
ACTION:
With the RWST inoperable, restore the tank to OPERABLE status within I hour or 2
be in~at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30-hours.
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SURVEILLANCE REQUIREMENTS - 4.5.5 The RWST shall be demonstrated OPERABLE:=
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a.
At least once per 7 days by:
b,
'1)
Verifying the contained borated water. level in the tank, and L
J 2)- ' Verifying the boron concentration ~of the water.
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At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when L
the outside air temperature is' either.less than 35'F or greater' s+
'than 100*F, and c.
AtD1 east once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST vent path temperature to be greater than or equal to 35'F when the outside c
air temperature is less'than 35'F.
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BRAIDWOOD - UNITS 1 & 2 3/4 5-11 AMENDMENT NO. 25 L
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LIMITING CONDITION FOR OPERATION a
3.9.8.1 At.least one residual heat removal.(RHR) loop shall be OPERABLE and i
['
in operation.*
[
APPLICABILITY:
MODE 6, when the water level above the ~ top of the reactor _-
vessel-flange is greater than or equal to 23 feet.
ACTION:.
1"
'With no RHR loop OPERABLE and in operation,-suspend all operations involving an increase in the_ reactor decay heat,1 cad or a reduction.in boron concentra-tion of the Reactor Coolant System and-immediately initiate corrective action y
to return the required RHR loop to.0PERABLE and operating status'as soon as-Li
~ possible.
Close all containment penetrations providing direct access from the
'l containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
)
r SURVEILLANCE REQUIREMENTS l
r 4.9.8.1 At least once per.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , one RHR loop shall be_ verified in operation-and circulating coolant at a flowrate of greater than or equal to 1000 gpm with--
i RCS temperature.less than or equal to 140 F.
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during the performance of CORE ALTERATIONS in the vicinity of the reactor
,i
' vessel: hot legs.
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.BRAIDWOOD - UNITS 1 & 2 3/4 9-9 AMENDMENT NO. 25
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Q'li REFUELING OPERATIONS 1
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LIMITING CONDITION FOR OPERATION y
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'3.9.8.2 Two residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.-
APPLICABILITY:. V16, when-the water level above the top of the reactor 1
vessel flange is ini than 23 feet.
n ACTION:
a.
With less than the required RHR loops.0PERABLE, immediately initiate L
corrective action to return the required RHR loops to OPERABLE L
status, or establish greater than or equal to 23. feet of water above the reactor vessel flange, as soon as possible.
b.
With no RHR loop in operation, suspend all operations involving a J
J
' reduction in boron concentration of the' Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
Close all containment penetrations providing direct access from the containment atmosphere to the outside pK atmosphere within 4' hours.
1, ll c
SURVEILLANCE REQUIREMENTS 4.9.8.2 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months one RHR loop.shall be verified in operation:
l 14 andcirculatingcoolantataflowrateofgreater:thanorequalto'.1000gpmwith jj RCS temperature less than or equal to 140 F.
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'BRAIDWOOD - UNITS 1 & 2 3/4 9-10 AMENDMENT NO. 25 9
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'REACTORC00i.ANTSYSTEM l
't i
-BASES-
]
PRESSURE / TEMPERATURE LIMITS (Continued)
.. The use of the composite curve is necessary to set conservative heatup 2 limitations t,ecause it is possible for conditions to. exist such that over the icourse of the,heatup ramp the controlling condition switches from the inside' 1
to the outside'and the pressure. limit must at all times be based on analysis b
of the most-critical-criterion.
Fina.11y,'the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature t
4 sensing instruments;by.the values-indicated on the respective curves.
ll Although the pressurizer operates in temperature ranges above those for which there is reason' for concern of nonductile failure, operating limits d
are provided to. ass'ure compatibility.of operation with the fatigue analysis performed in.accordance.with the ASME Code requirements.
t 4
-The OPERABILITY of two PORVst or two RHR suction valves, or an RCS vent.
7 1,
opening offat"least 2 square inches ensures that'the RCS will be protected-from a
pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one..or more-of the RCS cold legs are less than or equal to 350'F.
~;
Either PORV has'_ adequate' relieving capability to protect the RCS from overpres-surization when the transient is limited to either: (1) the start of.an idle RCP with the secondary' water temperature of the steam generator less:than or equal to 50'F'above the RCS cold. leg temperatures', or (2) the start of a centrifugal ch:rging pump andL its injection into. a water solid RCS.
~These two scenarios are analyzed to determine the resulting overshoots assuming a-single.PORV actuation with a stroke time of 2.0 seconds from full.
closed,to full open.. Figure 3.4-4 is based upon this analysis and represents.
the maximum allowable P0dV variable setpoint such,that, for the two overpres-surization transients noted,'the resulting pressure will not exceed the nominal 10 effective full power years (EFPY) Appendix G reactor vessel NDT 3
limits; 3/4.4.10 STRUCTURAL-INTEGRITY 7
L: '
The-inservice inspection and testing programs for ASME Code Class 1, 2, and.3 components ensure'that the structural integrity.and operational readiness d
1.
- ofLthese components will be maintained-at an acceptable level throughout the 1
4' life of the plant. ;These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by
-10 CFR 50.55a(g) except where specific written relief has been granted by the Commission' pursuant'to 10 CFR 50.55a(g)(6)(1).
4
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-BRAIDWOOD - UNITS 1 & 2 B 3/4 4-16 AMENDMENT NO. 25 p
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3/4.5 EMERGENCY CORE COOLING SYSTEMS. '
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BASESI I
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~ /4.5.1 ACCUMULATORS-3 The OPERABILITY of each Reactor Coolant System-(RCS) accumulator ensures E.
that a sufficient volume of borated water will be immediately forced into the
- l core through each of the cold;1egs in the event the RCS pressure falls below the pressure of the accumulators.
This initial surge of water into the core l
provides the initial ~ cooling mechanism during large RCS pipe ruptures.
The limits on accumulator volume, boron concentration'and pressdre ensure.
"i that the assumptions used for; accumulator injection in the safety analysis are me t.'
A contained borated water level between 31% and 63% ensures a volume of.
greater than or equal to 6995 gallons but less than or equal to 7217 gallons--
The accumulator power operated isolation valves are considered to be-
~" operating bypasses" in the context'of IEEE Std.. 279-1971, which requires that bypasses of a protective function be' removed automatically whenever permissive conditions are not_ met.'--
In addition, as these accumulator ' isolation valves u
. fail to meet single failure criteria, removal of power to.the valves 11s required.,
.The limits for operation with an accumulator inoperable for any reason except'an isolation valve closed minimizes the time exposure of the plant to a-LOCA-event occurring concurrent'with failure of an additional accumulator which'may result ~in unacceptable peak cladding' temperatures.
If a closed
. isolation valve cannot be.immediately opened,- the ful1 ~ capability of one;
. accumulator is not available and prompt action is required to place the reactor
~
in a: mode where this capability is-not required.
The requirement to' verify accumulator isolation valves shut with. power removed from the valve operator when the pressurizer is solid ensures the
' accumulators'will not inject' water and cause a pressure ' transient when the Reactor. Coolant System is;on solid. plant' pressure control.
q 3/4.5.2, 3/4.5.3 AND 3/4.5.4 ECCS SUBSYSTEMS The OPERABILITY.of two independent ECCS subsystems ensures that sufficient h,
emergency core cooling capability will be~~available in the event of a LOCA L
. assuming the loss of one subsystem through any single failure consideration.
t
- Either' subsystem' operating in conjunction with the accumulators is capable of a
supplying sufficient core cooling to limit the peak cladding temperaturest u
Q within acceptable limits: for all postulated break sizes ranging from the jg
- double ended break of the largest RCS cold leg pipe downward.
In addition, each ECCS. subsystem provides long-term core cooling capability in the si h
recirculation mode during the accident' recovery period.
1 5
l.
With the RCS temperature below 350'F, one OPERABLE ECCS subsystem is 1
L acceptable without single failure consideration on the basis of the stable 1
reactivity condition of the reactor and the limited core cooling requirements.
BRAIDWOOD --UNITS 1 & 2 B 3/4 5-1 AMENDMENT NO.25 p
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BASES-1 S
- ECCS SUBS'YSTEMS (Continued)-
The limitation for a maximum of one centrifugal charging pump to be '
OPERABLE and the Surveillance Requirement to verify all-charging pumps except-the required OPERABLE Charging pump to be inoperable in M0JE 4'with one er.
y more of the.RCS-cold legs less than or equal to 330'F, MODr 5,-and MODE 6 with the reactor vessel head;on, provides assurance that a mass-addition pressure transient can;be relieved by the operation of a single PORV or RHR suction relief valve.
Similarly, the requirement to verify all Safety Injection pumps a
'are inoperable in MODE 4 with-the temperature of one or more of the RCS Cold 1
Legs less than or ' equal to 330'F, in MODE 5 with pressurizer levnl greater tiian 1
1 5 percent (Level 409.5') and in MODEL6 with pressurizer level greater than 5 percent and the reactor vessel head resting on the reactor vessel flange, H
provides assurance that a mass addition pressure transient can be relieved by' a single PORV or RHR suction relief valve.
In MODE 5 and MODE 6 with pressurizer level less than or equal to 5 parcent,
at least-one Safety Injection pump or gravity feed from the RWST must be avail-able to. mitigate the effects of a loss of decay heat removal during partially drained conditions.
Surveillance-requirements assure availability, but prevent' inadvertent actuation during these modes.
The-desired flow path for the SI
,1 pump or gravity. feed varies with RCS configuration and is, therefore, procedurally
addressed.
.The Surveillance Requirements define what constitutes an adequate hot side vent for various plant conditions.
It was determined that removing the. reactor vessel: head was an adequate vent under alltconditions. Other venting alterna-tives haveirestrictions based on time from shutdown and RCS temperature. The
~
, values in--the surveillance were-taken from the graph on the following page.
The Surveillance Requirements provided to ensure OPERABILITY-of each
~
.~ component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.
Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance
.j that proper ECCS flows will be-maintained in the event of a LOCA.
Maintenance
~
of proper' flow resistance and pressure drop in the piping system-to each
. injection point is necessary to:
(1) prevent total. pump flow from exceeding
+
. runout conditions when the system is in its minimum resistance configuration,
'(2)' provide the proper flow split between injection points in accordance with y
the assumptions used in the ECCS-LOCA analyses, and.(3) provide an acceptable.
d
,i leve1~of-total ECCS-flow to all injection points: equal to or above that assumed if in the ECCS-LOCA analyses.. The> Surveillance Requirements for leakage testing l
.of ECCS check valves ensures that a failure of one valve will not cause an intersystem LOCA.
In Mode 3, with pressurizer pressure below 1000 psig, the accume'stors will be available with their isolation valves either closed but 1
energized, or open, whenever-a SI8809 valve is closed to perform check valve r
, leakage testing.
x BRAIDWOOD - UNITS 1 & 2 B 3/4 5-2 Amendment No. 25
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4 aaECCS SUBSYSTEMS-(Continued)
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3-s TIME AFTER SHUTOOWN MS)
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"":. s Vent Path Required to Prevent
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lBRAIDWOOD -' UNITS 1 & 2 B 3/4 5-3 AMENDMENT NO. 25 m xw i
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3/4.5.5 ' REFUELING WATER STORAGE TANK; n
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. The OPERABILITY of the refueling water st'orage tank (RWST) as part _of the.
A z
ECCS ensures that' a sufficient supply of borated water is. avail _able for _ injection.
L Lby the ECCS.in'the event of'a~LOCA.' The limits on.RWST minimum volume' and boron.
1 bo concentration ensure that: (1) sufficient water is available within' containment to:
permit recirculation cooling flow to the core, and.(2) the reactor will remain h
subcritical,in the cold condition following mixing of-the RWST and the RCS water:
L volumes with all' control rods' inserted except for the most reactive control i
assembly.
These' assumptions'are consistent with the LOCA analyses.
E l
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Thelcontained water volume limit includes an: allowance for water not:
-usable because o? tank discharge line location or other physical characteristics.-
0
- A minimum contained borated water level of 89% ensuresJa volume of greater than -
H or. equal to 395,000 gallons.-
H
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. The limits-on contained water volume' and boron concentration'of-the RWST 1
also ensure:a:pH value of-between 8.5 and;11.0 for the solution recirculated I
within containment after a'LOCA.. This pH band. minimizes the evolution of
.l iodine..and minimizes = the effect of chloride and caustic' stress corrosion ~ on P
. mechanical systems and components, u
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BRAIDWOOD - UNITS 1 & 2 B 3/4 5-4 AMEN 0 MENT NO. 25 i1 4
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REFUELING OPERATIONS-3 s
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' BASES L
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-3/4'9.'6-' REFUELING MACHINE; j
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The'0PERABILITY requirements for the refueling machine and auxiliary hoist y
ensure that: (1)' refueling. machines will be used for movement of drive rods.and k
fue11 assemblies, (2)- each refueling machine has sufficient load capacity to L
lift a' drive rod or fuel assembly, and (3) the core internals and reactor vessel-y4 are protected from excessive lifting force in the event they are inadvertently-
)
?'
engaged during lifting operations.
3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY j
g l
The restriction on movement of loads in excess'of the nominal weight of a 1
~
L fuel and control rod assembly and associated handling tool over other fuel l-assemblies in the storage pool areas ensures thr', in the event this load-is l
dropped:-(1) the' activity release will be limited to that contained in a
~
i single fuel assembly, and (2) any possible distortion of fuel in the storage p
racks will not result in a critical array.
This assumption is consistent with the activity release assumed in the safety analyses.
3/4.9.8' RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION
.l The requirement that' at least one residual herit removal (RHR) loop be in operation ensures that:-(1) sufficient cooling cr.pacity is available to remove
.oecay heat and maintain the water in the reactor vessel below.1.40*F-as required -
r during the REFUELING MODE, and (2) sufficient coolant circulation-is. maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification.
'1 The surveillance: requirement verifies that the RHR loop.is operating and circulating reactor coolant to ensure the capability of the RHR system to main-tain compliance with ' plant design limits.
The required RHR loop reactor coolant
.flowrate.is determined by the flowrate necessary to: ' '(1) provide sufficient n
decay heat removal capability, (2). maintain the reactor coolant temperature rise through the core within design limits, for compliance with flowrates assumed ~in the boron-dilution analysis, (3) prevent thermal and boron stratification in-the core, (4) preclude cavitction of the reactor coolant downstream of the RHR-flow control valve, and (5)' ensure that~ inadvertent boron dilution events can 3',
'be identified and terminated by operator action prior to the reactor returning jj
- critical.
I' M
~In addition, during operation of the RHR loop with the water level in the-ivicinity of the. reactor vessel nozzles, the RHR loop flowrate determination must 1
r-M*'
Ealso consider the RHR pump suction requirements.
At this water level, the RHR
'y
. pump <can experience vortexing or' cavitation conditions which would cause the l'
loss.of RHR pump operation, if the flowrate demand is too high.
Operation with e
reactor coolant water at this level is often called mid-loop operation. ~ Care
'E OL must be'.taken in determining the RHR loop flowrate, when operating with' water level in this region, to prevent loss of the PHR pump and subsequent loss of r
the RHR. loop for decay heat removal.
f
[ ^ BRAIDWOOD - UNITS 1 & 2 B 3/4 9-2 AMENDMENT NO. 25
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REFUELING OPERATIONS-4 L
1 BASES-p 3/4.9.8-RESIDUAL HEAT' REMOVAL. AND COOLANT CIRCULATION (Continued)
~
h-i The requirement-to have two RHR loops '0PERABLE when;there is less. than.-
l L
23 feet of water above the reactor vessel flange ensures that a single failure.
(of the operating RHR loop will not result in a complete loss of RHR capability.
With the reactor vessel head removed and at least'23 feet'of water above the reactor vessel-flange, a large heat sink is-available for core cooling.
Thus, oS'
> in the event of a failure of the operating RHR loop, adequate time.is provided H
to initiate'eme'rgency procedures to cool the core.
L
'3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM f
.t l!
The.0PERABILITY of this system ensures that the containment purge 1-penetrations'will be automatically isolated upon detection of high radiation
levels within the containment.
The OPERABILITY of this system is. required to
.(
restrict the release of radioactive material from the containment atmosphere to the environment.-
1 L
3/4.9.'10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL ~ and STORAGE POOL cThe restrictions on minimum water level ensure that sufficient water Edepth is available toiremove 99% of the assumed 10% iodine gap activity released L
from/the rupture of an irradiated, fuel assembly.
The minimum water depth.is c
consistent with the assumptions of the safety analysis.
J\\
f3/4.9.12: FUEL HANDLING BUILDING EXHAUST FILTER PLENUM u
O ThelisitationsontheFuelHandlingBuildingExhaustFilterPlenum-l
. ensure that'all radioactive material released from an irradiated fuel assembly L
will:be filtered through~the HEPA filters and charcoal adsorber prior to dis -
fchargeLto'the atmosphere.
The OPERABILITY of this system and the resulting iodine removal ~ capacity are consistent with the assumptions of the safety.
analyses. ANSI N510-1980.will be.used as a procedural guide for surveillance testing.
i
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_BRAIDWOOD - l' NITS 1 & 2 B 3/4 9-3 AMENDMENT NO. 25
~
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,

ML20059F114

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