Text

Proposed Tech Specs Reducing RHR Min Flow Rate,Removing RHR Autoclosure Interlock on RHR Sys Suction Isolation Valves & Allowing Operation of Safety Injection Pump to Mitigate Effects of Loss of DHR
	ML20070U080
Person / Time
Site:	Byron, Braidwood
Issue date:	01/31/1991
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20070U075	List: ML20070U073; ML20070U080;
References
	NUDOCS 9104080108
	Download: ML20070U080 (52)
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ATTACIOiENT 1 i-PROPOSED CHANGES TO APPENDIX A

-TECHNICAL-SPECIFICATIONS'FOR FACILITY

-- OPERATING LICENSES NPF-37/66 AND NPF-72/77 Byron NPF-37/66 Table of Contents IX XVI XVIII Revised Pages 3/4 4-41 3/4 5-5 3/4 5-9 3/4 5-10 3/4 5-11 3/4 9-9 3/4 9-10 B 3/4 4-16 B 3/4 5-1 B 3/4 5-2 B 3/4 5-3 B 3/4 5-4 B 3/4 9-2 Braidwood NPF-72/77 Table of Contents IX XVI XVIII

' Revised Pages 3/4 4-41 3/4 5 3/4 5-9 3/4 5-10 3/4 5-11

.3/4 9-9 3/4 9-10 B 3/4 4-16 B 3/4 5 B 3/4 5-2 B 3/4 5-3 I"

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

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LIMITlHG CONDITIONS FOR OPERATION AND-SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................

3/4 5 1 3/4,5.2 ECCS 30BSYSTEMS - T,yg >350*F...........................

3/4 5-3 FIGURE 4.5 3 RES10 VAL HEAT REMOVAL PUMP MINIMUM ACCEPTABLE PERFORMANCE CURVE...................................

3/4 5 6a

([

3/4.5.3 ECCS SUBSYSTEMS - T 1.s>u'.

avg < 350'F...........................

3/4 5 7 3/4.5.(f REFUELING WATER STORAGE TANK..........

3/45pil 3/4.6 -CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity....................................

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

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

3/4 6 4 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 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

-C0HTAINMENT ISOLATION VALVES.............................

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

3/4 6 3/4.6.4 COMBUSTIBLE GAS CONTROL Hy d r o g e n Mo n i ta r s........................................

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

3/4 6-26 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety valves............................................

3/4 7-1 BYRON - UNIT 1 IX AMEN 0MENTNO.f m.

nw..,__,

---e e-,e-

.,-,o 2

v

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.-a-s,.,,. < --

2 sc.

. _ ~ _ _.

r Insert C 3/4.5.4 ECCS SUBSYSTEFS - Tavg LESS THAN OR EQUAL TO 200*F Fressuriter Level Greater Than 5 Percent (Level 409.5')

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

3/4 5-10 l

i

(.

BASES SECTION PAGE 3/4.4.5 STEAM GENERATOR5..........................................

B 3/4 4-3 3/4.4.6 REACTOR COO LANT SYSTEM LEAKAGE............................

B 3/4 4-4 3/4.4.7 CHEMISTRY.................................................

B 3/4 4-5 3/4.4.B SPECIFIC ACTIVITY.........................................

B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS...............................

B 3/4 4-7 TABLE B 3/4.4-la REACTOR VESSEL TOUGHNESS-(UNIT 1)................

B 3/4 4-11 ','

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

B 3/4 4-12 J 4 c.

FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1Nev) AS A FUNCTION OF e;;.

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

B 3/4 4-13

FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RTNDT FOR REACTOR VESSEL STEELS EXPOSED TO IRRADIATION AT 550'F...........................

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

B 3/4 4-16 3/4.4.11 R EA C TO R V E S S E L H EAD V E NT S................................

B 3/4 4-17 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3 /4. S.1 A C C UMU LA TO R S. )......,.............,........................

B 3/4 5-1 c, e vv s. ~I 3/4.5.2 -and 3/4.5.3 ECCS SUBSYSTEMS...............................

B 3/4 5-1 j

g 3/4. 5.fi REFUE LING WATER STOP ArE TANK..............................

B3/45g1 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT.......................................

B 3/4 6-1 3/4.6.2-DEPRESSURIZATION AND COOLING SYSTEMS......................

B 3/4.6-3 3/4.6.3 CONTAINMENT I SO LATION VALVES..............................

B 3/4-6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L......................'.............

B 3/4 6-4 e

a BYRON - UNITS 1 & 2 XVI

7 BASES SECTION PAGE 3/4.9.6 REFUELING KACHINE.........................................

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 NEAT REMOYAL AND COOLANT CIRCULATION.............

B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM........................

B3/49-g3 3/4.9.10 and 3/4.9.11 WATER LEVEL REACTOR VESSEL and STORAGE P00L............................................

B 3/4 9-3 3/4.9.12 FUEL NANDLING BUILDING EXHAUST FILTER PLENUM SYSTEM.......

B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10,.1 SHUTDOWN MARGIN...........................................

B 3/4 10-1 3/4.10.2 GROUP NEIGHT, INSERTION, AND POWER DISTRIBUTION LINITS....

B 3/4 10-1 3/4.10.3 FHYSICS TESTS.............................................

B 3/4 10-1 3/4.10. 4 REACTOR COO LANT L00PS.....................................

B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SNUTD0WN.....................

B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS.........................................

B 3/4 11 3/4.11.2 GASEOUS EFFLUENTS........................................

B 5/4 11-3 3/4.11.3 SOLIO RADI0 ACTIVE WASTES.................................

B 3/4 11-7 3/4.11.4 TOTAL D0SE..............................................., B 3/4 11-7 y

3/4.12 RADIOL 0GICAL ENVIRONMENTAL HONITORING 3/4.12.1 HONITORING PR0 GRAM.......................................

B 3/4 12-1 3/4.12.2 LAND USE CENSUS..........................................

B 3/4 12-1 1

3/4.12. 3 INTERLABORATORY COMPARISON PR0 GRAM.......................

B 3/4 12-2 I

BYPON - UNITS 1 & 2 XVIII

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORY shall be demonstrated OPERABLE by:

Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV a.

actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV 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 atleastonceper18 months (;and Verifying the 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 c.

when the PORY is being used for overpressure protection.

l 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:

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ForRHRsuctionreliefvalve8708S$

a.

3 I)

By -ver44ying at-least-+nce per 31 day; thet R"R ACS stien 4

Iso 14t4en-Ve4v; RHS702A-45 cpan with pcwer to the velve

+peester r:::v:c, :nd

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-23 verifykg at least once pergl2 hours thatdRH8702B h/open.

For RHR suction relief valve'b708Af ML b.

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( +}--By-vee 4fy4eg-et4eest-ence per-31-daye-that-AH&7018-4 Op;n

- 4th pc ar te-th: v:lve-operater r; :ved, :nd n

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verify Mt at least once per le hours that RH8701A p open.

A c.

Testing pursuant to Specification 4.0.5.

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 vent (s) is being used for overpressure protection.

S l

Except wnen the vent pathway is provided with a valve which is locked, sealed, 1

or otherwise secured in the open position, then verify these valves open at least once per 31 days.

The-spec 4ffed-l&-month-intervemay-be extended-to-32-montht-for-Gyele-1-only.

l BYROH - UNITS 1 & 2 3/4 4-41 Amendment No.14-

I' EMERGENCY CORE COOLING SYSTEMS SURVElttANCE REQUIREMENTS (Continued) 1)

For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and Of the areas affected within containment at the completion of 2) each containment tntry when CONTAINMENT INTEGRITY is established.

c.

At least once per 18 months by:

RHR Verifyingautomaticholathn-andinterlockactionofthd.n 1)

System from the Reactor Coolant System by ensuring that With+ simulated or actual Reactor Coolant System pressure 6-signal greater than or equal to 360 psig 4hc inteH oc u preventsthe valves from being opened y M-W4th-a-fr4maated-or-actual Aeac4ar Coolant Syston-pr4ssa

-signal-greater-than--or-,4Qaa.1-to 662 -ps49-the inter Locks 4444 c 4 w 64-the-v 41* e 6-to-a utoma L144 My4tose -

A visual inspection of the containment sump and verifying that 2) the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evicence of structural distress or abnormal corrosion, At least once per 18 months, during shutdown, by:

e.

Verifying that each automatic valve in the flow path actuates 1) to its correct position on a Safety inje: tion test signal and on a RWST Level-Low-Low test signal, and Verifying that each of the following pumps s' tart automatically 2) upon recttpt of a Safety Injection actuation test signal:

t a)

Centrifugal charging pump, b)

Safety injection pump, and c)

RHR pump.

I By verifying that each of the following pumps develops the indicated l-differential pressure on recirculation flow when tested pursuant to f.

Specification 4.0.5:

L.

1)

Centrifugal charging pump 1 2396 psid, 2)

Safety Injection pump 1 1412 psid, and in accordance with Figure 4.5-1 N

3)

RHR pump AMEN 0MENTNO./

3/4 5-5 BYRON - UNITS 1 & 2

I' 4

EMERGENCY CORE CCOLING SYSTEM o

ECCS SUBSYSTEMS - Tave LESS TRAN OR EQUAL TO 200 F PRESSURIZER LEVEL GREATER THAN 5 PERCENT (LEVEL 409.5')

LIMITING CONDITION FOR OPERATION 3.5.4.1 All Safety Injection pumps shall be inoperable.

APPLICABILITY:

MODE 5 with pressurizer level greater than 5 percent, and MODE 6 with pressurizer level greater than 5 percent and the reactor vessel head resting on the reactor vessel flange.

ACTION:

With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to inoperable status within I hours.

SURVEILLANCE REQUIREMENTS:

4.5.4.1 All Safety Injection pumps shall be demonstrated inaperable 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 energized for testing or for f1J11ng accumulators provided the dischargo of the pua.p is isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valvo secured in the closed position.

Byron - Units 1& 2 3/4 5-9 l

EMERGENCY SORE COOCM[G SYSTEM o

ECCS SUBSYSTEMS - Teve LESS TRAN OR EQUAL TO__200 F PRESSURIZER LEVEL LESS TRAN OR EQUAL TO $ PERCENT (LEVEL 409.5')

LI.MITING CONDITION T2R OPERATICN 3.5.4.2 At least one Safety Injection pump and flowpath shall be available, or the het side of the MCS must be adequetely vented and have valve 1.

alignments to allow gravity feed from the RWST.

APPLICABILITY:

Either MODr 5 or MODE 6 with prossurizer level Tes.s than or squal to S pe.ecent.

ACTION:

If neither safety Injection pump is available and the hot side of the RCS is not adoquately

rented then immediately initista corrective action to restore either condition or establish pressurizer level greater than 5 percent.

SURVEILLANCE REQUIREMENTS t 4.5.4.2.1 At least one Safety In demonstratedavailable,whenrequ,jectionpumpshallbe red, by verifying at least on:e per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that 1) the motor circuit breakers tre racked in and open with the control switch in the pull out position, and 2) en OPERABLE flowpath exists from the RWST to the RCS, or 4.5.4.2.2 The RCS shall be demonstrated to be edequately vented, when required, by verifying at least once per 12 rours 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 in removed, it has been at

{

least 140 houre since shutdown and the RCS is 140'r or i

less, or

3) Three pressurizer s3Ctsy 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 lose, or

4) Two pressurizer safety valves are removed, it has been at least 850 houre since shutdown and the RCS is 140'r or

less,

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

Byron - Units 1 & 2 3/4 5-10

EMERGENCY CORE COOLING SYSTEMS, 5

]/4.5.l g.,':3.ING WATYR STORAr/ g'tfjj LIMITitG DHDIT'!

FOR OPERATION i

s k.

1.5./ The ref wilri;l water storage tank (RWST) and the heat traced portion of the RWST vent path shall be OPERABLE with; A minimum contained borated water level of 89%,

a.

L b.

A mbimum Doron concentratior, of 2000 ppm, A minin.um water temperature c? 3b'F, and k

c.

d.

A maximun water temperature of 100*F.

,o APPLICABILITY:

MODES 1, 2, 3, and 4.

'5 ACTICtN:

k

[_

With the RJST iretrible, restore the tank to OPE %ABLE status within 1 but or be in at least HOT SiANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in C?LD SHUT 00W within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

' j' i

t SURVE1LLANCE_400'.REMENTS 4.5 The RWST sht11 be demonstrated OPERABLE:

k At least once per 7 days by:

a.

i 1)

Veri /ying the conteined borated water level in the tank, and

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Verifying tbt boton concentra*. ion of the water.

b.

At least once per 2f. hoars by vr+.rt fying the RWST temperature when F

the outside air 1

twoerature is either less than 35"F or greater than 100*F, and At least once per 24 h2urs by veri *fying the RWST vent path c.

twperature te be gr ute.r than or aqual to 35'F when the outside air tunperature if Ier s than 35 F.

x SvRON UNITS 1 & 2 3/45-/11

RE8UEL!NG OPERATf0NS 3/4.9.8 RESIDUALHEAT(1EMOVALANjLC00LANTCIRCU,gTION MGHWATERLEVEL L191 TING CONDITION rom OPERATION 3.9.8.1 At least one residual heat removal (RHR) lock shall be OPERABLE and in operation."

APDLICABILITY:

MODE 6, when the water level above the top of the rer.: tor vessel flange is greater than or equal to 23 feet.

ACTION:

With no RHR loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentra-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 , SURVEILLANCE REOUIREMENTS 4.9.8.1 At4eastane-RHR-Ioop shall be -verified -in -operation and circulating reactor coolant at--e--flow Pete 4f greater than or equal to 2000 ppm at least ence-per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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i Fine Rr1R loop may be removed from operation for up to I hour per 8-hour period during ti.e performance of CORE ALTERATIONS in the vicinity of the reactor vessel het legs.

EYRON - UNITS 1 & 2 3/4 9-9

i REFUELING OPERATIONS l

LOW WATER LEVEL LIMITING CONDITION FOR 06'89'.T10N 3.9.8.2 Two residual heat removal oneRHRloopshallbeinoporation.\\(RHR)loopsshallbeOPERABLE.andalleast a

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

ACTION:

i a.

With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops tu OPERABLE status, o* 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 reduction in boron concentrat'.on of thi. Aeactor Coolant System and imediately initiate corrective action to return the required RHR loco to operation.

Close all containment penetrations providing dir s 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 .

i SURVEILLANCE REOUIREMENTS 4.9.8.2 At 4 east one-RHR-4eop-shal4-be-verified-in-operation-and-circulat4ng i

reactor-coolant 4t -a --flow-f ate -+f-greater-than-oe-equal-to. 2800-gpm -at-least-i once-per-1240ur6r 5*

co' em p:> 11 3;s; P.1 t!; /' i n." h e vu d e d n ep' u. u,' s e.. '. '

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" P r i or-t o-i ni ti a l--c r it i c a li ty r-t h e -RHR -4 eep-m ay-be - remov ed - f rom -opera t i on -f or i

up -to hour - pe e-2-hov e - pe r iod -du r i ng - t he-pe r f orma nc e -o f40R E-A L T E RAT ION S-4n-the-v ic i ni ty-of-the-re ac to r--v e s s e bho t -4 egsv-i l

BYRON

UNITS 1 & 2 3/4 9-10

RE ACTOR COOLANT SY3'ilM BA,SES PRESSURE /TE@ERATURE LIMITS (Continued)

'lhe use of the composite curve is necessary to set conservative heatup limitations because it is possible for ceditions to exist such that over the course of the heatup ramp the controlling cordition switches from the inside to the outsids and the pretsure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the rerspective curves.

Although the pressurizer operates in temperature range's above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analp is perfoemed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, or two RHR suction valves, or an RCS vent i

opening of at least 2 squarc inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix 0 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 wi.th 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 overshoots assuming a single PORV actuation with a stroke time of 2.0 seconds from :ull closed to full open.

Figure 3.4-4 is based upon this analysis and represents the maximum allowable PORV variable setpoint such that, for the two overpres-surization transients noted, the resulting pressure will not exceed the Neinal 10 effective full power years (EFPY) Appendix G reactor vessel NDi limits.

RHR-RC+-tvetion-41eht4en-vehet-8MlA-and-BMBA-eee-4nterleeked+RF Or "A"-trai-fr wide renge pretsure trensmitter end velves-e?MB-end-67020-ere-inte r iceked with e "0" trein-wide rang; pressure-teensmuter.

R:::v4ng p;-:t fr::

v:1 ve c4418-and-SMBAr-Pr+v*nts- *-s4*91*-44449ee-4+6*-46*dveet+*t4y holat469-4eth4HR-wet 4en-rel4ef-vehes-wh44e-malete(ning RHR ite44t4en-cepeM44ty 107 bettt-RHR-fiew peths.

i 3/4.A.10 STRUCTURAL INTEGRITY Tne 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 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)(i).

BYRON

UNITS 1 & 2 B 3/4 4-16

a i

3/4. 5 EMERGE Q CORE COOLING SYSTFMS BASES 4

3/A.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators.

This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions use6 for accumulator injection in the safety analysis are A contained boratec water level between 31% and 63% ensures a volume of met.

greater than or equal to f995 gallons but less than or equal to 7217 gallons.

The accumulator power operated isolation valves are considered to be

" operating bypassts" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed autaatically whenever permissive 4

conditions are not met.

In addition, as these 3ccumulater isolation valves fail to meet single failure criteria, removal of rwer to the valves is required.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the timt expesure of the plant to a LOCA event occurring concurrent with failuFe of an deitional accumulator which may result in unacceptable peak cladding tems.ratures.

If a closed isolation valve cannot be immediately opened, the fu'l capability of one accumulator is not available and prompt action is reqeired to place the reactor in a mode where this capability is het required.

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

ed 3I4. i. 't j

3/4.5.2 4 4 3/4.5.3 ECCS SUBSYSTEMS The OPbtABILITY of two independent ECCS subsystems enstret that sufficient emergency core cooling capability will be available in the overt of a LOCA assuming the loss of one subsystem through any single f ailui'e ennsideration.

Either subsystem operating in conjunction with the accumulato>', is capable of supplying suf ficient core cooling to limit the peak cladding unmperatures within acceptable limits for all postulated break sizes rar/ ng from the double ended break of the largest RCS cold leg pipe downwbrJ.

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 30'F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

i BYRON' UNITS 1 & 2 B 3/4 5-1 r

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I' EWEe6ENCY C0er @00LfNG SYSTEMS BA!!$

1 ECCS SUESYSTEMS (Continued)

IN5ERT B 4

-The-4hMet4t+-br e mixima-ef cn; c;ettifugel chen'ng pump, tc b; F

TM L: in; th; brseillence R;; b ;;;nt te v;rify ell che q&n i

~&thty-lefteten pers; except-the re;vited OP "ACLC charging pw:g purp; :nd -

--abh b;10 3M94-Pedd:: ::: u re(+-44t- :

p tr.r-be-in;;e r -

-be silieved by the-optretica cf e single."CP". --::;-*6444en-pMs;ur; t+6M4ri ca-The Surveillance Requirements provided to ensure OPERABILITY of each are met and that subsystem OPERABILITY is maintained. compo for throttle valve position stops and flow balance testing provide assuranceS l

that proper ECCS flows will be maintained in the event of a LOCA.

of proper flow resistance and pressure drop in the piping systeh, to each Maintenance injection point is necessary to:

(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance conf'guration, (2) provide the proper flow split between injection points in accortance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an a:cepta level of total ECCS flow to all injection points equal to or above tnat assum in the ECCS-LOCA analyses.

of ECCS check valves ensures that a failure of one valve will not intersystem LOCA.

In Mode 3, with pressurizer pressure below 1000 psig, the accumulators wil.1 be asailable with their isolation valves either clos energized, or open, whenever a S18809 valve is closed to perform check valve leakage testing.

b S

' 214. 5. A R UUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for inje g u. a by the ECCS in the event of a LOCA.

concentration ensure that:

The limits on RWST minimum volume and boron permit recirculation cooling flow to the core, and (2) the rear; tor

-. suberitical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods in'serted except for the most reactive control assembly.

These assumptions are consistent with the LOCA analyses.

The contained water volume limit inc11 des an allowance for wate A minimum contained borated water level of 89% ensu or equal to 395,000 gallons.

also ensure a pH value of between 8.5 and 11.0 for the so within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosio mechanical systems and components.

BYRON - UNITS 1 & 2 B 3/4 5-2 AMEN 0MENTNO./>

.-,-.-.-.-uc...

- __ _ ~___ -_ -

- -.3--

ItJ 5E.RT

' ~

The limitation f or a maximum of one centrif ugal charging pump to be OPERABLE and the Surveillance Requirement to verify all Charging pumps except the required 0FERABLE Charging pump to be inoperable in MODE 4 with one or more of the RCS cold legs less than or equal to 330'r, MODE 5, and MODE 6 with the reactor vess-n. head on, provides assurance that a mass addition pressure transient can be relieved by the operation of a single P0kV or RHR suction relief valve.

Similarly, the requirement to verify all Safety injection pumps are inoperable in MODE 4 with the temperature of one or more of the RCS Cold Legs less than or equal to 3300F, in MODE 5 with pressuriger level greater than 5 percent (Level 409.5') and in MODE 6 with pressuriser level greater than 5 percent and the reactor vessel head resting on the reactor vessel flange, provides assurance that a mass addition pressure transient can be relieved by a single FORV or RHR suction relief valve.

In MODE 5 aad MODE 6 with pressurizer level less than or equal to 5 percent, at least one Safety Injection pump or gravity f eed f rom the RWST must be available to miti ate the effects of a loss of decay heat removal during F

partially drained conditions, surveillance requirements assure s.vailabill'y, but prevent inadvertent actuation cucing 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 teactvr vessel head was an adequate vent under all conditions. Other venting alternatives have restrictions based on time f rom shutdown and RCS temperature.

The values in the surveillance were taken f rom the graph on the following page.

l l

/sel:05957:27

I"F80ENCY CODE COOLING SYSTEMS BASES ICCS SUBSYSTEMS (Continued) 0

-TRCs = 100 F

___TACs = 140*F 30.5 2 SAFETIES

/

30.0

/' s '

NOT ACCEPTABLE

,/

25.0 20.0 18.3 3 SAFETIES

/'

15.0 s

/-

I f

ACCEPTABLE 10.0

/

AY" 8.5

/

5.0 0.0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 700.0 000.0 900.0 TIME AFTER SHUTOOWN pgs)

Vent Path Requried to Prevent RCS Pressurization BYRON UNITS 1 & 2 8 3/4 5-3 i

- -=.

3 Ew!RMNcv CORE COOLING $YSTEMS BASES 5

3/A.5.k REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS nsures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA.

The limits on iWST minimum volume and deren

)

concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the coresuberiticalinthecoldconditionfollowingal.;

volumes with all centrol rods inserted except for the most reactive control assembly.

These assumptions are consistent with the LOCA analyses.

The contained water volume Ifmit includes an allowance for water not 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 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 evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on Eechanical systems and components.

I e

BYRON - UNITS 1 & 2 g 3/4 5.b AMENDMENTNO.f e

RfNFL!w0 OPERA 9f 0N$

BA$t$

3/L L 5 t!NfLIE IMMINE The OPEAA8!LITY Mewiremente for the refueling machbe and ausiliary heist ensure that:

1) mfueling nachines will be sud far aevament of drive esas anc fuel asseatli(es (2) aach r9fuelirq nachine has evfficism 1HW capacity lift a drive mg er fue) assembly, and (3) the seen intamals and reacter vesse) are protacted free escessive liftin engspot during If f ting operations, g force in the event they are inairertently i

3 /4. 9. 7 CoMt TRAVEL = ptWT Ntl $T0sJct FACf LfiY The restriction en movement of leads in sucess of the nominal weight of a fuel and control red assantly and assectated hand 1(ag tot 1 over other fue?

assentiles in the sterage pool areas an6wres that in the even'; this lose is i

dropped: (1) the activity release will be limited to that containes in a single fuel assembly, and (2) any possible distortion of fuel in the storage rects will not result in a critical array. This assumption is consistent with the detivity release assmed in the safety analyses.

3 /a. 9. 8 RE$! DUAL W[AT Rt.McvAL AND C00LAWT C!t:ULAY10N The requirement that at least one nsidwal best reseval (MR) loop te in operation ensures that: (1) sufficient cooling capacity is available to esmove decay heat and uintain the water in the racter vessel belev 140'F as reewirec during ths RIFUELIM MODE, and (2) sufficient coolant circulation is saintainec throwgh the core to etnisite the effect of a beNn dilution incident anc p mvent boron stratification.

WMorA The requirement to have two NR loops CPERA8LE when there is less than 23 feet of water above the reacter vessel flange ensures that a sin 0le failure of the operating MR loop will not result in a complete less of MR capattlity.

With the reactor vessel head removed and at least 23 fut of water above the reactor ressel flange, a large heat sink is available for core cooling. Thws, in the event of a failun of the operating RNR loop, adequate time is provisec ta initiate energency procedures ta cool the aere.

3/a.9.9 COWTAflMIWT PVtCI ISCLATION SYSTEM The 07ttAI!LITY of this rystes ensures that the containment purge penetrations will be automatically isolated upon detection of high radiation l

1evels within tAe contaf tusent. The OPERA 41LITY of this system is retvind to restrict the release of raffenctive matarial free the containment assosphere to the anvironment.

1 i

BYt0N UNITS 1 & 2 8 3/4 9-2

)

~

lNSE AT A 1

IFr0*/I'lidv00d Itthmical $catificatiem $1333 3.Lt11 1he surveillance nghoment verifies that the MR loop is operating and circulating reactor coolant to ensure the capability of the MR system to maintain compitance with plant design limits. The required MR loop reactor coolant flowrate is deterstned by the flovrate necessary to:

(1) provide sufficient decay heat removal capability, (2) maintain the reactor coef ant tarperature rise through the core within design limits, for corpliance with flowrotes assumed in the boron dilution analysis.

(3) prevent thermal and boren stratification in the core, (4) preclude cavitation of the reactor coolant downstream of the MR flow control valve, are (5) ensure that inadvertent boren dilution events can be identified and tersinated by operator action prior to the reactor 4

returning critical.

In addition, during opetation of the MR loop with the water level in the vicinity of the reactor vessel nozzles, the MR loop flowrate i

deteritination must also consider the MR pwS suction requirements. At thir water level, the MR pure can experience vortexing or cavitation conditions which would cause the loss of MR purp operation, if the flowrite demand is too high. Operation with reactor coolant water at this level is eften called sid loop operation. Care sust be taken in determining the MR loop flowrate, when operating with water level in this region, to prevent loss of' the MR pnp and subsequent loss of the MR loop for decay heat removal.

0 0

l

(

l l

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

i SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................

3/4 5-1 avg 350*F...........................

3/4 5 3 3/4.5.2 ECCS SUBSYSTEMS - T FIGURE 4.5-1 RESIDUAL HEAT REMOVAL PUMP MINIMUM ACCEPT ABLE PERFORMANCE CURVE...............

3/4 5-6a 3/4.5.3 ECCS SUBSY$TEMS - T,yg <

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

3/4 5-7

gn,,,4 C.

3/4. 5./5 RE FUE LING WAT E R STO RAGE TANK.............................

3/4 5-ftl 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIKARY CONTAINMENT Containment Integrity....................................

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

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

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

3/4 6-6 A i r T e mp e r a t u r e..........................................

3/4 6-7 Contai nment Ves se) Structural Integri ty..................

3/4 6 8 Contai nment Purge Ventil ation System.....................

3/4 6-11 3/4.6.2 DEPRESSURIZATION AND C00LIRC SYSTEN5 Containment Spray System.................................

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

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

3/4 6-15 3/4.6.3 CONTAINMENT I SO LATION VALVES...............

3/4 6-16 TABLE 3.6-1 CONTAINMENT I SO LATION VALVES..........................

3/4 6-18 3/4.6.4 COM8USTIBLE GAS CONTROL Hydrogen Monitors........................................

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

3/4 6-26 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves............................................

3/4 7-1 BRAIDWOOD - UNITS 1 & 2 IX

_, = _. _.. _ - _ - - _ - _ _._. _ ___. _ _ _-. - _ - -.__

____..m.-_

5 Insert C 3/4.5.4 ECCS SUBSYSTEMS - Tavg LESS THAN OR EQUAL TO 200'r Pressuriser Level Greater Than 5 Percent (Level 409.5')

3/4 5 0 Pressurizer Level Less Than or Equal to 5 Fercent (Level 409.5')

3/4 5-10

, ~..

-v,w

,_-._,,,,....m.-.

y,,.., _,,,--,- -. - -,,,. --

..-ow,--.-,

y., -.y

BASES SECTION P,AGj 3/4.4.5 STEAM GENERATORS..........................................

B 3/4 4-3 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 3/4.4.9 PRESSURE / TEMPERATURELIMITS...............................

B 3/4 4-7 TABLE B 3/4.4-la REACTOR VESSEL TOUGHNESS (UNIT 1)................

B 3/4 4 11 TABLE B 3/4.4-1b REACTOR VESSEL TOUGHNESS (UNIT 2)................

B 3/4 4 12 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1NeV) 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 HDT FOR REACTOR VESSEL STEELS EXPOSED TO 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 3/4.5.1 ACCUMULATORS.,...........................................

4 an.s.s B 3/4 5-1 3/4.5.2 e 4 3/4.5.3 ECCS SUBSYSTEMS...............

3 3

B 3/4 5-1 3/4. 5. #'5 REFUELING WATER STORAGE TANK...........................

B 3/4 5-/4 s

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

B 3/4 6 1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS......................

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

B 1/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L...................................

B 3/4 6-4 b

BRAIDWOOD - UNITS 1 & 2 XVI

=,

1 BASES SECTION PAGE 3/4.9.6 REFUELING MACHINE.........................................

8 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 3/4.9.9 CONTAINMENT PURCE ISOLATION SYSTEM........................ B3/49/J 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE P00L............................................

B 3/4 9 3

)

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

B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARG I N......................................

8 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.....................................

t 3/4 10 1 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN.....................

B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS.........................................

B 3/4 11 1 3/4.11.2 GASEOUS EFFLUENTS........................................

8 3/4 11 3 3/4.11.3 SOLID RADI0 ACTIVE WASTES.................................

8 3/4 11 7 1 1 3/4.11.4 TOTAL 00SE...............................................

8 3/4 11 1 3/4.12 RADIOLOGICAL ENVIRONMENTAL HONITORING 3/4.12.1 MONITORING PR0 GRAM.....................

B 3/4 12 1 l

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

8 3/4 12-1 3/4.12.3 INTER LABORATORY COMPARISON PROGRAM.................

B 3/4 12-2 BRAIDWOOD - UNITS 1 & 2 XVIII l

REAC70R 800LANT Sv5 TEM SURVE!LLANCE RE0V!REMENis 4.4.9.3.1 Each PORV shall be cemonstrated 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 PORV is reovired OPERABLE and at least once per 31 days thereafter when the PORV is requirea OPERABLE; b.

Performance of a CHANNEL CALIBRATION on the PORV actuation channel atleastonceper18 months;(and

]

c.

Verifying the 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:

'M i

a.

For RHR suction relief valve,,8708B6 9 r

k 1}- By-te ri f yi ng--et-l e a s t-onc e-te r d ays -t ha t-- AHR-AC S-Guc t 40n-

-1 solation-Vahe RH8702A-it-open with-power-to4he-vahe-ope ra tor-removed r-4nd-

-2) N.By verifying at least once per[12 hours that RH87028 sala s b s? 4 = m A V

b.

For RHR suction relief valve,,8708Agr/

9

-1)-By-veef fying-a t-lees t-once-pe&31-days-that-AHB7018-4 6-open-with-poweedo-the-vahe-operator removed r-and-i x4

.n Vsim NP * ' * * "

-2)--ByverifyingatleastonceperlehoursthatRH8701AJs-open, s

3 c.

Testing pursuant to specification 4.0.5.

4.a 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 vent (s) is being used for overpressure protection.

Except when tne vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

-#The-speci fied-le-month-interval-may-be extended-to-32-months-for-cycle-1-onlyr BRAIDWOOD - UNITS 1 & 2 3/4 4-41 AMENDMENT NO. 4-

EMERGENCY CORE COOLING SYSTEMS t

SURVE1LLANCE REQUIREWEWTS (Continued) 1)

For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2) of the areas affected within containment at the completion of each contains,ent entry when CONTA!PMENT INTEGRITY is established.

l d.

At least once per 18 months by:

1)

Verifying automatic (solation and interloc6 action of the RHR

{

System from the Reactor Coolant System by ensuring that %

~

a)-With-a simulated or actual Reactor Coolant System pressure signal greater than or equG1 to 360 psig -the-inteelocks-prevents the valves f rom being openedf,and-

-4)

With_4 simulated.or-actual Asactor. Coolant-Jystea pressure.

-signal-greatar-than-or-equal-to-662--psig-the 4nteelocks-wil4-

-cause the-valves to-automatically-cleter i

2)

A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion, At least once per 18 months, during shutdown, by:

e.

1)

Verifying that each automt. tic valve in the flow path actuates to its correct position on a Safety Injection test signal and on' a RWST Level-Low-Low test signal, and 2)

Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signals a)

Centrffugal charging pump, b)

Safety injection pump, and c)

RHR punc.

By verifying that each of the following pumps develops the indicated f.

differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1)

Centrifugal charging pump l 2396 psid, 2)

Safety Injection pump 1 1412 psid, and In accordance with Figure 4.5-1 3)

RHR pump BRAIDWOOD - UNITS 1 & 2 3/4 5-5

3 EMERGENCY CORE COOLING SYSTEM o

ECCS SUBSYSTEMS - Tave LESS THAN OR EQUAL TO 200 F PRESSURIZER LEVEL GREATER THAN 5 PERCENT (LEVEL 409.5')

LIMITING CONDITION FOR OPERATION 3.5.4.1 All Safety Injection pumps shall be inoperable.

APPLICABILITY:

MODE 5 with pressurizer level greater than 5 percent, and MODE 6 with pressurizer levn.i greater than 5 percent and the reactor vesse'. head resting on the reactor vessel flange.

ACTION:

With d Safety Injection pump OPERABLE, restore all Safety Injection pumps to inoperable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS:

4.5.4.1 All Safety Injection pumps shall be demonstrated inoperable by verifylng 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 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.

-Dyron- - Units 1 & 2 3/4 5-9 Bnecb-A I

1

EMERGENCY CORE C00t!NG SYSTEM o

ECCS SUBSYSTEMS - Tave LESS THAN OR E.QVhk_TO 200 T FRESSURIZER LEVEL LESS TRAN OR EQUAL TO 5 PERCENT (LEVEL 409.5')

LIMITING CONDITION TOR OPERATION 3.5.4.2 At least one Safety Injection pump and flowpath shall be available, or the het side of the RCS must be adequately vented and have valve alignments to allow gravity feed from the RWST.

APPLICABILITY:

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

ACTION:

If neither Safety Injection pump is available and the het 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 regulred, by verifying at least on:e per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that 1)l switch in the pull out positien, and 2) the motor circuit breakers are racked in and open with the contro an OPERABLE flovpath exists from the RMST 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:

One of the following het side vent paths is available:

a.

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'r or less, or

3) Three pressurizer saftey 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'T 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'r or l

.less.

b. An OPERABLE flowpath that will permit gravity feed from the i

RWST is available.

%aGwk Byron - Units 1 & 2 3/4 5-10 r,

.-_~.-.

S.

EMERGENCY _ CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CON 0! TION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) and the heat traced portion l

of the RWST vent path shall be OPERABLE with:

a.

A minimum contained borated water level of 89%,

b.

A sinimum boron concentration of 2000 ppe, c.

A minimum water temperature of 35'F, and d.

A maximum water temperature of 100'F.

APPLICABILITY:

H0 DES 1.-2, 3, and 4.

ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or 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 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

\\-

SURVEILLANCE REQUIREMENTS 5

4. 5. ( The RWST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the contained borated water level in the tank, and 2)

Verifying the boron concentration of the water.

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 temperatur6 when the outside air temperature is either less than 35'F or greater than 100'F, and 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.

BRAIDWOOO - UNITS 1 & 2 3/45,9)l l

l

l REFUELING OPERATIONS 3/4.9.8 RESIDUAL HELJ RDCVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION F,0R (PERAT!0N 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation."

APPLICABILITY: W)DE 0, when the water level above the top of the reactor vessel fhnge is greater than or equal to 23 feet.

ACTION:

With no RHR loop 0FERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentra-tion of the Reactor Coolant System and imediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible.

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

e SURVE!LLANCE R[l@fXf Nf3 4.9.8.5 At Jesst-one401oop-sha114e-ver4f4+d4n-operation 4nd-circulating-react +c-cociant-et-e M:ve.vate of greater than or equel-to 2800 gpe-et-least-

-4nce -per-1240ure r j

h \\eg i m yt: G.ws eu MR \\c,c>p 6a.\\t W. WJI cck in opsycS'.s~

.tA ct:c c. Qtm& 5 qrechu su,,.

i 0..h 6: cn\\cE.

,Om io \\oct> p J 4 A,. 'tlCf5 h9 4 <c\\ess {a c.ce g ge

(

"The RriR loop may be remled from operation for up to I hour per 8-hour period

.during the performan::s c t CORi ALTERATIONS in the vicinity of the reactor vessel hot legs, l

BRAIDWOOD - UNITS 1 8 2-3/4 9 9

REFUELING OPERATIONS LOW WATER LEVEL LIMIT 1HG COWOITION FOR OPERATION t

3.9.8.2 Two residual heat removal one RHR loop shall be in operation.%(RHR) loops shall be OPERABLE, and at least l

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

ACTION:

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

corrective action to return the required RHR loops to OPERA 8LE 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 reduction in boron concentration of the Reactor Coolant System and imediately 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 atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.9.8.2 -At4 east-one-RHR-loop-shalbbe-verif4ed-in-operation-and-circulating-I

--re ac tor-coo l a nt-a t-a -fl ow-r a te-o f-gre a te r-than-o r - equa l-to -2800-gpe -a t4 ea s t- -

ance-pe r-12-hoves c J u + avet :ar \\tw s ca %(R !cep SL t! W ve.plie.cl m op<ohics CL Q C.a c d cd q G cha~.k ac c 5\\ew vche. o ? vsen e.s-A u c.r-e~ h Q

% te.o o og.s J & RC.I> te -e~eh ies s, du et e a$;. % d y,

> P ri o e-to -ini t i a l-c ri t4c a l i ty r-the-RHR-l oop - may - be.-removed -f rom -ope r a t ion-for-

-up-to-1-bou r-pe r-24our-pe r i od -du r i ng -the - pe e f o rea nc e-o f40RE-A LTERAT ION S-I n-

-the vicifttty-of-the-reactor-vessel hot legs.-

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BRAIDWOOD UNITS 1 & 2 3/4 9-10 3

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.,.,,.,e,.

-w.-,.,,.m,y.,

m

o 4

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPER.ATURE LIMITS (Continued)

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

Although the pressurizer operates in tamperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis perforined in accordance with the ASME Code requirmnts.

The OPERABILITY of two PORVs, or two RHR suction valves, or an RCS vent opening of at least 2 square inches ensures that the RCS will be protected from 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 PORY has adequate relieving capability to protect the RCS from overpres-suri2ation when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator ess 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 overshoots assuming a single PORY 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 PORV variable setpoint such that, for the two overpres-surization transients nmd, the resulting pressure will not exceed the nominal 10 effective full power years (EFPY) Appendix G reactor vessel NOT limits.

--RHR-RCS.4uction-4 solation-valves-8701A-and 87MA-are interlocked-withan-

"A"-train-wide-range-pressure transmittar-and-valves-87018-and-47028-are-4etet-t

-40cked-with-a FB"-train wide-range-pressure = transmitter Removing-power--from^

-valves-87018-and-8702Arprevents-a-single-failure froelnauertently isolating-

- bo t h - RH R -s uc t i on-re l i e fw e l ve s -wh i l e-sa i n ta i ni ng -RN R -4 s o l a t ion-< ap ab i l i ty-fee-l l

-both-RHR-f l ew-pa thsT-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 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 Comission pursuant to 10 CFR 50.55a(g)(6)(1).

BRAIDWOOD - UNITS 1 & 2 B 3/4 4 16

3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES _

3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a suf ficient volume of borated water will be imediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators.

This initial surge of water into the core provides the initial cooling mechanism during large RC5 pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are A contained borated water level between 31% and 63% ensures a volume of met.

greater than or equO to 6995 gallons but less than or equal to 7217 gallons.

The accumulate power operated is:uuon valves are considered to be 279 1971, which requires that

" operating bypasses" in the context c.f IEEE Std.

bypasses of a protective function be removed automatically whenever permissive In addition, as these accumulator isolatirn valves conditions are not met.

f ail to meet single f ailure criteria, removal of power to the vahes is 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 If a closed which may result in unacceptable peak cladding temperatures.

isolation valve cannot be imediately opened, the full capability of one accumulator i) 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.

u d 3/,.5 9 3/4. 5. 2. nd-3/4. 5. 3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one sub ystem through any single failure consideration.

Either subsystem operati g 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 In addition, double ended break of the largest RCS cold leg pipe ocwnward.

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 basis of the stable acceptable without single failure consideration on ti.]

reactivity condition of the reactor and the limited core cooling requirements.

BRA 10iOOD - UNITS 1 & 2 B 3/4 5-1

s 9

EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued) 'Id SRT'B

-The4 imitation for maximum 4fane4+ntr4 fugal--c+arging-pump-to-be-

--OPERABLE-and the Surve4444ece-Requireent-to-ver4 fy a14-charging pumps-4nd.

4afet; -Injection -pumos-except-the-required-OPERABLE-cheeging-pumeto be 4nepe--

abl+4elow 3304 provides455uranc+-thata. mass addition pressure-tr4n&4et--can be-relieved by the operat4on-of-444ngle RORb

{

The Surveillance Requirements provided to ensure OPERABILITY of each I

component ensures that at-a minimum, the assumptions used in the safety analyses I

are met and that subsysten OPERABILITY is maintained.

Surveillance Requirements for throttle valve 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 batween 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.0CA 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 518809 valve is closed to perform check valve leakage testing.

I 3/4.% t REFUELING WATER STORAGE TANK l

1 x

B 3/3 5 q !

The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures th'ats by the ECCS in the'a sufficient supply of borated water is available for injection 1

h e,nt of a LOCA.

The limits on RWST minimum volume and boron n

concentration ensure that(ing4't (1) sufficient water is available within containment to permit recirculation cool 10w to the core, and (2) the reactor will remain subcritical in the did condit g following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consfitent with the LOCA analyses.

The contained water volume limit inc a allowance for water not 4

Iusable because of tank discharge line location o(r other physical characteristics.

I A minimum contained borated water level of 89% ensureN volume of greater than or equal'to 395,000 gallons.

The limits on contained water volume and boron concentratio f the RWST also ensure a pH value of between 8.5 and 11.0 for the solution rect glated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on s

mechanical systems and components.

\\

BRA 10 WOOD - UNITS 1 & 2 B 3/4 5-2

L SE. RT 3

E The limitation f or a maximum of one centrif ugal charging pump to be OFEKABLE and the Surveillance Requirement to verif y all Charging pumps except the required 0FEKAfLE Charging pump to be inoperable in MODE 4 with one or more of the RCS cold legs less then or equal to 330'r, CODE 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 FORV or Rl!R suction relief vs1ve.

Similarly, the requirement to verify all Safety Injaction pumps are inoperable in MODE 4 with the temperature of one or more of the RCS Cold legs less than or equal to 3300p. in MODE $ with pressuriter level r,teater than 5 percent (Level 409.5') and in MODE 6 with pressuriser level greater than 5 percent and the reactor vessel head resting on the reactor vessel flange, provides assurance that a mass addition pressure transie9t can be relieved by a single PORV or RHR suction relief valve.

In MODE $ and M0tt 6 with pressuriser level less than-or equal to 5 percent, at least one Saf ety Injection pump or gravity f eed f rom the RWST must be available 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 f or the SI pump or gravity feed varies with RCS configuration and is, therefore, procedurally addressed.

The Surveillance Requir3ments-define what constitutes an adequate hot side vent for various pts.nt conditions.

It was determined that removing the reactor vessel head was an adequate vent under all conditions. Other venting alternatives have restrictions based on tims f rom shutdown and RCS temperature. The values in the surveillance were taken f rom the graph on the following page.

t

/sc1:0595T:27 1

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ~ - - -

EuggstNY CCRE CCCLING SYSTEMS BASES

_ECCS SUBSYSTEMS (Continued) 0 0

TRCS = 100 F 7 peg = 140 F 25.0

=

30.5 2 S AFETIES 30.0 f

[ '- ----

NOT l'

25.0 ACCEPTABLE

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a 3 S AFETIES (7

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=

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E s

g y

AC C E P,T A B L E 10.0 y

j PZR MANWAY -

5.0 f

0.0 0.0 100.0 200.0 300.0 400.0 5004 600.0 700.0 800.0 900.0 TIME AFTER SHUT 00NN (HRS)

Vent Path Requried to Prevent RCS Pressurization I

E G.> m C) tYRON - UNITS 1 & 2 B 3/4 5-3 k!

EWER 3ENCY CORE COOLING SYSTEMS BASES

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l i

I 5

3/a.5.hA REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of berated water is available for injection by the ECCS in the event of a LOCA.

The limits on RWST minimum volume and coron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain 4

suberitica' in the cold condition following mixing of the RWST and the RCS water volumes wiu. all control rods inserted except for the most reactive control assembly.

Ttese assumptions are consistent with the LOCA analyses.

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

A minimum contained berated 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 also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA.

This pH band sinimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

M~.:JK)

BYAON - UNITS 1 & 2 g3/45.gu AMENOMENTNO./

l

REFUELING OPERAT10k5 BASES 3/4.9.6 REFUEL,1NG MACHINE Tha OPEPABILITY requirements for the refueling machine and auxiliary hoist ensure that: (1) refueling machines will be used for movement of drive rods ind feel assemblies, (2) each refueling machine has sufflCient load capacity to lift 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 inadvertehtly engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY The restriction on movement of loads in excess of the nominal weight of a fuel 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 containti,in a y

single fuel assembly, and (2) any possible distortion of fuel in the storage

~

racks will not result in a critical array.

This 455umption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION 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 reacto vessel below 140'F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron diluti0n incident and prevent boron stratification, y EM M @

l The requirement to have two RHR loops OPERABLE when there is less then 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 he:t si^k is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core, 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment purge penetrations will be automatically isolated upon detection of high radiation levels within the containn nt, The OPERABILITY of this systtm is required to restrict the release of radioactive material from the containment atmosphere to the environment.

BRAICw000 - UNITS 1 & 2 B 3/4 9 2

]

lN S E RT A tveen/Basidweed Te dnical teneificatien Basei rytt11thee Ngdnuant verifies that the MR loop is operating and circulating reactor coolant to ensure the capability of the MR system to maintain coraliance with plant design limits. The required MR loop reactor coolant flowrate is detersined by the flovrate necessary to:

(1) provide sufficient decay heat removal capability. (2) saintain the r1 actor coolant tamperature rise through the core within design limits, for cow 9114nce with flowrates assuned in the boron dilution analysis.

(3) prevent thermal and boren stratification in the core. (4) preclude cavitation of the reactor coolant dornstream of the MR flow control valve and (5) ensure that inadvertent boron dilution events can be identified ar.d terminated by operator action prior to Os reactor returning critical.

In addition, during operation of the RhR loop with the water level in the vicinity of the reactor vessel nozzles, t5e MR lea flowrate deteruination siust also consider the MR pump suc'. ton requirements. At thir water level, ths MR puro can experience vorusing or cavitation conditions which would cause the loss of MR pump operation, if the flowrite dee.and is too high. Operation with reactor coolant water at this level is often called mid loop operation. Care must be t, den in determining the MR loop flowrate, when operating with water level in this region, to prevent loss of the MR pump and subsequent loss of the MR loop for decay heat removal,

(

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l

.o ATTACEMENT 2 DETAILED DESCRIPTION In response to the Recommended Action portion of Generic Lette, 66-17, Commonwealth Edison committed to submit any Technical Specification changes required te facilitate operation of the reactor coo 16nt system (RCS) at reduced inventory conditions.

Three changes were identified reduction of the residual heat removal (RHR) minimum flow rate, removal of the RHR autoclosure interlock on the RHR system suction isolation valves, and the 611owance of the operation of a safety injection pump to mitigate the effects of a loss of decay heat removal.

Minimum RHR Flowrate Mid-loop operation occurs when the plant is operating with the RCS partially drained in Modes 5 and 6.

The RCS water level is lowered in Modes 5 and 6 to facilitate removal and reinstallation of the reactor head during refueling outages. Operation with the RCS partially drained may also be necessary for the inspection ard mrintenance of RCS components such as reactor coolant pumps (1d steam gecer.:e-However, when the reacto: coolant level in the RCS loop pip.ng is lowerea, t here in a potential for air to be drawn into the RHR suction line (air (n.alnm-.t) due to RCS loop level fluctuations and/or the development of a vortex.

't entrainment into the RHR could cause air binding of t'ne RHR pumps and thus, result in the inadvertent loss of decay heat removal capability.

The tendency for vortex formation at the RHR suction line, and subsequent air entrainment into the RHR, is a function of the water level above the RRR suction nozzle and the RHR flowrate. The lower the level, or the higher the RHR flowrate, the greater the potential for a vortex to develop and air to be drawn into the RHR. Therefore, the likelihood of vortex formation due to partial draining of the RCS can be offset by reducing the RHR flowrate.

The required minimum RHR flowrate during mid-loop operations is based on the fo11 ewing concerns.

The ability of the RHR to remove decay heat such that RCS temperature can be controlled.

Sufficient flow is provided to ensure that reactor coolant temperature rise through the core does not exceed reactor vessel internals delta T limits.

Sufficient flow is provided to ensure that the re. actor coolant is mixed such that significant boron stratification does not occur.

Sufficient flow f* provided to ensure that the pressure drop across

?

.the RER bypass flow control valve does not result in cavitation.

Sufficient flow is provided to ensare that inadvertent boron dilution events can be identified and terminated by operator action prior to the reactor returning critical.

/cci:0595T:,

l

The first change involves-the minimum RER flow rate.

Currently, there is'no restriction on RHR' flow rates in Mode 5.

However, in Mode 6, a minimum flow rate of 2800 gpm is specified in Surveillance Requirements 4.9.8.1 (p. 3/4 9-9) and 4.9.8.2 (p. 3/4 9-10)... Connonvecith Edison proposes to reduce this flow rate to 1000 gpm. Westinghouse perf ormed an evaluation of this minimum flow reduction (WCAP-12207).

The Westinghouse analysis considered the f ollowing concerns:- (1) decay heat removal capability, (2) thermal considerations, (3) boron mixing and stratification, (4) control valve cavitation, (5) inadvertant boron dilution, and (6) the ef fect of reduced flow i

on RHR pwnp thrust bearing expected life.

The purpose of the RHR flow requirements and the potential for vortexing are described in a change to the Bases (p. B 3/4 9-2).

The additions to the Bases for 3/4 9.8 will cause the Bases for 3/4 9.9 to be moved to page B 3/4 5-2.

Changes to the Table of Contents are also included.

Also, we propose to delete a footnote to Specification 3.9.8.2 that applied prior to initial criticality.

This change is requetted for editorial purposes only.

The Westinghouse evaluation concluded that the minimum flow rate in the Technical Specifications could be eliminated. The evaluation stated that the minimum value.for RHR flow is dependent on plant conditions, therefore, no single value is accurate for all times. Westinghouse recommended that no flow rate be specified and that the required flow rate be controlled by plant procedures based on plant conditions.

However, the Nuclear Engineering l

Depa.tment recommended that a minimum value of 1000 gpm be placed in the Technical Specifications because Westinghouse did not consider flow rates lower than 1000 gpm in the non-LOCA accident analyses.

Engineering also recommended the addition of a Surveillance Requirement which verifies that the RCS temperature is being maintained at less than or equal to 140 degrees F.

Minimur, RHR flow rates for various plant conditions will be specified in proceaures.

The reduced RHR flow rate will primarily be used to facilitate mid-loop operations which are covered under Specification 3.9.8.2, Mode 6 with the water level less than 23 f eet above the reactor vessel flange. The revised flowrate is also requested for Specification 3.9.8.1 (Mode 6 with the water level greater than or equal to 23 feet above the flange) for consistency in l

Surveillance Requirements-and to al'.ow operational flexibility.

The Westinghouse evaluation showed that operation of the RHR pumps at lower than 3000 gpm results in decreased thrust bearing life. Therefore, operation at reduced flow rates will be minimized.

The RHR pumps are run at low flow rates (approximately 500 gpm) several times per year for survelliance testing or due to automatic initiations. When run at these low flow levels, operatius is limited to under 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (typically 30 minutes).

The only extended operation (over three hours) at low flow rates (as low as 1000 gpm) would be l_

during mid-loop operation.

However, mid-loop operation is not common at l

Byron /Braivwood due to tlie presence of loop stop valves. The loop stop valves can be used to isolate individual loops as necessary to perf orm maintenance i

that would otherwise require mid-loop operation.

l l

/sc1:0595T:5

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+

Westinghouse found that dus to_ the shape of she RHR hydraulic thrust curve, any slight drop in the flowrate below 3000 spm or minor variation in pump _

hydraulic characteristics can result'in a significant reduction in thrust bearing' life. -Any increase in wear on the thrust bearings induced by increased low flow operations will be taken into account in the maintenance and replacement programs'for the thrust bearings. As an additional measure to j

-reduce thrust bearing wear, normal operating flow will be increased to 3300 gpm. This increased flow will prevent minor flow varjations from reducing flow below 3000 gpm (the flow at which increased pump wear begins).

Etmoxal _ of the_Au10el0 Ant _InitIlod The sec*nd proposed change involves the removal of the autoclosure interlock (AC1) on the RRR system suction isolation valves.

Inadvertant actuation of the ACI function has been a significant contributor to loss of decay heat removal events. The purpose of the ACI is to close the RilR isolation valves (RH8701A,_RH8701B, RH8702A and RH8702B) if the pressure in the RCS rises above approximately 662 psig. The RhR isolation valves are slow acting (clnse within two minutes) and, therefore, are not relied upon for overpressure protection. The RHR relief valves are used to mitigate a pressure transient, such as an unexpected pump start.

Each train of isolation valves is interlocked with a wide range RCS pressure transmitter.

Failure of a pressure transmitter can cause the isolation of one train of-RHR.

This can result in a loss of decay heat removal if only one train of RHR is operating and that train becomes isolated.

Each transmitter is interlocked with one valve on each RHR train.

In order to prevent the failu e of a single transmitter from isolating both RHR trains, one valve in each train (Rhd702A and RH8701B) is required to'be open with._the power to the valve operator removed. This assures that at least one train of RHR relief valves are available for overpressure protection. Once the ACI is removed the potential-for a spurious closure of the 1sniation valve is greatly reduced.

-Therefore,. removing the ACI will increase the reliability of the RHR system.

Removing the-ACI requires several changes to the Technical Specifications and

-Bases. First, we propose to delete the references in surveillance 4.4.9.3.2 (p. 3/4 4-41) which state that valves RH8702A and RH8701B must be open with the power removed.

These references will no longer be necessary after the ACIs are removed because the transmitters will no longer be able to close the isolation valves.

In addition, we request a change to the surveillance

.intervel in Specification 4.4.9.3.2 from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The purpose of this-surveillance is to assure that the isolation valves are open when the j '

RHR suction relief valv_es are being used for cold overpressurization.

Increasing the surveillance interval is justifiable because the probability I

that the isolation valves could be inadvertently closed is greatly reduced after the removal of the ACI.

Surveillance 4.5.2.d.1.b (p. 3/4 5-5) requires testing the ACI function. - Once I

the ACI function is removed, this surveillance will no longer be needed.

Thergfore, we request that this specification be deleted.

In addition, the reference to the ACI function for the RHR isolation valves in Bases Section

-3/4.4.9 (p.

B 3/4 4-16) should also be deleted since this information will no longer be applicable once the ACI is removed. The details of each of these changes is described below:

/sc1:0595T:6

. ~...

~

Changes (1) through (6) are applicable to Technical Specification 3/4.4.9.3, "RCS Overpressure Protection Systems". Change (7) applies to Technical-Specification 3/4.5.2, "ECCS Subsystems - Tavs > 350 degrees F".

Change (1) deletes the surveillance recuirement, 4.4.9.3.2.a.1, which requires that the RHR RCS suction isolation valve RNB701A be verified open with power to the valve operator removed at least once per 31 days whenever the RHR suction relief valves are being used f or cold overpressure protection.

Change (2) renumbers surveillance 4.4.9.3.2.a.2 to 4.4.9.3.2.a and adds valve RHB702A to the surveillance which requires that valve RH8702B be verified open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the RHR suction relief valves are being used for cold overpreasure protection.

Adding valve RH8702B to the surveillance vill ensure that free communication exists between the RCS and the RHR system when the relief valves are providing cold overpressure protection.

The surveillance interval is also being changed f rom once per every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to once per every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months surveillance interval is consistent with the surveillance interval for verifying the Pressurizer Power Operated Relief Valve (PORV) isolation valves are open when using the PORVs for cold overpressure protection.

Change (3) (similar to change (1)) deletes the surveillance requirement, 4.4.9.3.2.b which requires that valve RH8702B be verified open with power to the valve operator removed at least once per 31 days whenever the RHR suction relief valves are being used for cold overpressure protection.

Change (4) [similar to change (2)] renumbers surveillance 4.4.9.3.2.b.2 to 4.4.9.3.2.b and adds valve RHS701B to the surveillance which requires that valve RH8701A be verified open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the RHR suction relief valves are being used for cold overpressure protection.

Similarly, the surveillance interval is changed f rom once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Change ($) deletes the discussion of the requirement to remove power from valves RH8701B and RH8702A f rom the Bases of Section 3/4.4.9.3.

Change (6) deletes a footnote from Specification 4.4.9.3.1.b that applied to Cycle 1 only. Also, the system identifier "RH" was added to the relief valves l

8708A and 8708B in Specification 4.4.9.3.2.

Theso changes are requested for editorial purposes only.

Change -(7) deletes the words " isolation and" f rom surveillance requirement 4.5.2.d.1.

The surveillance requirement 4.5.2.d.1.a is renumbered to be l

incorporated into 4.5.2.d.1.

The surveillance currently requires that the automatic isolation and interlock action of the RHR system from the RCS be verified every 18 months by ensuring that with a simulated or actual RCS pressure signal greater than or equal to 662 psig that the interlocks will cause the valves to automatically isolate.

In addition, the words "with a" are replaced by the word "any", the words "the interlocks" are deleted and

" prevent" is changed to " prevents". This change clarifies the intent of the surveillance to assure that the interlock prevents the isolation valves from opening after receiving any signal indicating RCS pressure greater than or equal to 360 psig.

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The Westinghouse Owners Group -(WDC) p3rformad cn analysis justifying the rsmoval of the ACI for four reference plants. The Nuclocr Enginasting Department performed an evaluation comparing the design of the ACI at Byron /Braidwood with its reference plant, Callaway.

Several differences were

-identified in the design of the basic piping, logic and wiring diagrams of the ACis at the two plants. A detailed review determined that none of_the differences had a significant. impact on the applicability of the WOG analysis to Byron /Braidwood. Therefore the WOG analysis may be used as justification to oupport the removal of the ACI at Byron /Braidwood.

Analyses were performed to demonstrate the impact of removal of the ACI on Event V (Loss of Coolant Accident outside containment through the RHR relief valves) f requency, RHR system reliability and overpressure transients. The analyses performed compared the results with and without'the ACI. However, the results were contingent upon providing an alarm to alert the operator that a RCS-RHR series suction isolation valve (s) is not fully closed and that double isolation is not being maintained. The modification will not impact the opening circuitry, nor will it effect the MOV position indication in the control room.

The setpoint f ce the alarm will be within the range of the open permissive setpoint pressure and the RER-system design pressure minus the RER pump head pressure. Operating precedures will be revised to direct the operator to take the necessary actions to clo.9 the open valve (if it is not L

closed), or if this is not possible, to return.o the safe shutdown mode of operation.

The analyses performed indicates an overall increase in safety due to the removal-of the ACI, implementation of the modification, and procedural changes.

Safetv Inleetion Pm p_Operab111ty_in Mode.s 5 and 6 The proposed Technical Specification changes are intended to provide clarification on the status of the Saf ety Injection (SI) pumps in Modes 5 and 6.

Currently, the Tec vteal Specifications make no reference to the status of the SI pumps in MODES 5 and 6.

The new specifications also protect the RCS against cold over-pressurization and ensure the availability of at least one SI pump or gravity feed from the RWST to mitigate the effects of a loss of decay heat removal during mid-loop operations. The proposed changes provide for the addition of two new Technical Specifications and Bases, which are as follows:-

3/4.5.4.1' ECCS SUBSYSTEMS Tavg LESS THAN OR EQUAL TO 200* F, PRESSURIZER LEVEL GREATER TRAN 5 PERCENT (LEVEL 409.5') (p. 3/4 5-9), and 3/4.5.4.2 ECCS SUBSt TEMS Tavg LESS THAN OR EQUAL.TO 200' F. PRESSURIZER LEVEL LEF "HAN OR EQUAL TO 5 PERCENT (LEVEL 409.5') (p. 3/4 5-10)

In addition, this requires renumbering the REFUELING WATER STORAGE TANK

-Technical Specification and Bases to 3/4.5.5 (p. 3/4 5-11). Additions to the Bases for 3/4 5.4 will cause the Bases for 3/4 5.5 to be moved to page B 3/,4 5.4.

Changes to the Table of Contests z.re also included.

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Technical Specification 3/4.5.4.1 requires that all SI pumps be demonstrated INOFERABLE in MODE 5 with pressuriter level greater than 5 percent and in MODE 6 with pressurizer level greater than 5 percent and the reactor vessel-head resting on the reactor vessel flange.

The purpose of this Technical Specification is to specifically address the status of the SI pumps in MODES 5 and 6 with consideration to cold overpressure protection.

The associated ACTION STATEMENT requires that with a SI pump OPERABLE that all SI pumps be restored to inoperable status.within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS state that all SI pumps are to be demonstrated inoperable once.per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the motor circuit breakers are secured in the open position.

Technical Coccification 3/4.5 4.2 requires that at least one SI pump and flow path or an RCS hot side vent s.iat will permit gravity feed f rom the-Refueling Water Storage Tank (PWST) be available while in either MODE 5 or 6 with pressurizer level less than or equal to 5 percent.

The purpose of this Technical Specification 16 to ensure the availability of at least one SI pump or gravity feed from the RWS1 to mitigate the effects of a loss nf decay heat removal event during mid-loop sperations. The flowpaths for the SI pump and gravity feed are dependent on plant conditions and, therefore, are addressed by plant procedures.

The assor.iated ACTION STATEMENT requires with neither: SI pump available or without an adequate hot side vent, that corrective actions be initiated immediately to restore either condition or establish pressurizer level greater than 5 percent.

SURVEILLANCE REQUIREMENTS state that at least one SI pump is to be demonstrated available, when required, by verifying that the motor circuit breakers are racked in and open with the control switch in the ' pull out" position at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Also, 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 one of the following hot side vent paths is available:

a.

The reactor vessel head is removed, or b.

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.

c.

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.

d.

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.

The abo'

requirements were determined in a Westinghouse analysis of the therma. hydraulic response to a loss of RHR (CAE-89-155 CCE-89-152).

The graph f rom which the values in the Surveillance Requirement were taken is included in the Bases for the new specification.

The above proposed Technical Specifications are similar to those approved for Callaway. However, there are dif ferences in the surveillance interval and the water level reference. The Callaway Specification 3/4.5.4.1 requires that the SI pumps be demonstrated inoperable at least once per 31 days. The proposed Byron /Braidwood Sp cification would require that the SI pumps be demonstrated inoperable every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This interval is considered more appropriate considering the relatively short time that the plant will be operated in this mode. This change is more conservative than the Callaway specification because it assures compliance with the limiting condition for operation on a-more frequent basis.

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Callaway used the flange elevation as the water level reference in the applicability for both new specifications.

Byron /Braidwood Station decided to use five percent pressurizer level (level 409.5') as the reference water level.

The reason for this dif f erence is that under various circumstances, the water level may need to be raised or lowered several times when the level is at the flange elevation.

If the flange elevation was used as-the reference level, operators would have to rack the breaker in or out (to make the SI pump either available or inoperable) each time the level was moved above or below the flange, fewer plant activities that require level changes occur at a pressurizer level of five percent.

This level will still provide the necessary eurge volume in the pressuriter while minimizing the burden on operations.

Another difference between the Callaway and the proposed Byron specifications is the option to use gravity feed from the RWST instead of an available SI pump if the RCS is properly vented. This option permits operational ilexibility while still assuring adequate capability to mitigate the 1

consequences of a loss of-RHR during mid-loop operations.

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ATTACHMENT 3 j

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10CFR50.92 SICNIFICANT RAZARDS CONSIDERATION EVALUATION c

g FOR PROPOSED TECHNICAL SPECIFICATIONS TO PREVENT

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LOSS OF DECAY HEAT REMOVAL EVENTS AT BYRON UNITS 1 AND 2 JE 1

2.".L' re

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t 1 IntroductiDD

.as As required by 10CFR50.91(a)(1) this evaluation is provided to demonstrate that a proposed license amendment to reduce the minimum required Residual yeat Removal (RHR) flowrate during mid-loop operation, to remove the RHR

'ht autoclosure interlock and to add SI pump operability requirements in Modes,5 and 6 represents no significant hazards consideration.

In accordance wigh the three factor test of 10CTR50.92(c), implementation of the proposed Jicpna3 amendment was analyzed and found not to:

1) involve a significant iperea9e in the probability or consequences for an accident previously evaluatedl;er-2) create the possibility of a new or different kind of accident from anyt accident previously evaluated; or 3) involve a significant reductiontfp a, margin of safety. Each of the three groups of changes was evaluated ms ;y 1

follows.

n rm.

T vi t-t Minimum RHR Flowrg13 a 3, a so.

Conformance of the proposed amendments to the standards for a determingflon of no significant hazard as defined.in 100F150.92 (three f actor test) is shown in the following:

1.

Operation of the Byron /Braidwood. Units 1 and 2, in accordance w(Ah,the proposed license amendment does not involve a significant increase kp;qfe, probability or consequences of any accident previously evaluated.

emove.

A reduction in RHR flow during mid-loop operation will potentially 43pqp4:

those transients explicitly analyzed in Modes 5 and 6.

The only event analyzed for these modes in Chapter 15 of the-Byron /Braidwood UFSARufAn. he f

malfunction of the CVCS that results in a decrease in boron concentfabten,in the. reactor coolant. The CVCS malfunction event can be impacted by a reduction in RHR flow'in the following two areas:

1) A reduction in fxpd4 cit RHR flowrote assumptions and 2) the vessel mixing assumption during a poron dilution.

The Mode 5' and 6 analyses -do not assume an explicit RHR floy yelue, and the RHR flowrates are assumed to be sufficient to provide adequate vessel circulation to prevent boron stratification and support the boron dilution transient mixing assumptions.

3 ;,

u.

't In addition, since a CVCS malfunction event in Mode 6 is prevented by

(

administrative controls which-isolate the RCS from any potential sourqe of unborated water, only the Mode 5 analysis could be impacted.

However itehas n

been' determined that a reduced RHR flew of 1000 gpm or greater would ngt,

invalidate thi Byron accident analysis assumptions. Therefore, the cuqtapt Mode 5 analysis remains valid.

11.c dA

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m 2.

The proposed license omendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Acceptable RHR flowrates that are consistent with the plant conditions would ce specified in the plant procedures. The RHR flowrates would be such that:

1) the RHR would be capable of decay heat removal to control the RCS temperature, 2) the reactor coolant temperature rise through the core would not exceed reactor vessel internala delta T limits, 3) the reactor coolant would be mixed to prevent significant boron stratification f rom occuring, 4) the fressure drop ac"oss the RHR flowrate control valve would not result in eavitation, and 5) inadvertent boron dilution events could be identified and terminated by operator action prior to the reactor returning critical.

Thus, a reduction in RHR flow would not increase the probability of a CVCS malfunction event and the possibility of an accident which is different than any aircady evaluated in the UFSAR would not be created.

3.

The proposed license amendment does not involve a significant reduction in a margin of safety.

l Currently, the Byron /Braidwood Technical Specifications do not specify RHR flowrate requirements for operation in Mode 5.

Mode 6 operations, however, require a minimum RHR flowrate of 2800 gpm (Survelliance Requirements 4.9.8.1 and 4.9.8.2).

The Technical Specifications place limitations on the RHR during mid-loop operation by specifying a minimum flow requirement for the purpose of decay heat removal and the number of RHR trains which must be operable.

They do not, however, contain restrictions based on minimizing air entrainment in the RdR as a result of vortexing which may occur during mid-loop operation under certain conditions.

The fuel cladding (fission product barrier) is protected in Modes 5 and 6 by providing cooling and maintaining core shutdown. Adequate decay heat removal is provided to address the cooling requirements, and sufficient mixing ensures that the boron dilution analyses remain valid.

Therefore, the amount of time available to identify and terminate a boron dilution event is unaf f ected.

Thus, a reduced RHR flowrate during mid-loop operation does not involve a significant reduction in a margin of safety.

ReconLaL1 h e._Au t os i nsu re_Intulod The following is an evaluation of significant hazards consideration to the changes for removal of the autoclosure interlock:

1.

The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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4..

., r c The RHR suction relief valves are used as e means of cold overpressure protection.

The cold' overpressure protection system is designed to ensure the limits of Appendix 0 to 100FR Part 50 are not exceeded when one or more of the RCS cold legs are less than or equal to 350 degrees F.

Transient analyses were performed to determine the worst case mass input and heat input events (refer to UFSAR, Section 5.2.2.11.2).

Removal of the ACI does not impact the transient analyses.

However, removal of the ACI helps ensure that the RHR suction relief valves are available to mitigate potential overpressure transients. Additionally, removing the ACI reduces the potential for inadvertent isolation of the RER. system which can cause a Low Temperature Overpressure (LTOP) transient (reduced letdown combined with a loss of decay heat removal) while also isolating an overpressure mitigation path.

Therefore, removal of the ACI does not involve an increase in the probability or the consequences of an ac:ident previously evaluated.

In fact, removal of the ACI has a positive impact on LTOP mitigation.

Analyses were also performed to confirm that one RHR relief valve has the capability of maintaining the RER system maximum pressure within code limits (refer to UFSAR, Section 5.4.7.2.3).

Removal of the ACI does not affect this analyses.

Should a peak pressure occur while the RER system suction isolation valves are open, the pressure effect on the-low pressure RHR system would be mitigated by the RHR suction relief valves. The deletion of the ACI feature has no ef fect on the ability of the RER system to survive pressure transients when the RHR system is connected to the RCS, since the RHR suction isolation valves are slow acting and no credit is taken for their actuation.

Therefore, removal of the ACI will not involve an increase in the probability or consequence of an accident previously evaluated.

The impact of the ACI to Event V, LOCA outside containment, frequency was also considered. Analysis demonstrates that the probability of the occurrence or consequences of an accident are not increased.

The dominant failure mode is rupture of the valve disc in each of the two series motor-operated valves (MOVs) in the RHR suction line when closed during normal power operation.

This failure mode is independent of the ACI. Another less influential contributor to Event V frequency was found to be rupture of one valve while the other valve has failed open. The results demonstrate that, in this case, removal of the ACI is beneficial when compared to retaining it.

2.

The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously considered.

The effect of an overpressure transient will not change due to the removal of the ACI.

The RHR suction relief valves were designed to maintain the RHR system pressure within design limits. Although the ACI isolates the RCS from the RHR suction relief valves on high RCS pressure, overpressure protection of the RHR system is provided by the RHR suction relief valves not by the slow acting suction isolation valves. The purpose of the interlocks is to assure double isolation between the RHR system and the RCS when the plant is at normal operating conditions. The interlock prevents the possibility of an Event V due to operator error.

l

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L.

r*v Removal of the ACI will not place _ths plant in any new or unanalyzed

~

condition. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously considered.

3.

The proposed amendment will not involve a significant reduction in a margin of safety.

Analyses were performed to demonstrate the impact of removal of the ACI on Event V frequency, RHR system reliability and overpressure transients.

je analyses performed compared the results with and without the ACI. However, the results were contingent upon providing an alarm to alert the operator that a RCS-RHR series suction isolation valve (s) is not fully closed and that double isolation is not being maintained. The modification will not impact the opening circuitry, nor will it effect the MOV position indication in the control room. The setpoint for the alarm will be within the range of the open permissive setpoint pressure and the RHR system design pressure minus the RHR pump head pressure. Operating procedures will be revised to direct the operator to take the necessary actions to close the open valve (if it is not closed), or if this is not possible, to return to the safe shutdown mode of operation.

The analyses performed indicates an overall increase in safety due to the removal of the ACI, implementation of the modification, and procedural changes.

Analyses performed indicates that the reliability of the RHR system is unchanged during RHR initiation and that it is improved during short and long term coolin6 as a result of the deletion of the ACI. Therefore, the margin of safety has actually increased.

Safety Injection Pump Opergb_111tv in Madgs 5 and 6 The following is an evaluation of significant hazards consideration relative to the proposed changes for SI operability in Modes 5 and 6:

1.

The proposed amendment does not involve a significant increase in the probability o: consequences of an accident previously evaluated.

The probability of the occurrence of an accident is net increased since proposed Technical Specification 3/4.5.4.1 requires that all SI pumps be demonstrated inoperable in MODE 5 with pressurizer level greater than 5 percent and in MODE o with pressurizer level greater than 5 percent and the reactor vessel head resting on the reactor vessel flange due to cold overpressure protection concerns.

In addition, proposed Technical Specification 3/4.5.4.2 requires the availability of at least one SI pump or an RCS hot side vent in MODE 5' and MODE 6 with pressurirst level less than or equal to-5 percent to mitigate the consequences of a loss of decay heat removal event during mid-loop operations. The availability of an SI pump under these circumstances does not present-a concern regarding cold overpressure protection since aufficient air volume-exists which allows the -

operator time to mitigate the transient. This is in contrast to the analyzed cold' overpressure transients, in which the RCS is assumed to be water solid at the onset of the event. To prevent a SI pump from inadvertently being started by a signal, but allow them to be manually started from the control room, surveillance requirements include verifying that the motor circuit breakers are racked in and'open with the control switch in the " pull out" position.

Therefore, the occurrence of an accident previously analyzed in the Final and Updated Safety Analyses Reports is not increased.

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The consequences of en accident es previously enalyzed in the Final and Updated Safety Analysis Reports are not increesed. However, the availability of at least one SI pump provides for the mitigation of the effects of a loss of decay heat removal event during mid-loop operrtions.

It has subsequently been demonstrated that for some cases, i.e. the combination of inadequate RCS venting and the existence of a cold leg opening, operation of at least one safety injection pump is required to prevent the core f rom uncovering.

The option to vent the RCS and use gravity feed from the RWST has been analyzed and will have no impact on the probability or consequences of a previously analyzed accident.

2.

The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously considered.

The provisions of these Technical Specification changes are for the purpose of mitigating the consequences of a loss of decay heat removal during mid-loop operations. Operation of at least one SI pump is required in some cases to prevent the core from uncovering as supported by the performance of thermal hydraulic analysis. The only new configuration allowed by this specification is the potential of having a SI pump available in Modes 5 and 6.

The potential overpressurization accident has been analyzed and accounted for in the specification by requiring pressurizer level to be less than 5 percent if a SI pump is available. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously considered.

3.

The proposed amendment does not involve a significant reduction in a margin of safety.

Operation of the SI pumps under these circumstances is not a cold overpressure.

concern because of the amount of air volume which exists in the RCS which would allow the operator time to mitigate the transient. This is in contract to the analyzed mass addition transient which assumed a water solid RCS. To prevent the pumps f rom being tnadvertent.ly starting by a signal, surveillance requirements require verifying that -th4 fircuit breakers are racked in and open with the control switch in the " pull out" position. Having either a SI pump available or gravity feed f rom the RWS7 vill help mitigate the consequences of a loss of decay heat reroval event. Therefore, the margin of safety is not significantly reduced.

Conclu11cn Based on the preceding evaluations, it la concluded that operation of Byron /Braidwood Units 1 and 2 in acesrdance with the proposed amendment, does not create an unroviewed safety quection. increase the probability of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, nor reduce any margins to plant aafety. Therefore, the license amendment does not involve a Significant Hazards-Consideration as defined in 10CFR50.92.

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ATTACEMENT 4 e

ENVIRONMENTAL ASSESSMENT Commonwealth Edieon has evaluated the proposed amendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10CFR51.21.

It has been determined that the proposed change meets the criteria for a catagorical exclusion as provided for under 10CTR51.22(c)(9)

This proposed amendment involves three groups of changes:

1) reduction of the RHR flow requirement in Mode 6, 2) removal of the autoclosure interlock on the RER system suction isolation valves, and 3) operability for SI pumps in Modes 5 and 6.

The purpose of these changes is to facilitate operation of the RCS at reduced invente y conditions. All changes are consistent with the guidance provided by the NRC in Generic Letter 88-17 and industry initiatives to mitigate loss of decay heat removal events.

The proposed. change does not involve a significant hazards consideration as discussed in Attachment 3 to this letter. Also, this proposed amendment will not involve significant chenges in the types or amounts of any radioactive ef fluents nor does it affect any of the permitted release paths.

Its addition, this change does not involve a significant increase in individual or cumulative occupational exposure. Therefere, this change meets the categorical exclusion permitted by 10CFR51.22(c)(9).

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