ML20082A284

Proposed Tech Specs 3/4.4.4 & 3/4.4.9.3,addressing Concerns of Generic Ltr 90-06 Re PORVs & LTOP

ML20082A284
Person / Time
Site:	Byron, Braidwood
Issue date:	06/28/1991
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20082A261	List: ML20082A259 ML20082A284
References
GL-90-06, GL-90-6, NUDOCS 9107100260
Download: ML20082A284 (27)

Text

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I REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES I LIMITING CONDITION FOR OPERATION Mk 3.4.4 gpower-operated relief valves (PORVs) and thair associated block valves shall be OPERABLE.

APPLICABILITY:- MODES 1, 2, and 3 ACTION: #1 4A/tfA}ddhdddf0Ih' k aWd '

t a.

With one or more PORV(s) inoperable because of excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(si to OPERABLE status or close the associated block valve (s)'; otherwise be in at least HOT STANDSY L

within 6' hours and in g SHUTDOWN within the following hours.

b. With one PORV inoperable due to causes other than excessive seat

( leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve;-restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in E% 9-H O T SHUTDGWNwithinthefollowingghours.

+- c. WithbothPORV[sfinoperabledueto'causesotherthanexcessiveseat l685tCAM leakaoe. within~1 hour either restore u d :f tL PORV(qF to OPERABLE jig "

status or close Dek-associated block valveM and remove power from the in block valve W and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and W^ SHUTDOWNwithinthefollowing{ hours.

d. re: ;r e,e uiou vaive(s) inuvelc, within 1 huui.

1) re he block valve (s) to OPERABLE status # esen ne block q

• • 'g valve (s) and muva pqntthaAleck v3PTels), or close the PORY g g mnd r; = : pcwei from 1ts assoltated-solanald__valytLapd 2) apply the SCTI^" ei u, y r-above as annmnMMn f;r th; i;;l;t;d 707j Q

e. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, eacn PORV shall be demonstrated OPERABLE at least once per 18 months Dy:

a.

6 $6It&& 04 4 AM Performance of a CHANNEL CALIBRAfION . "

)

7 Operating the valve through one complete cycle b li trave .

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per  % ans.

92 days by operating the valve through one complete cycle of full travel unless

-the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. of Specification 3.4.4.

"h; am iiiea 15 mv.,t',

ir,tarval may--be ;xte-de': t0 32 esnth5 'Graycle 1 erly BRAIDWOOD - UNITS 1 & 2 3/4 4-12 9107100260 910628 PDR AMENDMENTNO.[

P ADOCK 05000454 PDR

- INSERT A l

d. With one or more block valves inoperable, within i hour restore the block valve (s) to OPERABLE status or place its associated PORV in manual control. Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore any remaining inoperable block valve to OPERABLE siatus within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

INSERT B:

b. Operating solenoid air control and check valves on associated air accumulators in the PORV control system through one complete cycle of full travet, qd

/scl:lD615:51

REACTOR COOLANT SYSTEM f

OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4. 3 At least one of the following Overpressure P.rotection Systenpr/shall be O RABLE: /

a,

/

Tworesidualheatremoval(RHR)suctionreliefvalv/esiachwith tpoint of 450 psig 1%, or

/

b. Two wer-operated relief valves (PORVs) with li,ff. Setpoints that vary 4th RCS temperature which do not exceed the limit established in Figu 3.4-4, or

/

/

c. The Reactor oolant System (RCS) depressuf'ized with an RCS vent of greater than equal to 2 square inched

/

APPLICABILITY: MODES 4 an 5,andMODE6wittythereactorvesselheadon.

ACTION:

a. With one PORV and one R N suct on relief valve inoperable, either restore two PORVs or two R/suction rel'ef valves to OPERABLE status within 7 days or depressu $e and vent tra RCS through at least a 2 square inch vent within tri next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b.

/

With both PORVs and both RHR su ion relief valves inoperable, depressurize and ve t'the RCS th Nugh at least a 2 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c. In the event the"PORVs, or the RHR suc ion relief valves, or the RCS vent (s) are us,e'd to mitigate an RCSe pr(ssure transient, a Special Report shall fte prepared and submitted togthe Commission pursuant to Specificatipn 6.9.2 within 30 days. The r ort shall describe the circumstapdes initiating the transient, the ffect of the PORVs, or the RHR puction relief valves, or RCS vent (s) n the transient, and any co fective action necessary to prevent rec rence.

d. The rovisions of Specification 3.0.4 are not app icable, i f i i

BRAIDWOOD - UNITS 1 & 2 3/4 4-39

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECT 10N SYSTEM

.IMIT]NG CONDITION FOR OPERATION REEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEEWEEEEEEEEESEEEEEkES 3.4.9.3 At least two overpressure protection devices shall be OPERABLE, and each device shall be either:

a.

A residual heat removal (RHR) suction relief valve with a lift setting cf less than or equal tc 450 psic, or

b. A power operated relief valve (PORV) with a lift setpoint tnat varies with RCS temperature which does not exceed the limit establishec in Figure 3.4-4. ,

APPLICABILITY: MODES 4, 5, and 6 with the reactor vessel head on.

ACTION:

a. With one of the two reavited overpressure protection cevices inoperable in Moce 4. restore t.o overpressure protection devices to GPERABLE status witnin 7 days or cepressurize and vent the RC5 tnrough at least a 2 square inch vent within the next 8 nours.

b. With one of the two reauired overpressure protection devices inoperable in MODES 5 or 6, restore two overpressure protectiop devices to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or vent the RCS througn at least a 2 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c. With botn of the recuired overpressure protection devices inoceratie, depressurire anc vent tne RCS througn at least a 2 square incn vent witnin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c. With the RCS vented per ACTIONS a, b, or c, verify the vent path.ay at least once per 31 days when the patn ay is providec Dy a valve (s) that is locked, sealed, or otherwise secured in the open position; otherwise, verify the vent patnway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

e. In the event either the PORVs, RHR suction relief valves, or the RCS vents are usec to mitigate an RCS pressure transient, a Special Report shall be preoared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs, RHR suction relief valves, or RCS vents on the transient, and any corrective action necessary to prevent recurrence,

f. The provisions of Specification 3.0.4 are not applicable.

SasERT c d

.v

- REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS _7 h W Y $AL AY & wtst \

4.4.9.3.1 Each PORV shall be demonstrated OPERABLEk by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operatien,%eHFi- 21 day; vier +n anter %g e condition ia which th: POP" ;; ,equ;r;d OPE MSLE

.and at least once per 31 days h ert ftee when the PORV is required OPERABLE; Q

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

c. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

shan +ha DOR" u being uted 'er escrprc :ur; prettetwm 4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:

a. For RHR suction relief valve RH87088 verify at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that valves Rf;8702A and RH87028 are open.

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

c. Testing pursuant to Specification 4.0.5.

4.4.-0.0.0 -The "0 cent (s) shall L ;;r ' icd-to be spen-st--icost once pcr i

5n"re* uhe H hu veni,(3) is being used4er-cycrpre3sure pcotcction.

,m 3 ^ ^ -

• C ( 'k ba 'f e f + --pathi;;y 3 prOVN d "#th ? V3lVO MO i0 EIE '

er nherwisa cecur4d-in the aperr+00itie", the' ;cr i fy thec valvc; Open 5fu-Jccet once per H *y" BRAIDWOOD - UNITS 1 & 2 3/4 4-41 AMENDMENT NO.

U;'C10R COOT 1M 5Yg riy l

t LM5.F5. - ~ --

I

- h%

3/4.4.2 SAFETY VALVES pressurized above its safety Limit of 2735 psig.The pressurizer C relief cepacity of a single safetto relieve G0,000 lbs per The hour of sa condition which could occur during shutdown.3 valve is edequate to relieve any overpressur In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides i overpressure relief capability anu will prevent RCS overpressurization In addition, the Overpressure Prote:. tion System provides a diverse means f o.

protection against RCS overpressurization at low temperatures.

Ducing operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

The combined relief capacity of all of there valves is greater than the maximum ,

i surge rate resulting from a complete loss-of-load assuming no Reactor trip dumpalso and assuming no operation of the power operated relief valves or steam valves.

Demonstration of-the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI 1

of the ASME Boiler and Pressure Code.

t' 3/4.4.3 PRESSURIZER, ,.

l l The. limit on the maximum water volume (1656 cubic feet) in the pressurizer of operation assumed in the SAR. assures that the parameter is maintained with i

assumptions. The limit is consistent with the initial SAR The 12-hour periodic surveillance is suf ficient to ensure that the The maximum water volume also ensures that a steam bubb RCS is not a hy<fraulically solid system. The requirement that a minimum i

number of pressurizer heaters be OPER/,BLE enhances the capability of the plant to control Reactor Co:,lant System pressure and establish natural circulation.

3/4.4.4 RELIEF VALVES i

The power-operated relief valves (PORVs) and steam bubble function to '

relieve RCS pressure during all design transients up to and 'ncluding the design step load decrease with steam dump. Operation of the PORVs minimizes ,

M.d Each PORV has a remotely operated block valve to providei capability should a relief valve become inoperable.

'p [ The PORVs are equipped with automatic actuation circuitry and manual 1

'0 control captbility. 1 Because no credit for PORV operation is taken in the FSAR analyses for Mode 1, 2 & 3 transients, the PORVs are considered OPERABLE in ,

either the manual or automatic mode.  ;

I It should be noted that the automatic mode is the preferred configuration as this provides pressure relieving l capability without reliance on opera, tor action. l

i. BRAIDWOOD .- UNITS 1 & 2 B 3/4 4-2 CHrNGED BY LETTE9 DATED May 8, 1989 i e l.

f

4

- INSERT D The OPERABILITY of the PURVs and block valves is determined on the basis of their being capable of performing the following functions:

l A. Manual control of PORVs to control reactor coolant system pressure.

This is a function that is used for the steam generator tube rupture accident and for plant shutdown. This function has been classified as safety related for more recent plant designs, i B. Maintainin0 the integrity of the reactor coolant pressure boundary. This ,

is a function that is related to controlling identif od leakage and ensuring 1 the ability to detect unidentified reactor coolant pressure boundary l leakage.

C. Manual control of the block valve to: (1) unblock an Isolated PORV to allow ;t to be used manual control of reactor coolant system pressure (ltem A), and (2) isolate a PORV with excessive seat leakage (item B). ,

D. Manual control of a block valve to isolate a stuck-open PORV.

Surveillance Requirements provide the assurance that the PORVs and block valves can perform their functions. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being performed to restore an inoperable PORV to operable status.

Surveillance requirement 4.4.4.1.b has been added to include testing of the mechanical anc; electrical aspects c antrol systems for air-operated PORVs.

Testing of PORVs in HOT STANDBY or HOT SHUTDOWN is rec uired in order to simulate the temperature and pressure environmental effects on PORVs. In many PORV designs, testing at COLD SHUTDOWN is not considered to be a representative test for assessing PORV performance under normal plant operating conditions.

/scl:lD615:52

REACTOR COOLANT SYSTEM

_ BASES f

PRESSURE / TEMPERATURE LIMITS (Continued)

The use of the composite curve is necessary to set conserva course of the heatup ramp the controlling cond e to the of themostoutside and critical the pressure limit must at all times be ysis criterion. based on an Finally rate data are, adjusted for possible errors in the pres sensing instruments by the values indicated on the respective curves .

which there is reason for concern of nonductile fa are provided performed to assure in accordence withcompatibility the ASME Codeofrequirements operation with the fatigue an

,g og,,pog /u/

The OPERA 8ILITY of two PORVs, or two RHR suction # valves, opening of at least 2 square inches ensures that the RCS will be p pressure transients which could exceed the limits of Appendix G to 10 'A C

Either PORV has adequate surization when the transient is limited to either:

relieving capability es-

'5 w

flCP with tha secondary water temperature of the steamorgenerator l equal to 50'F above the RCS cold leg temperatures, or (2) the start of (

y a centrifugal charging pump and its injection into a water solid RCS, >

n These two scenarios are analyzed to determine the resulting overs L assuming closed to ft:11 a single PORV actuation with a stroke time of 2.0 seconds open. F Figure 3.4-4 is based upon this analysis and represents .

• i the maxiaus allowable noted PORY variable setpoint such that,erpres- for the two ov surization transients t 10 effective full power year,s (he resulting pressure will not exceed the nominal EFPY) Appendix G reactor vessel NOT limits.-

((>

W p

"A" RHR train wide RCS range suction pressure isolation transmitter valves and 8701A valves 8701Band an 8702A a and 8702B locked with a "B" train wide range pressure transmitter. n er-Removing power from both RHR flow paths.both RHR suction relief valv y for 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME , 2, Code Class of Iffe of these the plant.components will be maintained at an acc These programs are in accordence with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda 10 CFR 50.55a Commission g except pur(su) ant towhere 10 CFR specific written relief has 50.55a(g)(6)(i). e been gra BRAIDWOOD - UNITS 1 & 2 B 3/4 4-16

REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION both 3.4.4 .AW power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: H0 DES 1, 2, and 3.

ACTION:

alk par m.cJ6cd h k, . Lini vdk-(9

a. With one or more PORV(s) inoperable because of excessive seat leakage, / '

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> eithe restore the PORV(s) to OPERABLE status or close /

the associated blor4 valve (s)folherwise be in a Fleast HOT'5TAHO W ~ ,

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and iny 40t1T gMM wM W MMng y M. l

b. With one PORY inoperable due to causes other than excessive seat leakage, within I hour either restore the PORV to OPERABLE status l or close the associated block valve and remove power from the block '

valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 40tD'NOT SHUTDOWN within the following .Wg hours. 4 leul one.

c. With both PORVis$ inoperable due to causes other than excessive seat

,b leakage, within I hour either restore cach ef-the'PORV(1ry to OPERABLE status or close%hete associated block valveM and remove power from the block valve 4) and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and incogHUTDOWNwithinthefollowingJghours.

d. -With-one'-o r-mo re-b loc k-v a l ve( S bi nope aabl e Wi th i n-1-hou re lbre ttore-ths41oc k-va l ve( s )- to-OPERABLE-s ta tus -o r- c l o s e - the- b l oc k-Rep 6 4 + valve (o)-and-remove-power-from-the-block-valve (s)cor-close-the-PORV-1,2,4 A -

( and-remove-power-from-(ts-associated-solenoid-valvet-and4)-apply-the-AGT40N-of-broFerabover-as-appropriate-for-the-isolated-PORV(+)r

e. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by: ,

g ga. Performance of a CHANNEL CALIBRATIONh-and eMe

• 'O*^ " I, l c% Operating the valve throagh one complete cycle of full travelf ddig MObES 3 or 4.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless I the block valve is closed with power removed in order to meet the requirements l of ACTION b, or c. of Specification 3.4.4.

,n -  ?

i -#Th: :pt eff4 ed-18-month-4 n t e rv el-may-be-ex tended-to-32-mon t h s-f o r-cyc4e-1-onlyr l

BYRON - UNITS 1 & 2 3/4 4-12 AmendmentNo.)[

l IllSERL A

d. With one or more block valves inoperable, within one hour restore t he block valve (s) to Ol'EltAHLE status or place its associated l'ORV in manual control. Restote at least one block valve to OPERAULE status within the next hour if both block valves are luoperable; restore any remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise be in at least 110T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 110T SilUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

IllSTET_D

b. Operating solenoid alt control and check valves on associated air accumulators in the l'ORV control system through one complete cycle of full travel, and (3147z/0063z/031391)

i REACTOR COOLANT SYSTEM i .

OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION )

3.4.9.3 At least one of the following Overpressure Protection Systems shal e OPERABLE:

a. Two residual heat removal (RHR) suction relief valves each th a Setpoint of 450 psig i 1%, or b,

/

power-operated relief valves (PORVs) with lift 5 tpoints that va with RCS temperature which do not exceed the) mit established in Figure 3.4-4a for Unit 1 (Figure 3.4-4b for it 2), or l

c. The Re greater th CoolantSystem(RCS)depressu/ri2dwithanRCSventof equal to 2 square inches APPLICABILITY: H0 DES 4 an 5, and MODE 6 with,A reactor vessel head on.

MTION:

a. With one PORV and one R s tion relief valve inoperable, either restore two PORVs or two,RR suction rellef valves to OPERABLE status within 7 days or depre)5uriz nd vent the RCS through at least a '

2 square inch vent t'hin the n hours.

b. With both PORVs,and both RHR suction gelief valves inoperable, depressurizejfid vent the RCS through at least a 2 square inch vent within 8 houts,

c. In the vent the PORVs, or the RH1 suction r y ef valves, or the RCS vent R3 p6rt{ shall

) arebeused to mitigate prepared an RCS and submitted to pressure traq(sient, the Cor ssion pursuant atoSpecial Specification 6.9.2 within 33 days. The report shal describe the circumstances initiating the transient, the effect of e PORVs, or the RHR suction relief valves, or RCS vent (s) on the tr ient, and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

l BYRON - UNITS 1 & 2 3/4 4-39 AMENbMENT NO. 37

/

n

I J

. AlfEERK_C  ;

REACTDR COOLANT SYSTEM OVERPRESSLIRE PROTECTION SYSTEM ,

LIMITING CONDITION FOP OPERATION )

3.4.9.3 At least two overpressure protection devices shall be OPERABLE, and each device shall be either I

a. A residual heat removal (RIIR) suction relief valve with a lift setting of less than or equal to 450 psig, or

b. A power operated relief valve (PORV) with a lift setpolnt that varies ,

with RCS temperature which does not exceed the limit established in i Figure 3.4-4.-

APPLICABILITY: MODES 4, 5, and 6 when the head is on the reactor vessel.

ACTION:

a. With one of the two required overpressure protection devices Inoperable in Mode 4, restore two overpressure protection devices to

= OPERABLE status within 7 days or depressurize and vent the RCS through at least a 2 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. With one of the two required overpressure protection devices inoperable in MODES 5 OR 6, restore two overpressure protection devices to OPERABLE etatus within 24-hours or vent the RCS through at least a 2 square luch vent within.the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c. With both of the required overpressure protection devices inoperable, depressurize and vent the-RCS through at least a 2 square inch vent witnin 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

d. With the RCS vented per ACTIONS a, b, or c, verify the vent pathway at least once per-31 days when the pathway is provided by a valve (s) that is locked, sealed, or otherwise secured in the open position; otherwise, verify the vent pathway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e. In the event either the PORVs, RilR suction relief valves, or the RCS vents are used to mitigate an RCS pressure-transient, a Special Report shall-be prepared and submitted to the Commission pursuant to .

l Specification 6.9.2 within 30 days. Ths report shall describe the circumstances initiating t he transient, the effect of the PORVs, RHR suction relief valves, or RCS vents on the transient, and any  ;

corrective action necessary to prevent recurrence,

f. The provisions of Specification 3.0.4 are not applicable.

(3147z/0063z/031391)

l

, , RfACTOR COOLANT $YSTEN

$URytILLANCE REQUIREMENTS g, & PORVs see. La.4 med & eoW meyrus,re. peduke.,x 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

Perfomance of an ANALOG CHANNEL OPERATIONAL TEST on the PORY  !

actuation channel but excluding valve operation, witth 314::

-emf at least e onc; per $1 days thr=ftr- When the PO OPERABLE: and b.

Perfomance at least once of pera la CHANNEL months; and CALIBRATION on the PORY actuation channel

c. Verifying"the PORY isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ch= th "! h Mh; :=d fer =ren:== ;=txthr 4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when t thefollows:

as RHR suction relief valves are being used for cold overpressure protection i

n.

For RHR suction relief valve RH07003 verif that valves RH8702A and RH87028 are open y at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

b. For RNR suction relief valve RH8708A verif that valves RH8701A and RH87018 are open. y at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

c. Testing pursuant to Specification 4.0.5.

4.4.0.3.3 The 200 ser,t(s hell M serified te M : r. ;; h n t : =; ; r 12Mw.e'-betMvent(e,};sMir,,-e.4fe,eeri...ee.rei,re%;ti.a.

i 6

_ _aA .

I Owb. h v wh ' . hiwN w wig b Yu vh ws: $st w wN etww g AA . 3 35_f5'5N!*$_5'i."i5$.!

N'" 'Ho N"NI"'~9 k''~' '"i 5 t"" V'I '#" "F'~' 2 7., . . , . .

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

~

REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at 1 v temperatures.

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum, surge rate resulting from t complete loss-of-load assuming no Reactor trip -

and also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3/4.4.3 PRESSURIZER The' limit on the maximum water volume (1656 cubic feet) in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR.

assumptions. The limit is consistent with the initial SAR The 12-hour perindic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation.

The maximum water volume also ensures that a steam bubble is-formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.

3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves.

Each PORV has a remotely operated block valve to provide a positive shutoff All capability should a relief valve become inoperable.

pd 8 The PORVs are equipped with automatic actuation circuitry and manual p control capability. Because no credit for PORV operation is taken in the FSAR analyses for Modes 1, 2 and 3 transients, the PORVs are considered OPERABLE in either the manual or automatic mode. It should be noted that the automatic mode is the preferred configur6 tion, as this provides pressure relieving capability without reliance on operator action.

BYRON - UNITS 1 & 2 B 3/4 4-2 CHANGED BY LETTER DATED May 5, 1989

~

l INSERT _IL i i

The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:

A. Manual control of POSVs to control reactor coolant system pressure. This is a function that is used for the steam generator tube rupture accident and for plant shutdown. This function has been classified as safety related for more recent plant designs.

.B. Maintaining the integrity of the reactor coolant pressure boundary. This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentifled reactor coolant pressure boundary leakage.

i C. Manual control of the block valve to (1) unblock-an isolated PORV to allow it to be used for manual control of reactor coolant system pressure i (1 tem A), and (2) isolate a PORV with excessive seat leakage (Item B). )

I D. Manual control of a block valvo to isolate a stuck-open PORV. '

I Surveillance Requirements provide the assurance that the PORVs and block j valves can perform their functions. The block valves are exempt from the surveillance requirements to cycle the valves when they have been. closed I to comply with the ACTION requirements. This precludes the need to cycle j the valves with full system differential pressure or when maintenance is i being performed to_ restore an inoperable PORV to operable status.

Surveillance requirement 4.4.4.1.b has been added to include testing of the mechanical and electrical _ aspects of control systems for alr-operated PORVs.

• Testing of PORVs in HOT STANDBY or HOT SHUTDOWN la required in order to nimulate the temperature and pressure environmental effects on PORVs. In many PORV-designs, testing at COLD SHUTDOWN is not considered to be a representative test for assessing PORV performance under normal plant operating conditions.

(3147z/0063s/031391)

~ - --'-

lEACTOR C001. ANT. SYSTEM BASES _

PRES 5URE/ TEMPERATURE LIMITS (Continued)

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to axist such that over the course of the heatup reap 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 temperature 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 i

c performed in accordance with the ASME Code requirements.t um Mv u .m te

p. w% <t m q The OPERABILITY of two p0RVs, or two RNR suction 4 valves, or an RCS vent opening of at least i 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 PORV has adequate relieving capability to protect the RCS from ove nres-

.surization when the transient is imited to either: (1) the start of an idle RCP with the: secondary water temperature of the steam generator less than or equal to 50*F above the ACS cold leg temperatures, or (2) the start of a centrifugal charging pump and its injection into a water solid RC5.

These two scenarios are analyzed to determine the resulting overshoots assuming a single P0RV actuation with a streks time of 2.0 seconds from full .

closed to full open. Figure 3.4 4 is based upon this analysis and rnpresents '

the maximus allowable PORY variable setpoint such that for the two overpres-surization transients noted the resulting pressure wi$1 not exceed the Appendix 0 reactor vessel NbT limits (noe' nel 10 effective full power years for Unit 1 only).

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing progress for A$ME Code Class 1, 2, and 3 components ensure that-the structural integrity and operatinnal readiness ,

of these components will be maintained at an acceptable level th nughout the

-life of the plant. ~These programs are in accordance with section XI of the '

A$NE 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 ths Consission pursuant to 10 CFA'50.55a(g)(8)(1). <

BYR0N - UNITS 1 & 2 B 3/4 4-16 AMENDMEN,T No. 38

ATTACHMENT C1 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS l TECHNICAL SPECIFICATION 3/4.4.4 Commonwealth Edison has evaluated this proposed amendment and .

determined that it involves no significant hazards considerations. According to 10 CFR l 50.92(c), a pro aosed amendment to an operating license involves no significant hazards consic erations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

The basis for this determination is as follows:

LIST THE DOCUMENTS IMPLEMENTING THE PROPOSED CHANGE: GL 90-06, OSR 91-025.

UNIT: 1&2 APPLICABLE MODES: 1,2,3 OTHER RELEVANT PLANT CONDITIONS: NOME SYSTEMS AFFECTED: RY,RC EQUIPMENT NAME(S): 1/2RY455A,1/2RY456 LIST THE APPLICABLE SAR SECTIONS WHICH DESCRIBE THE AFFECTED SYSTEM, STRUCTURE, OR COMPONENT (SSC) OR ACTI%TY.

INCLUDE THE ACCIDENT ANALYSIS SECTIONS. LIST ANY OTHER CONTROLLING OR REFERENCE DOCUMENTS.

UFSAR 5.2, 5.4, 7.2, 7.3, 7.6,15.0,15.2,15.3,15.6, E.23, E.24, E.49, E.50, E.56; NUREG 0800 SECTIONS 5.2.2,7.6; NUREG 0730 ITEMS II.D.1, ll.K.2:

10CFR50 APP A GDC 15,31; APP G; BY/BW SGTR ANALYSIS; NUREG 1316 " TECHNICAL FINDINGS AND REGULATORY ANALYSIS RELATED TO GENERIC ISSUE 70".

DESCRIBE HOW THE CHANGE WILL AFFECT PLANT OPERATION.

CONSIDER ALL APPLICABLE MODES. INCLUDE A DISCUSSION OF ANY SYSTEM INTERACTIONS WHICH MAY BE AFFECTED. CONSIDERATION SHOULD BE GIVEN TO THE FOLLOWING AREAS: MECHANICAL, ELECTRICAL, I&C, EO, STRUCTURAL, FIRE PROTECTION, SITE / ENVIRONMENTAL IMPACTS, RADIOLOGICAL CONCERNS, SECURITY, AND FLOODING.

/scl:lD615:28

t The proposed change will have little effect on plant operation. There are no changes in the present system interactions. The change will terminate cooldown requirements at Mode 4, which is consistent with the Modes 13 applicability of Specification 3/4.4.4. This will avoid a potential conflict with S 3ecification 3/4.4.9.3, which will allow credit to be taken for either two RH suction rol et valves or one PORV and one suction relief valve to comply with the LCO.

Additionally, the cycling of the PORV will be limited to Modes 3 and 4. This change is being made in order to reduce the potential for a PORV stickin0 open during aower operation. This is consistent with ASME Sectiori XI, Paragraph IWV 3412 (ref MUREG 1316).

Additional surveillances address the PORV actuation circuitry and the PORV control air system. These surveillances are currently being performed, so their incorporation into the Technical Specifications will have no physical effect on the plant.

DESCRIBE HOW THE CHANGE WILL AFFECT REACTIVITY MANAGEMENT:

There is no impact on Reactivity Management.

DESCRIBE HOW THE CHANGE WILL AFFECT EOUIPMENT FAILURES.

INCLUDE ANY NEW FAILURE MODES AND THEIR IMPACT DURING ALL APPLICABLE OPERATING MODES.

The proposed changes will eliminate the cycling of the PORV for testing puraoses whi e the unit is at power. This is expected to reduce the overall failure rate of t1e PORV over the life of the plant because of being performed, so their incorporation into Technical Specification 3/4.4.4 will have no impact on the plant.

Because the testing methodology for the testing of the PORVs and block valves remains unchanged, there are no new failure modes introduced.

IDENTIFY EACH AFFECTED ACCIDENT OR TRANSIENT:

Steam Generator Tube Rupture coincident with Loss of Offsite Power.

Loss of Nonemergency AC Power to Plant Auxiliaries.

Loss of Normal Feedwater Flow.

Turbine Trip - Maximum and Minimum Reactivity Feedback Cases.

LIST EACH AFFECTED TECHNICAL SPECIFICATION SECTION: 3/4.4.4 DETERMINE IF THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAR MAY BE INCREASED. ANSWER THESE QUESTIONS SEPARATELY FOR EACH AFFECTED ACCIDENT.

AFFECTED ACCIDENT- SGTR SAR SECTION: 15.6.3

/scl:ID615:29

MAY THE PHOBABILITY OF THE ACCIDENT BE INCREASED?

No. Tha subj:ct ch ngs only eff: cts the Pressuriz r PORVs, block valves, and the PORV actuation circuitry. There exists no physicalinteraction between the subject equipment and the factors which contribute to S/G tube degradation which would increase the likelihood of catastrophic tube failure. The proposed changes are intended to increase the reliability and availability of the PORVs, and are expected to result in an overallincrease in the protection of the public health and safety (ref NUREG 1316).

MAY THE CONSEQUENCES OF THE ACCIDENT (OFFSITE DOSE) BE INCREASED?

I No. The proposed changes will not increase the unavailability of the PORVs to be utilized in the mitigation of the SGTR. The PORVs are used for depressurizing the RCS to at or below the ruptured S/G pressure. By allowing the PORV to be placed under manual control as opposed to doenergizing it for an inoperable block valve, the availability of the PORV ls increased. Because the PORVs' availability is enhanced, the resultant dose to the public in the event of a SGTR is not adversely impacted. j MAY THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY INCREASE 7 No. The reduced testing of the PORVs is expected to reduce the probability of a PORV sticking open at power (NUREG 1316). Because the additional surveillances are currently being periormed, their incorporation into the Technical Specifications will have no effect on the probability of a malfunction of equipment important to safety.

MAY THE CONSEQUENCES OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY INCREASE 7 No. No change in testing methodology is being proposed, and the equipment is not being operated in a new or different manner, in the event that a block valve becomes inoperable, the associated PORV is placed in manual control, vice deenergized closed. Under these conditions, if the PORV fails open (unlikely for a valve whose failure position is closed), the resulting transient is bounded by the inadvertent opening of a safety valve. The consequences of this event have been analyzed anc found acceptable (FSAR 15.6.1).

BASED ON THE INFORMATION PROVIDED ABOVE, DOES THE CHANGE ADVERSELY IMPACT SYSTEMS OR FUNCTIONS SO AS TO CREATE THE POSSIBILITY OF AN ACCIDENT OR MALFUNCTION OF A TYPE DIFFERENT FROM THOSE EVALUATED IN THE SAR7 No. The proposed change does not introduce new equipment, and installed equipment is not operated la a new or different manner, The testing methodology is unchanged, and is consistent with the equipment design. The elimination of cycling the PORVs at power will not introduce any new or different failure mechanisms.

DETERMINE IF THE MARGIN OF SAFETY IS REDUCED (l.E. THE NEW VALUES EXCEED THE ACCEPTANCE LIMITS).

Because the proposed changes are designed to enhance the overall availability and reliability of the PORVs, no adverse impact on the margin of safety is evident.

/scl:lD615:30

O e

AFFECTED ACCIDENT: Loss of Nonemergency Power to SAR SECTION:

15.2.6 the Plant Auxiliaries.

MAY THE PROBABILITY OF THE ACCIDENT BE INCREASED?

No. The subject change only affects the Pressurizer PORVs, block valves, and the PORV actuation circultry. There exists no physical interaction between the subject equipment and the f actors which contribute to the loss of the offsite power distribution network. The proposed changes are Intended to increase the reliability and availability of the PORVs, and are ex 3ected to increase the overall protection of the public health and safety (NUREG 13' 6). The likely causes of this event are the loss of offsite power coincident with a turbine trip or loss of onsite distribution system. The PORVs are air operated and are being provided with Class 1E power, so their failure will not be felt on the distribution system, and the block valves are provided with sufficient current intenupting features so as to ensure that a fault associated with the block valve will not have a detrimental effect on either the offsite or onsite distribution network.

MAY THE CONSEQUENCES OF THE ACCIDENT (OFFSITE DOSE) BE INCREASED?

No. The proposed charsges will not increase the unavailability of the PORVs to be utilized in the mitigation of a loss of power event. The PORVs will function to limit RCS pressure to a value below the lift s9tting of the safety valves (Figure 15.2 9).

Because the changes are designed to enhance the reliability and availability of the PORVs, there will be no adverse impact on offsite doses.

MAY THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY INCREASE?

No. The reduced testing of the PORVs is expected to reduce the probability

. Because the additional of surveillances are currently being performed, their 1310) incorporation into the Technical a PORV sticking open at power (NUREG Specifications will have no effect on the probability of a maifunction of equi 3 ment important to safety. The testing which is currently being performed is cons stent with the design and installation of the affected equipment.

MAY THE CONSEQUENCES OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY INCREASE?

No. No change in testing methodology is being proposed, and the equipment is not being operated in a new or different manner. In the event that a block valve becomes inoperable, the associated PORV is placed in manual control, vice deenergized closed. Under these conditions, if the PORV f alls open (unlikely for a valve whose f ailure position is closed), the resulting transient is bounded by the inadvertent oaeninpot a safety valve. The consequences of this event have been analyzed and founc acceptable (UFSAR 15.6.1).

/sct:lD615:31

1 BASED ON THE INFORMATION PROVIDED ABOVE, DOES THE CHANGE ADVERSELY IMPACT SYSTEMS OR FUNCTIONS SO AS TO CREATE THE POSSIBILITY OF AN ACCIDENT OR MALFUNCTION OF A TYPE DIFFERENT FROM THOSE EVALUATED IN THE SAR?

No. The proposed change does not introrbce new equipment, and installed equipment is not op6 rated in a new or different manner. The testing methodology is unchanged, and is consistent with the equipment design. The elimination of cycling the PORVs at power will not introduce any new or different f ailure mechanisms.

DETERMINE IF THE MARGIN OF SAFETY IS REDUCED (l.E. THE NEW VALUES EXCEED THE ACCEPTANCE LIMITS),

Because the proposed changes are designed to enhance the overall availability and reliability of the PORVs, no adverso impact on the margin of safety is evident.

AFFECTED ACCIDENT: Loss of Normal Feedwater Flow SAR SECTION:

15.2.7 MAY THE PROBABILITY OF THE ACCIDENT BE INCREASED?

No. The subject change only affects the Pressurizer PORVs, block valves, and the PORV actuation circuitry. There exists no physical interaction between the subject equipment and the f actors which contribute to the loss of normal feedwater flow. The proposed changns are intended to increase the reliability and availability of the PORVs, and are expected to increase the overall protection of the public health and safety (NUREG 1316). The likely causes of this event are the loss of offsite power, pump f ailures, or valve malfunctions. There exist no physical or systemic interactions which could adversely affect the reliability or operation of the feedwater pum as and valves, and the reliability of the offsite power network is completely indepenc ent of the components addressed by this change.

MAY THE CONSEQUENCES OF THE ACCIDENT (OFFSITE DOSE) BE INCREASED?

No. The propored changes are designed to increase the availability of the PORVs for the mitigation of a loss of feedwater event. The PORVs will function to limit RCS pressure to a value below the lift setting of the safety valves (Figure 15.211).

Because the changes are designed to enhance the reliability and availability of the PORVs, there will be no adverse impact on offsite dosos. The limiting failure of these scenarios would be the loss of a single train of auxiliary feedwater. The oroposed changes will not allow continued operation of the plant with the PORVs c egraded beyond the level that is currently acceptable.

MAY THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY INCREASE?

No. The reduced testing of the PORVs is expected to reduce the probability of a PORV sticking open at power (NUREG 1316). Because the additional surveillances are currently being performed, their incorporation into the Technical Specifications will have no effect on the probability of a malfunction of equipment important to safety.

/scl:lD615:32

1 MAY THE CONSEQUENCES OF A MALFUNCTION OF EQUIPMENT i IMPORTANT TO SAFETY INCREASE? l No. No chan e in testing methodology is bein proposed, and the equipment is not being operatedfn a new or different manner. In tke event that a block valvo l becomes inoperable, the associated PORV is placed in manual control, vice l deenergized closed. Under these conditions, if the PORV f ails open (unlikely for a <

valve whose failure position is closed), the resulting transient is bounded by the inadvertent opening of a safety valve. The consequences of th6 event have been analyzed anc found acceptable (FSAR 15.6.1).

BASED ON THE INFORMATION PROVIDED ABOVE, DOES THE CHANGE ADVERSELY IMPACT SYSTEMS OR FUNCTIONS SO AS TO CREATE THE POSSIBILITY OF AN ACCIDENT OR MALFUNCTION OF A TYPE DIFFERENT FROM THOSE EVALUATED IN THE SAR?

No. The proposed change does not introduce new equipment, and installed equipment is not operated in a new or different manner. The testin0 methodology is unchanged, and is consistent with the equipment design. The elimination of cyc ing the PORVs at power will not introduce any new or different failure mechanisms.

DETERMINE IF THE MARGIN OF SAFETY IS REDUCED (l.E. THE NEW VALUES EXCEED THE ACCEPTANCE LIMITS).

Because the proposed changes are designed to enhance the overall availability and reliability of the PORVs, no adverse impact on the margin of safety is evident.

AFFECTED ACCIDENT: Turbine Trip (Both Cases) SAR SECTION: 15.2.3 MAY THE PROBABILITY OF THE ACCIDENT BE INCREASED?

No. The subject change only affects the Pressurizer PORVs, block valves, and the PORV actuation circuitry. There exists no physical Interaction between the subject equipment and the tactors which contribute to a turbine trip with either maximum or minimum reactivity feedback conditions. The likely causes of this ment do not involve components or parameters germaine to the PORVs. There exist no physical or systemic interactions which could adversely affect the reliability or operation of the main turbine, the generator or their associated controls.

MAY THE CONSEQUENCES OF THE ACCIDENT (OFFSITE DOSE) BE INCREASED?

No. The proposed changes are designed to increase the availability of the PORVs for the mitigation of a turbine trip. This event has been analyzed assuming both operable and noperable PORVs under both maximum and minimum reactivity feedback conditions. The offsite dose consequences for all four cases are minimal (FSAR 15.2.3.3), so it can be concluded that no increase in the consequences of an accident could result.

MAY THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY INCREASE?

No. The reduced testing of the PORVs is expected to reduce the probability of a PORV sticking open at power (NUREG 1316). Because the additional surveillances are currently being performed, their incorporation into the Technical Specifications will have no effect on the probability of a malfunction of equipment important to safety.

/scl:lD615:33

MAY THE CONSEQUENCES OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY INCREASE? '

No. No change in testing methodology is being proposed, and the equipment is not being operated in a new or different manner. In the event that a block valve l becomes inoperable, the associated PORV is placed in manual control, vice deenergized closed. Under these conditions,if the PORV fails open (unlikely for a calve whose failure position is closed), the resulting transient is bounded by the inadvertent o:>ening of a safety valve. The consequences of this event have been analyzed and found acceptable (FSAR 15.6.1).

BASED ON THE INFORMATION PROVIDED ABOVE, DOES THE CHANGE  !

ADVERSELY IMPACT SYSTEMS OR FUNCTIONS SO AS TO CREATE THE POSSIBILITY OF AN ACCIDENT OR MALFUNCTION OF A TYPE DIFFERENT FROM THOSE EVALUATED IN THE SAR?

l No. The proposed change does not introduce new equipment, and installed l

equipment is not operated in a new or different manner. The testing methodology is unchanged, and is consistent with the equipment design. The elimination of cycling the PORVs at power will not introduce any new or different failure mechanisms.

DETERMINE IF THE MARGIN OF SAFETY IS REDUCED (1.E. THE NEW VALUES EXCEED THE ACCEPTANCE LIMITS).

Because the proposed changes are designed to enhance the overall availability and reliability of the PORVc, no adverse impact on the margin of safety is evident.

/scl:lD615:34

e ATTACHMENT C2 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS TECHNICAL SPECIFICATION 3/4.4.9.3  ;

i-Commonwealth Edicon has evaluated this proposed mendment and determined that it involves no significant hazards considerations. According to 10 CFR 50.92(c), a pro aosed amendment to an operating license involves no significant hazards consic era!!ons if operation of the facility in accordance with the proposed i amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any .

accident previously evaluated; or

3. Involve a significant reduction in a margin of safety. -

The basis for this determination is as follows:

LIST THE DOCUMENTS IMPLEMENTING THE PROPOSED CHANGE: NRC i GL 90 06, OSR 91-025. i UNIT: 1&2 APPLICABLE MODES: 4,5,6 WITH HEAD ON VESSEL i OTHER RELEVANT PLANT CONDITIONS: NONE SYSTEMS AFFECTED: RC,RY,RH EQUIPMENT NAME(S): 1/2RY455A,1/2RY456 i

DESCRIBE THE PROPOSED CHANGE AND THE REASON FOR THE CHANGE:

The proposed change is belag submitted in accordance with NRC Generic ';

Letter 90 06, which is intended to resolve Generic issues 70 and 94. Generic Issue 94,

" Additional Low Temperature Overpressure Protection for Light Water Reactors" addresses concerns with the implementation of the resolution to Untosolved Safety issue A 26, " Reactor Vessel Pressure Transient Protection (Overpressure Protection)".

The resolution of USl A 26 involved several administrative actions to reduce the potential for overpressure transients. These include, but are not limited to minimizing the time the RCS is water-solid, restricting the number of high pressure pumps operable below 330 Degrees F, minimizing the RCS to S/G temperature differential, '

and providing a PORV setpoint which varies with RCS pressure and temperature.

Cespite these controls, low temperature overpressure transients have continued to o: cur, The net effect of NUREG 1326, which provides the resolution to Generic issue 94,is to limit the amount of time that a plant can be in a condition conducive to an overpressure transient with only a single overpressure mitigation device operable to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The Technical-Specification revisions being submitted to address this portion of Generic Letter 90-06 accomplish the following:

1) The LCO is rewritteq to require any two overpressure mitigation devices when the RCS is nct depressurized through a 2 square inch or larger .

vent. This is a slight modification of the current LCO, which requires either two PORVs in the Armed LOW Temperature mode or Iwo RH suction reliefs to be operable.

0/ scl:lD615:35

o o

2) With only one overpressure mitigation device operable in Mode 4, the current 7 day allowed outage time is retained. With only one overpressure mitigation dovice operable in Mode 5 or u ode 6 with the vessel head on, the allowed outage time is reduced to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3) Surveillance Raquirement 4.4.9.3 is rewritten as an Action Statement because of lts conditional nature.

LIST THE APPLICABLE SAR SECTIONS WHICH DESCRIBE THE AFFECTED SYSTEM, STRUCTURE, OR COMPONENT (SSC) OR ACTIVITY. INCLUDE THE ACCIDENT ANALYSIS SECTIONS. LIST ANY OTHER CONTROLLING OR REFERENCE DOCUMENTS.

FS AR 3.9, 5.2, 5.4, 7.2, 7.6; NUREG 0800 SECTION 5.2.2; NUREG 1326 " Additional Low-Temperature Overpressure Protection for Light Water Reactors";

IEEE 279-1971,3381971.

DESCRIBE HOW THE CHANGE WILL AFFECT PLANT OPERATION.

CONSIDER ALL APPLICABLE MODES. INCLUDE A DISCUSSION OF ANY SYSTEM INTERACTIONS WHICH MAY BE AFFECTED. CONSIDERATION SHOULD BE GIVEN TO THE FOLLOWING AREAS: MECHANICAL, ELECTRICAL, l&C, EO, STRUCTURAL, FIRE PROTECTION, SITE / ENVIRONMENTAL IMPACTS, RADIOLOGICAL CONCERNS, SECURITY, AND FLOODING.

The proposed revision willlimit the amount of time a plant is permitted to operate in a configuration where an overpressure transient poses a concern with only one overpressure mitigation device operable. In Mode 5 and Mode 6 with the head resting on the reactor vessel, this allowed outage time is limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This change will also allow credit to be taken for any two c aerable overpressure mitigation devices to comply with the provisions of the LCO. Th s change wi:1 not impact any of the system interactions in the areas described above.

DESCRIBE HOW THE CHANGE WILL AFFECT REACTIVITY MANAGEMENT:

This change has no impact on reactivity management.

DESCRIBE HOW THE CHANGE WILL AFFECT EQUIPMENT FAILURES.

INCLUDE ANY NEW FAILURE MODES AND THEIR IMPACT DURING A' APPLICABLE OPERATING MODES.

The proposed change does not alter the manner in which equipment is maintained or o aerated. There is no reduction in the level of surveillance conducted on i equipment usec to comply with the LCO, Because the equipment is not being operated in a new or different manner, and no new equipment is being introduced, no new failure modes are being introduced.

IDENTIFY EACH AFFECTED ACCIDENT OR TRANSIENT:

The installed overpressure mitigation devices are designed to prevent overpressurization upon the inadvertent start of either a centrifugal charging pump or a reactor coolant pump.

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LIST EACH AFFECTED TECHNICAL SPECIFICATION SECTION: 3/4.4.9.3 DETERMINE IF THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAR MAY BE INCREASED. ANSWER THESE QUESTIONS SEPARATELY FOR EACH AFFECTED ACCIDENT.

AFFECTED ACCIDENT: RCS Cold Overpressurization SAR SECTION:

5.2.2, 5.4, 7.6.9 MAY THE PROBABILITY OF THE ACCIDENT BE INCREASED?

No. There is no change being proposed in the controls designed to limit the occurrence of an overpressure transient. These include, but are not limited to minimizing the time the RCS is water solid, restricting the number of high pressure pumps operable below 330 Degrees F, minimizing the RCS to S/G temperature differential, and providing a PORV setpoint which varies with RCS pressure and temperature. The changes only serve to limit the a lount of time a plant is vulnerable to a potentially damaging overpressure transient whh Fmited overpressure protection available.

MAY THE CONSEQUENCES OF THE ACCIDENT (OFFSITE DOSE) BE INCREASED?

There is no change being proposed which would impact any potential offsite dose attributable to an overpressur zation event. The proposed revisions do not alter the source term, the containment isolation or closure capability requirements, or allowable release levels.

MAY THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY INCREASE 7 No. There is no change being proposed to the level of surveillance required to demonstrate compilance with the LCO. The installed cverpressure mitigation devices will continue to be operated and tested in a manner consistent with their design and Installation. The proposed changes are Intended to enhance the level of overpressure protection during periods of vulnerability.

MAY THE CONSEOUENCES OF A MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY INCREASE?

No. The proposed change does not impact the f actors which determine the potential offsite dose in the event of a failure of the RCS due to a cold overpressurization event, b/ limitin the amount of time that a olant may be in a vulnerable condition with only a sin e mitigation device availab e, the overall availabi!!ty of overpressure protecti n features is enhanced. This will enhance the ability of the plant to withstand an overpressure transient concurrent with the f ailure of an operable overpressure rnitigation device.

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BASED ON THE INFORMATION PROVIDED ABOVE, DOES THE CHANGE ADVERSELY IMPACT SYSTEMS OR FUNCTIONS SO AS TO CREATE THE POSSIBILITY ff AN ACCIDENT OR MALFUNCTION OF A TYPE DIFFERENT Ff:l?M THOSE EVALUATED IN THE SAR?

No. The proposed change does not introduce new equipment, and installed equipment is not operated in a new or different manner. The testing methodology is unchanged, and is consistent with the equipment design. The level of surveillance required to demonstrate compliance with the LCO remains unchanged.

DETERMINE IF THE MARGIN OF SAFETY IS REDUCED (1.E. THE NEW VALUES EXCEED THE ACCEPTANCE LIMITS).

There is no reduction in the margin of safety. No change is being proposed to the setaolnts of the RH suction relief valves or the methodology and parameters used to estaalish the PORV cold overpressure protection setpoints. Additionally, by limiting the allowed outage time in Modes 5 and 6, the overall availability of overpressure protection is enhanced. As such, the margin of safety is not reduced.

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.t ATTACHMENT D ENVIRONMENTAL ASSESSMENT STATEMENT Braidwood Station has evaluated the proposed amendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has been determined that the proposed change meets the criteria for a categorical exclusion as provided for under 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50, the change affects the requirements pertaining to the use of components located within the restricted area as defined in Part 20 of this chapter, and the change involves no significant hazards considerations. There is no change in the amount or type of releases made offsite, and there is no significant increase in individual or cumulative occupational radiation exposure.

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