ML20065T438

Proposed Tech Specs Re Heatup & Cooldown Curves & PORV Low Temp Overpressure Protection Setpoints

ML20065T438
Person / Time
Site:	Braidwood
Issue date:	12/19/1990
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20065T434	List: ML20065T432 ML20065T438 ML20065T446
References
NUDOCS 9012280123
Download: ML20065T438 (23)

Text

l ATTACHMEN1 B Proposed changes to Technical Specifications of facility Operating License NPF 72 and 77:

1. Braidwood Station Onsite Review

2. BwAP 1205-3T3, Request for Offsite Review

3. Revised pages:

Index VIII 3/4 4-32 3/4 4-33 3/4 4-34 3/4 4-35 3/4 4-36 3/4 4-39 3/4 4-40a 3/4 4-40b 3/4 4-41 B 3/4 4-7 B 3/4 4-8 0 3/4 4-11 B 3/4 4-12 0 3/4 4-15 8 3/4 4-16 9012280123 901219 PDR ADOCK 05000436 P PDR tt1LD615-6

l LIMITING CONDITIONS FOR OPERATION AND SL'RVEILLANCE REQUIREMENTS SECTION PAGE I 1

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......

3/4 4-23 3/4.4.7 CHEMISTRY................................................ 3/4 4-24 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CNEMISTRY SURVEILLANCE

' REQUIREMENTS........................................ 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

>l pCi/ GRAM DOSE EQUIVALENT I-131.................. 3/4 4-29 TABLE 4.4-4 REACTOR CON _ ANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................... 3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-32 FIGURE 3.4-2a REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)...... 3/4 4-33 FIGURE 3.4-2b REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... 3/4 4-34

, FIGURE 3.4-3a REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)...... 3/4 4-35 FIGURE 3.4-3b REACTOR COOLANT SYSTEM C00.J0WN LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... 3/4 4-36 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE....................... 3/4 4-37 Pressurizer.............................................. 3/4 4-38 Overpressure Protection Systems..........................

3/4 4-39 FIGURE 3.4-4a. NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTIONSYSTEMAPPLICABLEUPTO10EFPY.h4.I. . 3/4 4-40 A 9/4.4.10 STRUCTURALINTEGRITY.....................................

3/4 4-42 N fjY 3/4.4.11 REACTOR COOLANT SYSTEM

' VENTS'............................. 3/4 4-43 3

,4-4b benhs) f05 fAtMM Q S$f VJua4 r(6S

%udau. fa,.h W<t opeym f4 dan 3lY 4*Oy sp B444 0

.BRAIDWOOD - UNITS 1 & 2- VIII

I r

j REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM s

LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2a and 3.4-3a for Unit 1 (Figures 3.4-2b and 3.4-3b for Unit 2) during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100*F in any 1-hour period,

b. A maximum cooldown of 100'F in any 1-hour period, and c.

A maximum temperature change of less snan or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldow limit curves.

At all times.

APPLICABILITY:

-ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to-determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> an, continued operation or be'in at least HOT STANDBY within the d reduce the RCS T,yg and pressure to less than 200'F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS '

'4.4.9.1.1 The~ Reactor Cor int System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.'4.9.1.2- The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine _ changes in material properties, as required by 10 CFR Part 50, Appendix H,:in accordance with the schedule in Table 4.4-5. The.results of ,hese examinations shall be used to update FiguresfC3744) and 3. 4-4cx. p U it I (Fipts 3. V t/6 p 4 2/ 2) .

o.+2 mt 3.y-u y us yqw s.q a ,,.A 3,4-sy, uh),

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i curve applicable for%stup rates up to 100*F/hg for the service period up to 32 EFMand contains s.trgins of 107 and 60 psig for 2600 possible instrument errors

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FOR THE SEWICE PERIOD UP 70 32 EFPY#d-350 0

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s REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2) 1 BRAIDWOOD - UNITS 1 & 2 3/4 4-34

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CURVES APPLICABLE FOR HEATUP RATES UP TO 100'F/HR FOR THE SERVICE PERIOD UP TO i 16 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUkiNT ERRORS.

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FIGURE 3.& tb REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFEY (UNIT 2)

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

l Curves applicable to periodinstrument possible up to 32 errors D"PT*and contains margins of 107 and

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FIGURE 3.4-3a I

REACTOR COOLANT SYSTEM C00LOOWN LIMITATIONS 8 4 d&alA. t kky~ ^du<d > APPLICABLE p UP TO 32 EFPY (UNIT 1) g

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Curves applicable for cooldown rates up to 1001/hr for the service period up to 16 D7I and ocotsins enrgins of 107 and 60 psig for s possible instrument errors i 2500 ' )

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/ REACTOR COOLANT SYSTEM COOLDOWN

/

LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)

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BRAIDWOOD - UNITS 1 & 2 3/4 4-36

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CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100*F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY. CONTAINS MARGIN OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS.

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INDICatto TEWACRATURE (CCC.r) i FIGURE 3.4-3b REACTOR COOLANT SYSTEM COOLDOW LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2) ,

l BRAIDVD0D - UNITS 3 & 2 3/4 4-36

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:

a.

Two' residual heat removal ~(RHR) suctio'n relief valves each with a Setpoint of 450 psig i 1%, or

b. Two power-operated relief valves (PORVs) with lift Setpoints that vary with RCS temperature which do not exceed the limit established d m & l(3,tl-4b h llh

c. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2 square inches.

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

ACTION:

a. With one PORV and one RHR suction relief valve inoperable, either restore two PORVs or two RHR suction relief valves 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 />.

I b. With both PORVs and both RHR suction relief valves inoperable, depressurize and vent the RCS through 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 suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the l

circumstances initiating the transient, the effect of the PORVs, or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence,

d. The provisions of Specification 3.0.4 are not applicable.

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

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FIGURE 3.4-4 A  ;

L b NOMINAL PORV PRESSURE RELIEF SETFOINT VERSU9 RCS. TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE 'UP TO 10 EFPY f g () ,

t LBRAIDWOOD - UNITS 1 & 2 3/4 4-40 m f.aWsd AMENDME.NTN0.

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FIGURE 3.4-4b NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLO OVERPRES5URE PROTECTION SYSTEM (UNIT 2)

BRAIDWOOD -~ UNITS I & 2 3/4 4-40b

Tta. -.

REACTOR C00Ly _ SYSTEM y SURVEILLANCE REQUIREMENTS 4.4.9.3.1: Each PORV shall be demonstrated OPERABLE by:

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

- i

.b. . Performance of a CHANNEL CALIBRATION on the PORV actuation channel E 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 /> when the PORV is being used for overpressure protection.

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

a. For RHR suction relief valve 8708B:

~

.- ~1) By verifying _at least once per 31 days that RHR RCS Suction 1

Isolation _ Valve RH8702A is~open with power to the valve operator removed, and

2) By verifying at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that RH87028 is open,

b. For RHR~ suction relief valve 8708A:

1) By verifying at least c.nce per 31 days that RH87018-is open with power to the valve operator removed, and '

i

2) By verifying.at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that RH8701A is open.

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

• when the_ vent (s) is being used for overpressure-protection.

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

L #T6 epeci44ed 'n month interval may .be extended _to 32-months-fw-cycae-1-only. ~

~

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

REACTOR COOLANT SYSTFM

'(

( BASES SPECIFIC ACTIVITY (Continued) take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomenon. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Sr.: tion III, Appendix G:

1. The reactor r;oolant temperature and pressure and system heatup and M oldown rates (with the exception of the pressurizer) shall be 1(mited in accordance with Figures M-4 and +.-44 for the service period spe.:ified thereon: g4 h 3.4 d)

<3A~3"(34 ~ 3 b)

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and (s.- 3A-2+ 3d-2b) g' 3.H e. (3,4-3 L )

b. Figures %(-f and Gr4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

, 2. These limit lines shall be calculated periodically using methods provided below,

3. The secondary side of the. steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below

.70'F,

4. The pressurizer heatup and cooldown rates shall not exceed 100 F/hr

! and 200 F/hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F, and

5. System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

l The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the 1973 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel and Code.

BRAIDWOOD - UNITS 1 & 2 B 3/4 4-7 I

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE' LIMITS (Continued)

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temr,erature, RT NDT, at the er.d of 32* effective full power years fo*r Unit 1 (16 effective full power years for Unit 2) of service life. The 32 EFPY for Unit 1 (16 EFPY for Unit 2) service life period is chosen such that the limiting RTNDT at the 1/4T location in the core region is greater than the RT f the limiting unitradiated material.

NDT

'p Mj '

The selection of such a limiting RTNDT assures that all components in the Rea tor Coolant System will be ooerated conservatively in accordance with O\ applicable Code requirements.

$ Ql[ tar S 3/62/-)b h NN t RT The

the reactor results ofvessel these materials have in tests are shown been Tabletested to determine B 3/4.4-la.h Reactor(their initial M opbtionandresultantfastneutron(Egreaterthan1MeV)irradiationcan M

cause an increase in the RTNDT. Therefore, usted reference temperature, M i based upon the fluence copper content and 7....

El in question, can be prudicted using Figure B 3/4.4-1 res content of the material and the largest value of l ART c

' NsNDT h :1gyp E! g e ner Regulatory Guide 1.99, Revision -4ffects-of- 2, "Redd. um, ,

=nt4-or hedicted R4444t4on-h=;p to Reactor Vessel Materials" h

Jr the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2.

a and cooldown limit curves of Figures 3.4 2 and-3,4:3 include predictedThe heatup qi #a"dJustments for this shift MTND,. at the end of 32*EFPY for Unit 1 (16 EFPY

r- for Unit 2) as well as adjustments for possible errors in the pressure and temperature sensing instruments. A l

Values of ART determined in this manner may be used until the results NDT from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen with-drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the capsule.

of the reactor vessel material by using the lead factor and the withdrawal time ART The heatup and cooldown curves must be recalculated when the NDT determined from the surveillance capsule exceeds the calculated ART HDT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975 4A Matu2l Q; p&&L'Q ar% kwe kn suacd u accek shd faf. G.d. W BRAIDWOOD - UNITS 1 & 2 u smbVtH 8 3/4 4-8 u empw.

, s .

INSERT i Revised heatup and cooldown curves have been generated for Unit 2 in accordance with Regulatory Guide 1.99 Revision 2. For Unit i the curves remain the sametj However, the applicability date has been reduced per Regulatory Guide 1.99 Revision 2 to 4.5 EFPY for heatup and 12.0 EFPY for cooldown.

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1 1

'";/4'

-C -= .)g 4

TABLE B 3/4.4-la en ..

s REACTOR VESSEL; TOUGHNESS U (UNIT 1) 6 8 Average

' Shelf Energy T RT c- MATERIAL Cu P NDT NDT NMWD" MWD"*

i'i COMPONENT Heat No. SPEC.  %  % *F F* ft-lbs ft-lbs I d Closure Head Dome D1398-1 A533B, C1. 1 .06 .559 -~30 ~30 129 - I w Closure Head Ring 49C1126-1-1 A508, C1. 3 .02 .009 -20 -20 123 -

e- Closure Head Flange 2030-V-1 A508, C1. 2 .11 .009 -20 -20 163 -

ro Vessel Flange 122N357VA1 A508, C1. 2 -

.010 -10 -10 106 -

Inlet Nozzle 21-3257 A508, C1. 2 .09 .008 -20 -20 144 -

Inlet Nozzle 21-3257 A508, C1. 2 .09 .010 -10 -10 144 -

Inlet Nozzle 22-3313 A508, C1. 2 .07 008 -10 -10 130 -

Inlet Nozzle 22-3313 A508, C1. 2 .07 .010 0 0 115 -

Outlet Nozzle 22-3025 A508, C1. 2 .13 .013 -10 -10 125 -

Outlet Nozzle 4-3329 A508, C1. 2 .08 .009 -20 -20 156 -

5 Outlet Nozzle 4-3383 A508, C1. 2 .08 .008 -20 -20 147 -

1 Outlet Nozzle 11-5226 A508, C1. 2 .09 .007 -10 -10 125 -

1 = Nozzle Shell SP7016 A508, C1. 2 .04 10 155 -

a Upper She11 H

• 49D383/ A508, C1. 3 .05 .008[.7b

.008 3010 -30 122 173 49C344-1-1 Lower Shell n1 490867/ A508, C1. 3 .03 .007[.13) -20 -20 135 151 l 49C813-1-1 l Bottom Head Ring 49D148-1-1 A508, C1. 3 .05 .008 -50 -50 147 -

Bottom Head Dome C4882-1 A5338, C1. 1 .14 -20 -20 123 -

Upper Shell to M% WF-562 .04 .010(.(7)

.015 40 40 80 -

Lower Shell Girth Weld Weld HAZ -70 <-10 151 -

!

• Normal to rafor working direction.

• • Major working direction.

4 44 Nkab8 Y $N. $ 15ba l U $t b Aer- ]~ A st- t h 10 E  %

• WM l

.~ .

$) m .;.gf Nf l ,l , f 3 '

F[

' ;L sE .

>.3

> r TABLE B 3/4.4-1b '

> REACTOR VESSEL TOUGHNESS O (UNIT 2). .

6 8 Average- .

. T RT Shelf Energy MATERIAL Cu P NOT NOT NMWD" c- COMPONENT g HEAT NO. MWO**

SPEC.  %  % *F F*

5 Closure Head Dome 49754-1 A5338, C1. 1 .16 .005 -60 ft-lbs ft-lbs d Closure Head Ring- -60 151- -

50C478-1-1 A508, C1. 3 .05 .006 -30 -30 128 -

g Closure Head Flange 2031-V-1 A508, C1. 2 -

.009 20

e. -Vessel Flange 20 135 -

124P455 A508, C1. 2 .07 .010 20 20 128 m Inlet Nozzle 41-5414 A508, C1. 2 .07 .008 -10 Inlet Nozzle -10 137 -

41-5414. A508, C1. 2 .07 .009 -10 -10 Inlet Nozzle 140 -

42-5417 A508, C1. 2 .09 .011 -10 -10 Inlet Nozzle 122 --

42-5417 A508, C1. 2 .09 .009 -10 -10 Outlet Nozzle 116 -

, 4-3502 A508, C1. 2 .09 .012 -10 -10 Outlet Nozzle '

155 -

11-5226 A5G8, C1. 2 .09 .009 -10 -10 m Outlet Nozzle 116 -

4-3481 A508, C1. 2 .07 .008 -10 -10 Outlet Nozzle 163 -

,o 11-5266 A508, C1. 2 .09 .010 10 10 3;: Nozzle Shell 117 -

SP7056 A508, C1. 2 .04 .005 30 30 115 -

a Upper Shell Fr** 490963/: A508, C1. 3 49C904-1-1

.03 .007['l -30 -30 119 147 Lower Shel1+*W 50D102/ A508, C1. 3 .06 .006hd-30 -30 144 168 50C97-1 Bottom Head Ring # 9 4701066-1 A508, C1. 3 .07 .008 -30 Bottom Head Dome -30 156 -

01429-1 A533B, C1. 1 .11 .010 -20 -20 Upper Shell h # WK. WF-562 120 -

.04 .015(.0)40 40 80 -

Lower Shell '" " i Weld Weld HAZ

-30 -30 145 -

e

• Normal to major working direction.

• • Major working direction.

%%Y N W .( $ ali99$MS E DW *" N#

- -- ~ --

.~

-4

( REACTOR COOLANT SYSTEM BASES PRES $URE/ TEMPERATURE LIMITS (Continued) M -34 NTI hkr -3bhs M h N A he notch in the cooldown curve of Figure 3.4-3 6 due to the added con-straint on the vessel closure flange given in Appendix G of 10 CFR 50. This constraint requires that, at pressures greater than 20% of the preservice system hydrostatic test pressure, the flange regions that are highly stressed by the bolt preload must exceed the RT of the material by at least 120'F. The NDT flange RT required.N T,w,+ ,&k 120'F h, ngeg ,A on

) Qthe cooldown curves and therefore th

) 4 (

HEATUP h'%

• te io b NC8 t

' Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K yg for the 1/4T crack i during heatup is lower than the K yp for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K IR 's f r steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the ldwer value of the allowable pressere calculated for steady-state and finite heatup rates is obtained.

l The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients estab'lished at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analyzed on an individual basis.

i

' Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows.

A composite curve is constructed based on a point-by-point comparison of the steady-state and finito heatup rate data. At any given temperature, the allowable pressure is taken to be tne lesser of the three values taken from the curves under consideration.

i l BRAIDWOOD - UNITS 1 & 2 B 3/4 4-15

, 3rp ~3 t nt

~ REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE 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 cochlown 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 performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, or two RHR su: tion 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 PORV has adequate relieving capability to protect the RCS from overpres-(' surization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50'F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and its injection int.o a wat<er solid RCS.

r 3.Ho. ( 3 4 -%) 4tc.

These two scenarios are analyzed to determine the resulting overshoots assuming a single PORV actuation with a stroke time of 2.0 seconds from full closed to full open. Figure 3r4-4 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.aomina4~

-effeethe fur pr/aer yearsiEPPYt Appendix G reactor vessel NDT limits.

RHR RCS suction isolation valves 8701A and 8702A are interlocked with an "A" train wide range pressure transmitter and valves 8701B and 87028 are inter-locked with a "B" train wide range pressure transmitter. Removing power from valves 8701B and 8702A, prevents a single failure from inadvertently isolating both RHR suction relief valves while maintaining RHR isolation capability for both RHR flow paths.

314.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2,

.and 3 componer.ts 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)(1).

l BRAIDWOOD - UNITS 1 & 2 8 3/4 4-16 '

ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Commonwealth Edison has evaluated this proposed amendment and determined t' t it involves no significant hazards considerations. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consideraticns if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences o' 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 proposed change does not result in a significant increase in the probability or consequences of accidents previously evaluated. Several of the changes involved are administrative in nature, and as such have no impact on the probability or consequencas of accidents. The changes to the Unit 2 heatup, cooldown and cold overpressure protection figures are the result of reanalysis performed in accordance with Regulatory Guide 1.99 Revision 2.

This revision effectively resulted in a shifting of the above curves in a more conservative direction. 1he shifting of these curves has no effect on the probability for occurrence of any accidents. The consequences for accidents would remain unchanged, the opening setpoint for cold overpressure protection will be at a lower value, thus ensuring that the Appendix G limits of 10 CFR 50 will continue to be met. The reduction in Effective Full Power Years (EFPY) for the Unit 1 vessel will ensure that the Appendix G limits will continue to be met until new curves are generated.

The proposed change does not create the possibility for a new or different kind of accident from any accident previously evaluated. The proposed change does not introduce any new equipment or change the fashion in which the installed equipment will be operated. The revised setpoints for the cold overpressure protection setpoints are still high enough to allow normal heatup and cooldown operations without requiring programmatic changes. The changes involved will place Unit 2 in compilance with the new methodology for the calculation of heatup, and cooldown curves outlined in Regulatory Guide 1.99 Revision 2. This change also addresses the impact of this reanalysis on the cold overpressurizaticn systems setpoints. The Unit I curves have had their effective dates revised to ensure compliance until new curves are generated.

The proposed :hange does not involve a significant reduction in a margin of safety. 'he margin to safety will remain unchanged. The reduced applicability date for the Unit I curve will ensure all current limitations are met up to and includir., that date. The changes made to the Unit 2 curves are in accordance with the new methodology outlined in Regulatory Golde 1.99 Revision 2. The Unit 2 curves are effectively being shifted in the more conservative direction, and as such will not reduce the margin to safety, m u4 u- 7 I

ATTACHMENT D ENVIRONMENTAL ASSESSMENT STATEMENT Braldwood 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 detumination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50, 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 cumulat)ve occupational radiation exposure l

IHLDs315-0

ATTACHMENT E ADDITIONAL INFORMATION The following was asked by the NRC with respect to the Byron submittal. This information is being provideo for the Braidwood submittal.

1. What is the vessel inside radius at beltline?

Unit 1 - 86.625" Unit 2 - 86,5"

2. What is the fluence rate for Braidwood Unit 1 0 4.5 EFPY

-(beltline)?

5 x 1018 n/cm2

3. What is the Braidwood vessels fabricator?

Babcock & Wilcox

4. What is the Braldwood I thickness t. bel tl i n-8.5"

5. When were the irradiation coupors taken from Braidwood Unit I?

During AIR 01, September, 1989.

1 L 6. What is the Braidwood 1 new adjusted RTNOT at 1/4T and '/4T for o

4.5 EFPY?

1/4T - 109,F j 3/4T - 84 F.

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f,NLD615-10

_ - _ _ _ - - - _ - - - _ _ - _ _ _ _ _ - _ - _ _ - - _ - _ . _ _ _ _ _-