Text

Proposed Tech Specs Relocating Reactor Vessel Surveillance Program Capsule Withdrawal Schedules IAW GL 91-01
	ML20141K001
Person / Time
Site:	Byron, Braidwood
Issue date:	05/21/1997
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20141J982	List: ML20141J978; ML20141K001; ML20141K018; ML20141K024; ML20141K037; ML20141K043;
References
	GL-91-01, GL-91-1, NUDOCS 9705280281
	Download: ML20141K001 (82)
	v • d • e

- --

ATTACHMENT B-1 A PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, FOR FACILITY OPERATING LICENSES NPF-37 AND NPF-66 BYRON NUCLEAR POWER STATION UNITS I AND 2 REVISED PAGES I

l-4 3/4 4-32 l

3/4 4-33 3/4 4-34 3/4 4-35 3/4 4-36 3/4 4-37 4

3/4 4-38*

3/4 4-39 3/4 4-40a 3/4 4-40b B 3/4 4-7 j

B 3/4 4-8 B 3/4 4-9 B 3/4 4-10 B 3/4 4-11 B 3/4 4-12 B 3/4 4-13 B 3/4 4-14 B 3/4 4-15 B 3/4 4-16 6-23

Page provided for continuity. No changes are proposed.

9705200281 970521 ADOCK0500g4 PDR

_. _ _ _ _ _ _ _ _ _ ~_ _ _ _ _ _ _. __ ___ _ _ _ _

1 INDEI i

DEFINITIONS

\\

SECTION PAGE 1.0 DEFINITIONS 4

1.1 ACTI0N........................................................

1-1 l

1.2 ACTUATION LOGIC TEST..........................................

1-1 1.3 ANALOG CHAlgfEL OPERATIONAL TEST...............................

1-1 1

4 j

1.4 AXIAL FLUX DIFFERENCE.........................................

1-1 1.5 CHANNEL CALIBRATION...........................................

1-1 1

1.6 CHANNEL CHECK.................................................

1-1 i

l 1.7 CONTAlletENT INTEGRITY.........................................

1-2 i.

CONTR0ttED LEAKAGE............................................

1-2 1.9 CORE ALTERATION...............................................

1-2 1.9.a CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS................................................

1-2 1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................

1-2 1.11 DOSE EQUIVALENT I-131........................................

1-2a 1.12 E-AVERAGE DISINTEGRATION ENERGY..............................

1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSETIME.....................

1-3 1.14 FREQUENCYII0TATION...........................................

1-3 1.15 IDENTIFIED LEAKAGE...........................................

1-3 1.15.a L,..........................................................

1-3 1.16 MASTER RELAY TEST............................................

1-3 1.17 MEMBER (S) 0F THE PUBLIC........................

1-3 1

1.18 0FFSITE DOSE CALCULATION MAh0AL..............................

1-4 1.19 OPERABLE - OPERABILITY.......................................

1-4 1.19.a OPERATING LIMITS REP 0RT.....................................

1-4 1.20 OPERATIONAL MODE - M0DE......................................

1-4 1.20.a P,..........................................................

1-4 l

1.21 PHYSICS TESTS................................................

1-4 1.22 PRESSURE 900NDARY LEAKAGE....................................

1-4 r-+1. 23 PROCESS CONTROL PR0 GRAM......................................

-5 1.24 PURGE - PURGING..............................................

1-5 1.25 QUADRANT POWER TILT RAT10....................................

1-5 1.26 RATED THERMAL P0WER..........................................

1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................

1-5 1.28 REPORTABLE EVENT.............................................

1-5 Q3,2t.* m u And vgw a m umm N m ).........-

y.3 )

BYRON - UNITS 1 & 2 I

AMENDMENTNO.p

DEFINITIONS OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alam/ Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Sections 6.8.4e and f, and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7.

I OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATING LIMITS REPORT 1.19.a The OPERATING LIMITS REPORT is the unit-specific document that provides operating limits for the' current operating reload cycle. These cycle-specific operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9.

Plant operation within these operating limits is addressed in individual specifications.

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

f.

1.20.a P shall be the maximum calculated primary containment pressure (44.( psi,g) for the design basis loss of coolant accident.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests perfomed to measure the fundamental nuclear characteristics of the core and related instrumentation: (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAXAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component ody, pipe wall, or vessel wall.

BYRON - UNITS 1 & 2 1-4 AMENDMENT NO. 81

... _ _.. =-

I i

INSERT A i

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.22.a The PTLR is the unit-specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, and the pressurizer power operated relief valve (PORV) lift settings for the current reactor vessel fluence period.

These pressure and temperature limits shall be determined for each fluence period in ~

accordance with Specification 6.9.1.11. Unit operation within these limits is addressed in LCO 3.4.9.1," Pressure / Temperature Limits," and LCO 3.4.9.3, " Overpressure Protection Systems."

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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS

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LIMITING CONDITION FOR OPERATION

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s 3.4.9.lfTheReac Coolant stem (excep the pre urizer) te erature nd d

p gr ssure shall limited i accordance th the mit lines own on gures f.4-2aand3.

a for Uni 1 (Figures

.4-2b an 3.4-3b for nit 2) ring heatup, coo own, crit ality, and i ervice ak and hyd static sting th:

a.

A maximu eatup of 10 F in an 1-hour per d,

j b.

Am mum cooldown f 100*F n any 1-ho perio, and c.

maximum temp ature nge of les than o equal 10*F any 1-hour peri during 'nservice by ostati and le testi opera ons 4

above the eatup a cooldown 1 it cur s.

APPLICABILITY:

At all times.

,,.4.fTN ACTION:

With any of the d: = limits,5 minutes; perform an engineering exceeded, restore the temperature and/or pressure, i

to within the limit within 3 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 continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T,yg and pressure to less than 200*F and j

500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant 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 sur'veillance specimens shall be removed and examined, to determine changes in material properties, 2

as required by 10 CFR Part 50, Appendix H, in accordance with the schedule irh P7~/J T:ti:

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IslCATED TDIPERATURE (*F) t FIGURE 3.4-2A REACTOR COOLANT SYSTEM HEATUP LIMITATIONS 1

APPLICA8LE UP TO 32 EFPY*(UNIT 1)

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MATERIAL PROPERTY BASIS ole /e,,4 s

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i REACTOR COOLANT SYSTEM HEATUP LIMITATIONS 1

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IGURE 3.4-3a REACTOR COOLANT SYSTEN C00LDOWN LIf1ITATI NS APPLICABLE UP TO 32 EFPY (UNIT 1)

BYRON - UNITS 1 & 2 3/4 4-35

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RON - UNITS 1 & 2 3/4 4 36 AMENDMENT NO. 3

i TABLE 4.4-5

=

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REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITH0RAWAL SCH

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UNIT-1 i

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CAPSULE VESSEL LEAD i

a NUMER LOCATION

. ACTOR WITHDRAWAL TINE (EFPY)*

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Withdrawal time may be modified to coincide with those. refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.

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4 REACTOR COOLANT SYSTEM 4

PRESSURIZER 1

LIMITING CONDITION FOR OPERATION i

l 3.4.9.2 The pressurizer temperature shall be limited to:

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

1 b.

A maximum cooldown of 200*F in any 1-hour period, and I

' c.

A maximum spray water temperature differential of 320*F.

APPLICABILITY:- At all times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits,

{

restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer i

remains acceptable for continued operation or be in at least HOT STANDBY within I

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i SURVEILLANCE REQUIREMENTS i

4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.

The i

spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

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,e REACTOR COOLANT SYSTEM.

OVERPRESSURE PROTECTION SYSTEMS

(

LIMITING CONDITION FOR OPERATION 1

3.4.9.3 At least two overpressure protection devices shal'1 be OPERABLE, and

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each device shall be either:

i

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a.

A residual he6t removal (RHR) suction relief valve with a lift f

setting of less than or equal to 450 psig, or.

b.

A power operated relief valve (PORV) with a lift setpoint that j

varies with RCS temperature which coes not exceed the limit established i G r: 2.' '. g g4 APPLICABILITY: MODES 4, 5, and 6 with the reactor vessel head on.

ACTIO_N_:

4 a.

With one of the two required overpressure protection devices inoperable in MODE 4, restore two overpressure protection devices to j

OPERABLE status within 7 days or depressurize and vent the RCS i

through at least a 2 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

With one of the two required overpressure protection oevices

{

inoperable in MODES 5 or 6, restore two overpressure protection i

devices to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or vent the RCS through j'

at least a 2 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

4 i

j c.

With both of the required overpressure protection oevices inoperable, depressurize and vent the RCS through at least a 2 square inch vent

{

within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

1 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 lockea, sealed, or otherwise secured in the open position; i

otherwise, verify the vent pathway every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , In the event either.the PORVs, RHR suction relief valves, or the RCS i

e.

l vents are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to i

i Specification.6.9.2 within 30 days. The report shall describe the l

circumstances initiating the transient, the effect of the PORVs, RHR j

suction relief valves, or RCS vents on the transient, and any corrective action necessary to prevent recurrence.

j f.

The provisions of Specification 3.0.4 are not applicable.

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. 37 BYRON - UNITS 1 & 2

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RON - UNITS 1 & 2 3/4 4-40b AMEND *iENT NO. 3

REACTOR COOLANT SYSTEM BASES i

SPECIFIC ACTIVITY (Continued) take corrective action.

Information obtained on iodine spiking will be used s

i to assess the parameters associated with spiking phenomenon.

A reduction in i

frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

4 3/4.4.9 PRESSURE / TEMPERATURE LIMITS k-

' he temperature and pressure changes during heatup and, cooldown are

\\

Clt t H g" '"

to be consistent with the requirements given in the ASME Boiler and imite Pressure essel Code,Section III, Appendix G i

WA 1.

Th reactor coolant temperature and pressure and system heat and i

In5ut cool wn rates (with the exception of the pressurizer) sha be limite in accordance with Figures 3.4-2 and 3.4-3 for t service B

period s ciftee thereon:

a.

Allowab e combinations of pressure and temper ure for specific temperat e change rates are below and to t right of the limit line shown.

Limit lines for cooldpwn rates between l

those presen d may be obtained by inte olation; and b.

Figures 3.4-2 an 3.4-3 define limi)l operation, other inherent to assure prevention of non-ductile failur only.

For no a plant characteristic e.g., p heat addition and pressurizer heater capacity, may 1 it th heatup and cooldown rates that I

can be achieved over cer't i pressure-temperature ranges.

2.

These limit lines shall be c cul ed periodically using methods provided below, 3.

The secondary side of e steam generat must not be pressurized above 200 psig if t temperature of the team generator,is below 70'F, 4.

The pressuriz heatup and cooldown rates shal ot exceed 100*F/hr and 200* F/

respectively.

The spray shall.not e used if the temperatur difference between the pressurizer and he spray fluid is grea r than 320'F, and 5.

Sys m preservice hydrotests and in-service leak and hyd tests s 11 be performed at pressures in accordance with the req rements f ASME Boiler and Pressure Vessel Code,Section XI.

e fracture toughness properties of the ferritic materials in the re tor ve 1 are determined in accordance with the 1973 Summer Addenda to Section I

the ASME Boiler and Pressure' Vessel and Code.

BYRON - UNITS 1 & 2 B 3/4 4-7

- - -. = -

REACTOR COOLANT SYSTEM j

BASES i

PRESSURE / TEMPERATURE LIMITS (Continued)

Heatup and cooldown limit curves are calculated using the most limiting j

lue of the nil-ductility reference temperature, RTNDT, at the end of

~

32 ffective full power years for Unit 1 (16 effective full power years f Unit ) of service life.

The 32 EFPY for Unit 1 (16 EFPY for Unit 2) s vice j

life p iod is chosen such that the limiting RTNDT at the 1/4T locatio in the j

core reg n is greater than the RT f the limiting unirradiated terial.

NOT l

The select 1 of such a limiting RT assures that all component in the NDT l

Reactor Coola System will be operated conservatively in accor ance with 1

applicable Code equirements.

l The reactor ssel materials have been tested to date ine their initial j

RTNDT; the results these tests are shown in Table B 3/.4-1.

Reactor i

l operation and resulta fast neutron (E greater than 1 V) irradiation can

]

}

cause an increase in th RT Therefore, an adjust reference temperature,

)

NDT.

l based upon the fluence, c er content and nickel ntent of the material l

)

{

in question, can be predict using Figure B 3/4.

1 and the largest value of l

4 ART computed by either Re latory Guide 1.99 Revision 2, " Radiation NDT Embrittlement.of Reactor Vessel aterials" or he Westinghouse Copper Trend i

Curves shown in Figure B 3/4.4-2.

The heat and cooldown limit curves of i

Figures 3.4-2 and 3.4-3 include pr icted djustments for this shift in i

RTNDT.at the end of 32 EFPY for Unit

(

EFPY for Unit 2) as well as adjust-1 ments for possible errors in the pres e and temperature sensing instruments.

j Revised heatup and cooldown curves h e en generated for Unit 2 in j

accordance with' Regulatory Guide 1 9 Revi ion 2.

For Unit 1, the curves l

j remain the same.

However, the a icabilit date has been. reduced per RG 1.99 i

Revision 2 to 29.5 EFPY for hea p.

The Byro Unit'I applicability date of 32 l

EFP,Y for cooldown remains the ame, j

Values of ART dete ned in this manner m be used until the results NDT from the material survei ance program, evaluated a ording to ASTM E185, are available.

Capsules wi be removed in accordance wi the requirements of

[

ASTM E185-73 and 10 CF Part 50, Appendix H.

The surve'llance specimen with-drawal schedule is own in Table 4.4-5.

The lead facto represents the rela-tionship between t fast neutron flux density at the loca ion of the capsule and the inner wa of the reactor vessel.

Therefore, the re ults obtained from j

the surveillan specimens can be used to predict the future diation damage to the reacto vessel material by using'the lead factor and the ithdrawal time of the caps e.

The heatup'and cooldown curves must be recalcul ed when the j-ART det rmined from the surveillance capsul'e exceeds the calcul ed ART NDT NDT for the quivalent capsule radiation exposure.

J j

lowable pressure-temperature relationships for various heatup an coo own rates are calculated using methods derived from Appendix G in i

5 tion III of the A5ME Boiler and Pressure Vessel Code as required by J

pendix G to 10 CFR Part 50, and these methods are discussed in detail in 1

WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975.

BYRON - UNITS 1 & 2 B 3/4 4-8 AMENDMENT NO.37

REACTOR COOLANT SYSTEM l

BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The general method for calculating heatup and cooldown limit curves is b ed upon the principles of the linear elastic trccture mechanics (LEFM) tec ology.

In the calculation procedures a semi-elliptical surface defee with depth of one quarter of the wall thickness, T, and a length of 3/

is assume o exist at the inside of the vessel wall as well as at the ou de of I

the vesse wall.

The dimensions of this postulated crack, referred in

' Appendix G ASME Section III as the reference flaw, amply execed e current capabilities inservice inspection techniques.

Therefore, the actor opera-tion limit curv developed for this refeiwnee crack are conser tive and provide sufficient safety rgins for pistection against non-ductile ilure.

To assure that the radiation rittlement effects are accounted for the calculation of the limit curves, e most limiting value of the nil-tility reference temperature, RTNDT, is ed and this includes the radi on-induced shift, ARTNDT, corresponding to e end of the period for ch heatup and cooldown curves are generated.

The ASME approach for calcu ting the alkwable limit curves 'for various heatup and cooldown rates specifie that th total stress intensity factor, 1

K, for the combined thermal and pre ure tresses at any time during heatup y

or cooldown cannot be greater than the ference stress intensity factor, Kg, for the metal temperature at that t K

is obtained from the referener g

fracture toughness curve, defined in Appendix to the ASME Code. The KIR curve is given by the equatio XIR = 26.78 + 1.223

[0.0145(T-RTNDT + 160))

(1)

Where:

K is the ref rence stress intensity factor as function of the metal IR i

temperature T and e metal nil-ductility reference temperature RT

Thus, NDT.

the governing e tion for the heatup-cooldown analysis is de ed in Appendix G of the ASME Co as follows:

CK K

IKIR (2) y It Where:

Kyg = the stress intensity factor caused by meinbrane (pressure)

stress, kit = the stress intensity factor caused by the thermal gradients, BYRON - UNITS 1 & 2 B 3/4 4-9 1

q 1

REACTOR COOLANT SYSTEM 3

BASES i

PRESSURE / TEMPERATURE LIMITS (Continued) i KIR = c nstant provided by the code as a function of temperature relative j

to the RT of the material, NDT C=

2.0 for level A and B service limits, and l

1.5 for inservice hydrostatic and leak test operations.

j

=

y At an time during the heatup or cooldown transient, K is de rmined by IR i

the metal tem rature at the tip of the postulated flaw, tlye app priate value j

for RTNDT, and e reference fracture toughness curve.

The th al stresses i

I resulting from t erature gradients through the vessel wal are calculated and then the corres nding thermal stress intensity factor KIT, for the reference flaw is comp ed.

From Equation (2) the pres re stress intensity factors are obtained an, from these, the allowable p ssures are calculated.

C00LDOWN For the calculation of the llowable pre ure versus coolant temperature during cooldown, the Code referen flaw is ssumed to exist at the inside of the vessel wall.

During cooldown, e con olling location of the flaw is always at the inside of the wall becah he thermal gradients produce tensile stressesattheinside,whichincreaseAthincreasingcooldownrates.

Allowable pressure-temperature relations are griera ed for both steady-state and finite

' cooldown rate situations.

From these rela 'ons, composite limit curves are constructed for each cooldown rate'of intere The use of the composite urve in the cool byn analysis is necessary because control of the cooldown pro dure is based on measbrement of reactor coolant temperature, whereas the }4miting pressure is actuail,y dependent on the material temperature at the tip 'of the assumed flaw.

During cogidown, the 1/4T vessel location is at a highe,r temperature than the fluid adjacent to the vessel ID.

This condition, of ed rse, is not true for the steady-sta situation.

It follows that at an given reactor coolant temperature, the developed during cooldown results n a higher value of K at the 1/4T locatio for finite IR cooldown rate than for steady-state operation.

Furthermore, i conditions exist such at the increase in K exceeds kit, the calculated a owable yg pressure uring cooldown will be greater than the steady ' state value, e above procedures are needed because there is no direct control tem rature at the 1/4T location; therefore, allowable pressures may unkn ingly.

b violated if the rate of cooling is decreased at various intervals along ooldown ramp.

The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

BYRON - UNITS 1 & 2 8 3/4 4-10

/d<t<,

e TABLE B 3/4.4-la Y

REAC10R VESSEL TOUGHNESS (UNIT 1) k Average Upper M

-Shelf Energy Normal to 50 ft-35 il Principal Principal T

RT Wr ing Wrking i Cu P

NOT esp.

NOT Direction Direction COMPONENT Heat No.

Grade Q%

1%)

("F) 1_F"J Et-1 (ft-Ib) :

Closure Head Dome C3486-1 TJB CL1

.10

.016

-1

< 40 Closure Head Ring IV4566 AS CL2

.11

.007

< 80 20 125 Closure Head Flange 124K358VA1

.011 60

<100 60 145 p

Vessel Flange 123J219VA1

.012 10

< 70 10 152

, Inlet Nozzle IV4684/3V1320 8

-10

< 40

-10 117 i

t" IV4684/3V1320

.12

.008

-20

< 40

-20 116 w

IV4695 7

-20

< 10

-20 116 h"

IV4695

.12

.00

-20

< 10

-20 119 Outlet Nozzle IV4656

.11

.007 0

< 10 0

131

~ IV4656

.11

.007

-20

< 10

-20 131 f

2V2557

.11

.007

< 10

-20 112 2V2557

.11

.008

-10 50

-10 94 Nozzle Shell 123J2

.05

.010 20 76 20 138 184 l

Upper Shell**

SP 933

.05

.010(.73) 40

<1 40 139 156 Lower Shell**

SP-5951

.04

.014(.64) 10

< 70 10 150 160 g Bottom Head Rin IV4672

.012 0

< 60 115 j!j Botton Head C2815-1 A5338 CLI

.19

.009

-30 40

-20 118 N Upper to r

WF336

.024

.010(.70)

-30 30

-30 77*

l' y Shell rth Weld **

ti

Normal to Principal Welding Direction

- Calculations per Regulatory Guide 1.99 Revision 2 use nickel content shown in parenthesis.

l{

t

~..

.. -.. - -. ~..

. =. - -. ~. - ~ -

- - - = - ~

defeb g

TABLE B 3/4.4-lb 5

REACTOR VESSEL TOUGHNESS f

(UNIT 2)

I g

Average Upper l

q Shelf Energy

=

Normal to.

e-35

~

Principal Principal i 50 ft-1

Working Working '

T RT u

Cu P

NOT NOT Direction Direction Com>onent Heat No.

Grade Q%

Q%

(*F)

WF*

O'F (ft-lb)

(ft-lb) l Closure lead Dome C4375-2 533 B C1.1.12

.013

-40

<20

-40 114 Closure Head Ring 48C1300-1-1 AS C1.3

.05

.007

<30

-30 108 Closure Head Flange 2029-V-1 A508 C

.011 0

<60 0

157 l

Vessel Flange 124L556VA1

.008 30

<90 30 129

[

Inlet Nozzle 51-2979

.7

-10

<50

-10 130

["

51-2979

.07

.009

-20

<40

-20 121 l

2" 42-5105

.0 008 0

<60 0

122 i

42-5105

.07

.0 0

<60 0

121 Outlet Nozzle 11-5052

.09

.010

-10

<50

-10 108 l

I 11-5052

.08

.007

-10

<50

-10 121 4-2953

.09

.010 0

<40

-20 133 l

4-2956

.09

.009

-10

<50

-10 121 Nozzle Shel1 4P-610

.014 10

<70 10 155 l

Upper Shell*

4 9/

A508 C1.3

.01

.007(.70)

-20

-20 149 149 ll I'I 90297 g

Lower Shell*

490330/

A508 C1.3

.05

.008(.73)

-20

<40

-20 127 159 l

I-I Q

49C298 j

g Botton Head ng 4801566 1-1 A508 C1.3

.07

.007

-30

<30

-30 126 4 Bottom d Dome C3053-1 A5338, C1.1

.06

.004

-30 40

-20 121 5 Upp Shell to Lower WF447 SAW

.059

.009(.62) 10

<70 10 l

(

11 Girth Weld *

~

" Weld HAZ

-60

<0

-60 143 f

Calculations per Regulatory Guide 1.99 Revision 2 use nickel content shown in parenthesis.

l

4

)

1 l

1019 1/4T.1 x 1018 N/cm2 j

8 s

!/

4 4

i

\\

/

z wu2w

-- 3/4T. 2.7 x 1018 N/cm.

2 a

2

\\

au-2O c:>

3 w

i 2 1018

- - - - - - - - - - - \\

8

\\

s 4

i i

2 10 "

0 10 20 30 40 50 EFFECTIVE FULL POWER (Years)

FIGURE B 3/4.4-1

' FAST NEUTRON FLUENCE (E > 1 MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE BYRON - UNITS 1 & 2 B 3/4 4-13

old8 m

D Oz C

"O 5

N~

I I i I I i i lillllllliillilillliti41 lillf illlillliliit: 9 d

0.30% COPPER BASE,0.25% WELD 'I

~~

'~

~

~'

i l' (M UPPER LIMIT i

lilbf

,,)

' f l

i

...ii l l []

M i

.~ 0.

COPPER BASE. O.20% WELD

,oio u

l,

+

p.

\\

d,,..I,pi.

5 r) q, M

.ii 8,,,g

ill gin' t-s I

,",illH.

N s

i s

i ov'"',

"g U

5 I

j :

h="

s8, r"4m i

i g

ni"T

,;;,,s d '"

e inni-

,g l

iin s''

,.,,,,,as""'

o

di 8

,o o

~~~

3 i

O

o

'i,,

q o

ui LOWER LIMIT

,m g,o i '""",,,,,,,o'"

e s

,n"""

i g,,,

. /

(

l m!

F y

,,,,,"g,,,,iii

.ia s

,,g,,,,

i l 1!!

d..

I

/

.,,a="'

_1 oi

=

z

='-...,"

g n",,

! l j:

g

g

j j

g. 4,,-='-

,,,ii"'

w e

80 l

.I l--'

s.K'

' t 8 ,

N

<3

_.w a

8

,['j:

[-,.

I 18

- i..

~

j M-

','Ji q,,

'h

"'"~

0.20% COPPER BASE,0.15% WELD !

o s

60 i;t i

r";j, c

3 d

! ! 5

,,,,' 15% COPPER BASE 0.10% WELD i i

u d,,,

l l !

Il m.,,,,s x

1 D

l

'"p---

]

N0.10 OPPER BASE,0.05% WELD

,;;,ir.is=

"I i

\\,

i il nl l

i 1

..p t

I l

i i

i

" i 9

p j

p

..}

..'h l

.I

[

l

! ll!

I I

h 1

l i

20 1s 2

4 s

e 19 2

4 s

10" 10 10 I

FAST NEUTRON FLUENCE IN/CM E>1 MeV)

FIGURE B 3/4.4-2

~

EFFECT OF FLUENCE AND COPPER ON SHIFT OF RT FOR NDT REACTOR VESSEL STEELS EXPOSED TO IRRADIATION AT 550*F

j REACTOR COOLANT SYSfEM i

BASES i

\\ PRESSURE /TEMPERATURELIMITS(Continued) 4 A notch in the cooldown curve of Figure 3.4-3 may be present due to th l

l add constraint on the vessel closure flange given in Appendix G of 10 CF 50.

l This nstraint requires that, at pressures greater than 20% of the pres vice system drostatic test pressure, the flange regions that are highly s essed l

by the b t preload must exceed the RTNDT of the material by at least 20'F.

l The flange TNOT + 120'F impinges on the cooldown curves and there re the j

notch is requ ed.

If no notch is present, this indicates that e vessel l

closure flange gion has been determined not to be limiting.

1 HEATUP Three separate c culations are required to determi e the limit curves for finite heatup rates. As is done in the cooldown a lysis, allowable 1

pressure-temperature rel ionships are developed for teady-state conditions as well as finite heatup r e conditions assuming t presence of a 1/4T defect at the inside of the ssel wall.

The the 1 gradients during heatup l

produce compressive stresses the inside of t wall that alleviate the l

tensile stresses produced by in rnal pressure The metal temperature at the j

crack tip lags the coolant tempe ture; ther ore, the K for the 1/4T crack IR l

during heatup is lower than the K or th 1/4T crack during steady-state yg l

conditions at the same coolant tempera e.

During heatup, especially at the end of the transient, conditions may a t such that the effects of compressive j

thermal stresses and different K

's or eady-state and finite heatup rates IR I

do not offset each other and the p essure-t erature curve based on steady-state i

conditions no longer represents lower bound f all similar curves for finite heatup rates when the 1/4T fla is considered.

Therefore, both cases have to I

be analyzed in order to assur that at any coola temperature the lower value 1

of the allowable pressure c culated for steady-st e and finite heatup rates

{

is obtained.

l The second portion f the heatup analysis concern the calculation of pressure-temperature mitations for the case in which a 1/4T deep outside surface flaw is ass d.

Unlike the situation at the ves 1 inside surface, l

the thermal gradie s established at the outside surface du ing heatup produce j

stresses which a tensile in nature and thus tend to reinfo e any pressure stresses prese These thermal stresses, of course, are depe ent on both i

the rate of h tup and the time (or coolant temperature) along t heatup ramp.

Furt recre, since the thermal stresses, at the outside ar tensile and

}

increase w h increasing heatup rate, a lower bound curve cannot be efined.

j Rather,-

ch heatup rate of interest must be analyzed on an individu basis.

F lowing the generation of pressure-temperature curves for both t j

stea -state and finite heatup rate situations, the final limit curves a j

pr uced as follows. A composite curve is constructed based on a point-by j

p nt comparison of the steady-state and finite'heatup rate data.

At any i

iven temperature, the allowable pressure is taken to be the lesser of the j-three values taken from the curves under consideration.

I

?

BYRON - UNITS 1 & 2 B 3/4 4-15 AMENDMENT NO.37 4

1 i

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

N The use of'the composite er

is necessary to set conservative heatup 1

itations because it is possible for conditions to exist such that over e

cou of the heatup ramp the controlling condition switches from the i de to the utside and the pressure limit must at all times be basen :n a ysis of the t critical criterion.

Finally, he composite curves for the heatup rate data an the cooldown rate data are a sted for possible errors in the pressure temperature sensing instrument by the values indicated on the respec e curves.

Although the press izer operates in temperature anges above those for which there is reason for oncern of nonductile fa re, operating limits are provided to assure compat bility of operatio with the fatigue analysis performed in accordance with t ASME Code re rements.

The OPERABILITY of two PORVs, two R suction relief valves, or one PORV and one RHR suction relief valve, 11 be protected from pressure transients an RCS vent opening of at least 2 square inches ensures that the RCS i

which could exceed the limits of ndix D to 10 CFR Part 50 when one or more of the RCS cold legs are less t or egaal 350*F.

Either PORV has adequate i

relieving capability to prote the RCS from ov ressurization when the tran-sient is limited to either:

) the start of an i le RCP with the secondary water temperature of the as generator less than dNequal to 50*F above the i

RCS cold leg temperatuyes, or (2) the start of a centrtfugal charging pump and itsinjectioninto ater solid RCS.

i These two enarios are analyzed to determine the result overshoots n. W r.; r, si it PDRV actuation with a stroke time of 2.0 secon fr:s full j

closed to 1 open.

Figure 3.4-4 is based upon this analysis an presents l

the maxi allowable PORV variable setpoint such that, for the two rpres' surization transients noted, the resulting pressure will not exceed the Ap

' dix G reactor vessel NDT limits (nominal 10 effective full power yea Unit 1 only).

4 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness l

of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

I BYRON - UNITS 1 & 2 B 3/4 4-16 AMENDMENT.NO. 38, 44

INSERT B All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and

.~ shutdown (cooldown) operations, power transients, and reactor trips. LCO 3.4.9.1,

" Pressure / Temperature Limits," limits the presnre and temperature changes during RCS heatup and cooldown to within the design assum.otions and the stress limits for cyclic operation.

The PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) contains pressure and temperature (P/T) limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature.

Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

LCO 3.4.9.1 establishes operating limits that provide a margin to non-ductile failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB). The vessel is A component most subject to non-ductile failure, and the LCO limits apply to the entire A S, except the pressurizer, which has different design characteristics and operating functions.

10 CFR 50, Appendix G, requires the establishment of P/T limits for specific material fracture toughness requirements of the RCPB materials. Appendix G of 10 CFR 50 requires an adequate margin to non-ductile failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use'of the American Society of Mechanical Engineers (ASME) Code,Section XI, Appendix G.

The neutron embrittlement effect on the material toughness is reflected by mereasing the Nil Ductility Reference Temperature (RTer) as exposure to neutron fluence increases.

The actual shift in the RTm1 of the vessel material will be established periodically by removing and evaluating the irradiated rer.ctor vessel material specimens, in accordance with ASTM E 185 and Appendix H of 10 CFR 50. The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99, Revision 2.

j i

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive.

At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, l

i 4

1 i

a

4 INSERT B (continued) different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls during heatup and cooldown, respectively.

The criticality limit curve includes the 10 CFR 50, Appendix G requirement that it be 2 40 F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.1.1.4, " Minimum Temperature for Criticality."

i The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in non-ductile failure of the RCPB, possibly leading to a nonisolable leak or loss-of-coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. The ASME Code,Section XI, Appendix E, provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

f OVERPRESSURE PROTECTION SYSTEMS The Low Temperature Overpressure Protection (LTOP) System controls the RCS pressure at low temperatures so the integrity of the RCPB is not compromised by violating the P/T limits of 10 CFR 50, Appendix G. The reactor vessel is the limiting RCPB component for l

demonstrating such protection. The PTLR provides the maximum allowable actuation logic setpoints for the pressurizer Power-Operated Relief Valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the 10 CFR 50, Appendix G requirements during the MODES in which the LTOP system is necessary.

The reactor vessel material is less ductile at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and the material becomes less resistant to stress at low temperatures. RCS pressure, therefore, is maintained low at low temperatures and is increased only within the limits specified in the PTLR.

The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only during shutdown; a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the RCS P/T limits by a significant 2

j

{

INSERT B (continued)'

1 amount could cause non-ductile failure of the reactor vessel. LCO 3.4.9.1, i

" Pressure / Temperature Limits," requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the PTLR limits.

i LCO 3.4.9.3, " Overpressure Protection Systems," provides RCS overpressure protection by having a minimum coolant input capability and having adequate pressure relief capacity.

j Limiting coolant input capability requires all Safety Injection (SI) pumps and all but one

{

i charging pump (a centrifugal charging pump) incapable ofinjection into the RCS and isolation 4

of the SI accumulators. The pressure relief capacity requires either two redundant RCS relief i

valves,'or a depressurized RCS and an RCS vent of sufficient size. One RCS relief valve or j

the open RCS vent is the overpressure protection device that acts to terminate an increasing j

i pressure event.

[

p

[

With minimum coolant input capability, the ability to provide core coolant addition is j

restricted. The LCO does not require the makeup control system deactivated or the SI

' actuation circuits blocked. Due ~to the lower pressures in the LTOP MODES and the

'l expected core decay heat levels, the makeup system can provide adequate flow via the j

j makeup control valve. If conditions require the use of more than one centrifugal charging pump for makeup in the event ofloss ofinventory, then pumps can be made available through manual actions.

i 1

The LTOP System for pressure relief consists of two PORVs with reduced lift settings, or two

?

Residual Heat Removal (RHR) suction relief valves, or one PORV and one RHR suction relief valve, or a depressurized RCS and a RCS vent of sufficient size. Two RCS relief valves are J

l.

required for redundancy. One RCS relief valve has adequate relieving capability to prevent _

j overpressurization for the required coolant input capability.

PORV Reauirements As designed for the LTOP System, each PORV is signaled to open if the RCS pressure approaches a limit determined by the LTOP actuation logic. The LTOP actuation logic monitors both RCS temperature and RCS pressure and determines when a condition is approaching the PTLR limits. The wide range RCS temperature indications are auctioneered

- to select the lowest temperature signal.

The lowest temperature signal is processed through a function generator that calculates a i

pressure limit for that temperature. The calculated pressure limit is then compared with the

. indicated RCS pressure from a wide range pressure channel. If the indicated pressure meets

'or exceeds the calculated value, a PORV is signaled to open.

):

i 3

INSERT B (continued)

The PTLR presents the PORV setpoints for the LTOP system. The setpoints are normally staggered so only one valve opens during a low temperature overpressure transient. Having the setpoints of both valves within the limits in the PTLR ensures that the 10 CFR 50, Appendix G limits will not be exceeded in any analyzed event.

When a PORV is opened in an increasing pressure transient, the subsequent relief will causei the pressure increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close. The pressure continues to decrease below the reset pressure as the valve closes.

RHR Suctionf elief Valve Reauirernents During the LTOP MODES, the RHR System is operated for decay heat removal and low pressure letdowa control. Therefore, the RHR suction isolation valves are open in the piping from the RC5 hot legs to the inlets of the RHR pumps. While these valves are open, the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS.

The RHR suction isolation valves must be open to make the RHR suction relief valves OPERABLE for RCS overpressure mitigation. The RHR suction reliefvalves are spring loaded, bellows-type relief valves with pressure tolerances and accumulation limits established by Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for Class 2 relief valves.

RCS Vent Reauirements Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain the RCS at containment ambient pressure in an RCS overpressure transient, if the reliesing requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass or heat input transient, and maintaining pressure belc / the P/T limits. The required vent capacity may be provided by one or more vent paths.

For an RCS vent to meet the flow capacity requirement, it requires removing a pressurizer safety valve; remoeg a PORV's internals, and disabling its block valve in the open position; or similarly establishmg any comparable vent. The vent path (s) must be above the level of reactor coolant, so as not to drain the RCS when open.

4

I.

ADNINISTRATIVE CONTROLS 4

CRITICALITY ANALYSIS OF BYRON AND BRAIDWD00 STATION FUEL STO 1

6.9.1.10 Fuel enrichment limits for storage shall be established and 4

documented in the CRITICALITY ANALYSIS OF BYRON AND BRAIDWD00 S l

STORAGE RACKS.

The analytical methods used to determine the maximian fuel enrichments shall be those previously reviewed and approved by the NRC in

" CRITICALITY ANALYSIS OF 8YRON AND BRAIDWD00 STATION FUEL ST j

The fuel enrichment limits for storage shall be detemined so that all applicable limits (e.g., suberiticality) of the safety analysis are met.

4

) N/

The CRITICALITY ANALYSIS OF BYRON AND 8AAIDWD00 STATION FU i

RACKS report shall be provided upon issuance of any charges, to the NRC i

C Document Control Desk, with copies to the Regional Admin <strator and the l

Resident Inspector.

I j

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

6.10 RECORD RETENTION i

Code of Federcl Regulations, the following records shall be reta i'

least the minimum period indicated.

}

6.10.1 The following records shall be retained for at least 5 years:

Records and logs of unit operation covering time interval at each a.

power level; b.

Records and logs of principal maintenance activities inspections, i

repair and replacement of principal items of equipmen,t related to nuclear safety; c.

All REPORTABLE EVENTS; d.

Records of surveillance activities inspections, and calibrations I

requiredbytheseTechnicalSpecifications; Records of changes made to the procedures required by e.

Specification 6.8; f.

Records of radioactive shipments; 4

Records of sealed source and fission detector leak tests and results; g.

j and h.

Records of annual physical inventory of all sealed source material j

of record.

i 6.10.2 The following records shall be retained for the duration of the unit

{

Operating License:

a.

Records and drawing changes reflecting unit design modifications i

made to systems and equipment described in the Final Safety Analysis Report; i

i b.

Records of new and irradiated fuel inventory, fuel transfers and j

assembly burnup histories; i

BYRON - UNITS 1 & 2 6-23 AMENDMENT ND. 50 I

INSERT C Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE I,IMITS REPORT (PTLR) j 6.9.1.11 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, as well as heatup and cooldown rates and pressurizer power-operated relief valve lift settings, shall be established and documented in the 1

PTLR for the following:

1)

LCO 3.4.9.1, " Pressure / Temperature Limits," and j

2)

LCO 3.4.9.3," Overpressure Protection Systems."

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP-14040-NP-A, Revision 2, January 1996, " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," as approved by the NRC by letter dated October 16,1995 (TAC M91749); and NRC Safety l

Evaluation Report dated t

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period i

and for any revision or supplement thereto.

t I

i i

1

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

i ATTACHMENT B-1B PROPOSED CHANGES TO PROPOSED IMPROVED TECHNICAL SPECIFICATIONS FOR FACILITY OPERATING LICENSES NPF-37 AND NPF-66 BYRON NUCLEAR POWER STATION, UNITS 1 AND 2 REVISED PAGES 5.0-43 B 3.4-10 B 3.4-15 B 3.4-17 B 3.4-67 B 3.4-75 B 3.4-77

i Reporting Requirements i

5.6 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) a.

RCS pressure and temperature limits for heat up. cooldown.

low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates, and Power Operated Relief Valve (PORV) lift settings shall be established and documented in the PTLR for the following:

LCO 3.4.3. "RCS Pressure and Temperature (P/T) Limits": and LCO 3.4.12. " Low Temperature Overpressure Protection":

b.

The analytical methods used to determine the RCS pressure and temperature limits shall* be those previously reviewed and approved by the NRC. specifically those described-in the following documents:

e-A 1.

WCAP-14040%. " Methodology used to Develop Cold Overpressure Mitigating System setpoints and RCS Heatup and Cooldown Limit Curves." Revision S.1 y7 )W 2:;,uuci 1994. Approval provided by NRC Safety Evaluation Reporty dated October 16.1995;aa 2.

Gm:d icuer w ns.

Inn.iai Pressure a.d Te e ture o :0-t." deted

^pprev:1

~

Li-its e

e m.1dcd & NRC Safety Evaluation Report, dated

---end 10per tem 500 Just"iaticr.

6 D. M rcr.c=' P ]

y c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.7 Post Accident Monitorina Reoort When a report is required by Condition C or H of LCO 3.3.3. " Post Accident Monitoring (PAM) Instrumentation." a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

(continued)

BYRON - UNITS ? & 2 5.0-43 Revision A

RCS P/T,B 3.4.3 Limits

\\

B 3.4 REACTOR COOLANT SYSTEM (RCS) l B 3.4.3 RCS Pressure and Temperature (P/T) Limits i

BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.

These loads are introduced by startup (heatup) and shutdown (cooldown) operations. power transients. and reactor trips.

This.C0 limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

{

The PTLR contains P/T limit curves for heatup. cooldown.

Inservice Leak and Hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature (Ref. 1).

Each P/T-limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation j

is within the allowable region.

Ao^'ph e LCO establishes operating limits that provide a margin to failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB).

The vessel is the component most subjec brittle failure, and the L

itsann(

the entire RCS (except the pressurizer).

The limits do not apply to the pressurizer. which has ifferent design characteristics and operating functions.

10 CFR 50. Appen (Ref. 2), requires the establishment of P/T limits for spe ific material fracture toughness requirements of the R materials.

'ference 2. requires an adequate margin to britth failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.

It mandates the use of the American Society of Mechanical Engineers (ASME) Code. Section JM. Appendix G (Ref. 3).

g (continued)

BYRON - UNITS 1 & 2 B 3.4-10 Revision A

l RCS P/T Limits

~

B 3.4.3' f

BASES i

ACTIONS A.1 and A.2. (continued) l Condition ~ A is modified by a Note requiring Required Action A.2 to be completed whenever the Condition is i

l entered.

The Note emphasizes the need to perform the l

evaluation of the effects of the excursion outside the I

allowable limits.

Restoration alone per Recuired Action A.1 is insufficient because higher than analyzec stresses may have occurred and may have affected the RCPB integrity.

B.1 and B.2 If a Required Action and associated Com)letion Time of Condition A are not met, the unit must ye placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress or a sufficiently severe event caused entry into an unacceptable region.

Either possibility indicates a need for more careful examination of the event. best accomplished with the RCS at reduced pressure.:r.d tc per:ture.

In reduced 3

pressure...

....r.....

conditions, the possibility of propagation with undetected flaws is decreased.

If the required restoration activity of Required Action A.1 cannot be accomplished within 30 minutes. Required Action B.1 and Required Action B.2 must be implemented to reduce pressure and temperature.

If the required evaluation for continued operation cannot be 1

accomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are

-l indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in Required Action B.1 and Required Action B.2.

A favorable engineering evaluation must be completed and documented before returning to operating pressure and temperature conditions.

Pressure and temaerature are reduced by bringing the unit'to MODE 3 within 6 lours and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

(continued)

BYRON - UNITS 1 & 2 B 3.4-15 Revision A

RCS P/T Limits

B 3.4.3 BASES- (continued) 1.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS.

Verification that operation is within the PTLR limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication-available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments. 30 minutes permits assessment and correction for minor deviations within a reasonable time.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown. and ISLH testing. This SR is not required during critical operations because the combination of LCO 3.4.2 establishing a lower bound and the Safety Limits establishing an upper. bound will provide adequate controls to prevent a change in excess of 100*F prior to entry into the performance condition of heatup and cooldown operations.

j

-wf-A,.Ressaaa REFERENCES 1.

WCAP-14040^ " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Junc 10?4.

2.

10 CFR 50. Appendix G.

CIjanga7 "4

1 3.

ASME. Boiler and Pressure Vessel Code. Section 134..

Appendix G.

dxi 4.

ASTM E 185-82. July 1982.

5.

10 CFR 50. Appendix H.

6.

Regulatory Guide 1.99, Revision 2. May 1988,

~

7.

ASME. Boiler and Pressure Vessel Code.Section XI.

Appendix E.

l 1

'4 BYRON - UNITS 1 & 2 B 3.4-17 Revision A 4

LTOP Syster

'B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 Low Temperature Overpressure Protection (LTOP) System BASES 4

BACKGROUND-The LTOP System controls RCS pressure at low temperatures so the integrity of the Reactor Coolant Pressure Boundary (RCPB) 1s not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50. Appendix G (Ref. 1).

The reactor vessel is the limiting RCPB component for 4

J demonstrating such protection.

The PTLR provides the maximum allowable actuation logic setpoints for the pressurizer Power Operated Relief Valves (PORVs) and the i

maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the Reference 1 requirements during the MODES in which LTOP is necessary.

The reactor vessel material is less ductile at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates. the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2).

RCS pressure. therefore, is maintained low at low temperatures and is increased only within the limits specified in the PTLR.

i 1

The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only while shutdown:

a pressure fluctuation can occur more quickly than an operator can react to relieve the condition.

Exceeding th RCS P/T limits by a significant amount could cause brittlge, nes - /wedd

[7kl4a2?crect'ngofthereactorvessel.

LCO 3.4.3. "RCS Pressure and Temperature (P/T) Limits." requires administrative i

control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the PTLR limits.

i (continued)~

BYRON - UNITS 1 & 2 B 3.4-67 Revision A

-.. ~ -. - - -

LTOP, System "B 3.4.12 BASES 1

LCO The elements of the LCO that provide low temperature (continued) overpressure mitigation through pressure relief are:

a.

Two OPERABLE PORVs:

1 b.

Two OPERABLE RHR suction relief valves:

One OPERABLE PORV and one OPERABLE RHR suction relief c.

valve; or d.

A'depressurized RCS and an OPERABLE RCS vent.

A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set to the limit required by the PTLR and testing proves its ability to open at this setpoint. and i

motive power is available to the te vehes and the+e its control circuitf.

Cpogy C'

An RHR suction relief valve is OPERABLE for LTOP when its RHR suction isolation valves are open. its setpoint is 4

s 450 psig, and testing has proven its ability to open at this setpoint.

An RCS vent is OPERABLE when open with an area of 3

= 2.0 square inches.

Each of these methods of overpressure prevention is capable

-of mitigating the limiting LTOP transient.

i I

I i

1 I

I j

I 3

(continued)

BYRON - UNITS 1 & 2 B 3.4-75 Revision A i

LTOP System-

"B 3.4.12 BASES ACTIONS C.1 and D.1 (continued)

An unisolated accumulator requires isolation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

This is only required when the accumulator pressure is at or more than the maximum RCS pressure for the existina temperature allowed by the P/T limit curves If'the Required Action and associated Ccapletion Time of Condition C are not met. Required Action D.1 must be performed in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Depressurizing the accumulators below the LTOP limit from the PTLR prevents an accumulator pressure from exceeding the LTOP limits if the accumulators are fully injected.

The Completion Times are based on operating experience that these activities can be accomplished in these time periods and on engineering evaluations indicating that an event requiring LTOP is not likely in the allowed times.

i j

U i

In MODE 4. with one required RCS relief valve inoperable.

the RCS relief valve must be restored to OPERABLE status within a Completion Time of 7 days.

Two RCS relief valves in any combination of the PORVS and the RHR suction relief valves are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.

The Completion Time considers that only ora of the RCS l

relief valves is required to mitigate an overpressure i

transient and that the likelihood of an active failure of the remMning valve path during this time period is very low.

(continued)

BYRON. UNITS 1 & 2 B 3.4-77 Revision A

ATTACHMENT B-2A PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, FOR FACILITY OPERATING LICENSES NPF-72 AND NPF-77 BRAIDWOOD NUCLEAR POWER STATION UNITS 1 AND 2 REVISED PAGES I

1-4 a

3/4 4-32 3/4 4-33 3/4 4-34 3/4 4-35 3/4 4-36 3/4 4-37 3/4 4-38*

3/4 4-39 3/4 4-40 3/4 4-40a J

B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-10 B 3/4 4-11 B 3/4 4-12 B 3/4 4-13 B 3/4 4-14 i

B 3/4 4-15 B 3/4 4-16 6-23

Page provided for continuity. No changes are proposed.

4 1

l 3Dil

("A DEFINITIONS SECTION p_gg 1.0 DEFINITI04 1.1 ACTI0N........................................................

j_3 1.2 ACTUATION LOGIC TEST..........................................

11 1.3 ANALOG CHANNEL OPERATIONAL TEST...............................

1_1 1.4 AXIAL FLUX DIFFERENCE.......................................

3_1 1.5 CHANNEL CALIBRAT10N...........................................

1-1

. ~ 1. 6 CHANNEL CHECK.................................................

1-1 1.7 CONTAllMENT INTEGRITY.........................................

1-2 l

1.8 CONTROLLED LEAKAGE............................................

1-2 1.9 CORE ALTERATION...............................................

1-2 1.9.a CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS...........................

1-2 1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................

1-2 i

1.11 DOSE EQUIVALENT I-131........................................

1-2 1.12 E-AVERAGE DISINTEGRATION ENERGY..............................

1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSETIME.....................

1-3 1.14 FREQUENCY NOTAT10N..................................,,,.....,

y3 1.15 IDENTIFIED LEAKAGE...........................................

1-3 1.15.a L,..........................................................

1-3 l

1.16 MASTER RELAY TEST............................................

1-3 l.17 MEMBER (S) 0F THE PUBLIC......................................

1-3 1.18 0FFSITE DOSE CALCULATION MANUAL..............................

1-4 1.19 OPERABLE - OPERABILITY.......................................

1-4 1.19.~a OPERATING LIMITS REP 0RT.....................................

1-4 1.20 OPERATIONAL MODE - M0DE..................................

1-4 1.20.a P,......................................................

1-4 1.21 PHYSICS TESTS................................................

1-4 1.22 PRESSURE BOUNDARY LEAKAGE....................................

+

1-4 1.23 PROCESS CONTROL PR0 GRAM......................................

1-5 1.24 PURGE - PURGING..............................................

1-5 l

1.25 QUADRANT POWER TILT RATI0....................................

1-5 1.26 RATED THERMAL P0WER..........................................

1-5 1.27 REACTOR TRIP. SYSTEM RESPONSE.TIME............................

'l-5 1.28 REPORTABLE EVENT..............................................

1-5

{/. 2 2. A. FAELW4dbe 7EA'NdfB4tl AM/75 AfAMT6424.).... (pye)

BRAIDWOOD - UNITS 1 1 2 I

Amendment No. 7/

r DEFINITIONS 0FFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alam/ Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. The 00CM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Sections 6.8.4.e and f, and (2) descriptions of the

'information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7.

OPERABLE - OPERABILITY J

1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATING LIMITS REPORT 1.19.a The OPERATING LIMITS REPORT is the unit-specific document that provides operating limits for the current operating reload cycle.

These cycle-specific operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9.

limits is addressed in individual specifications. Plant operation within these operating OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

2 1.20.a P (44.4 psi,) for the design basis loss of coolant accident.shall be the maxi g

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the core and related instrumentation:

in Chapter 14.0 of the FSAR, (1) described 50.59, or (3) otherwise approv(ed by the Commission.2) authorized under the provisio PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall zuse@, or vessel wall.

BRAIDWOOD - UNITS 1 & 2 1-4 AMENDMENT N0. 73

)

INSERT A PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.22.a The PTLR is the unit-specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, and the pressurizer power operated reliefvalve (PORV) lift settings'for the current reactor vessel fluence period.

These pressure and temperature limits shall be determined for each fluence period in accordance, with Specification 6.9.1.11. Unit operation within these limits is addressed in LCO 3.4.9.1, "Pressurefremperature Limits," and LCO 3.4.9.3, " Overpressure l

Protection Systems."

i I

i

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l

[

l P

a f

I I

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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS 4

g4

-REACTOR COOLANT SYSTEM MF" 4 4 N d Mm

&N wirQIiE4,uik Mw k*# k ? @ 1*>ueyM.L 3% P724.

LIMITING CONDITION FOR OPERATION

/

3.4.9.1(The eactor C lant System except the p ssurizer) t peratureane[

p essure sh 1 be lim ed in accor nce with the imit lines own on Fig es

/3.4-2a an 3.4-3a fo Unit 1 (Fig es 3.4-2b a 3.4-3b for it 2) duri

'heatup,

ooldown, iticality, a inservice ak and hydro tatic testi g with:

A maxi um heatup of 00'F in any -hour period b.

Am imum cooldow of 100*F in ny1-hourpe7iod,and c.

maximum temp ature change f less than r equal to 0*F in a y

-hour period uring inserv'ce hydrostat' and leak sting o ratio s above the he up and cool wn limit cur es.

APPLICABILITY:

At all times.

ACTION:

/n O A With any of the above limitt 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 Syster remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T,yg and pressure to less than 200'F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant 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 3 9h 4,4-A.-The-results-of-these-examinations = shal14eW-t: r; tt:

Ti;;re; 3.0-2; :nd 0.4-3; f;r Unit 1 (figuree 3.005 :nd 3.03b f;r Unit 2),

- M L ' 'a f:r Unit 1 (Fi;;r: 3. M b f;r Unit 2).

f P72'Q BRAIDWOOD - UNITS 1 & 2 3/4 4-32 AMENDMENT NO. 30

M i

MATRIAL PROPERTY BASIS I

i

' ~~ * ~

CONTROLLING MATERIAL:

CIRCUFERENTIAL WELD INITIAL RT,et:

4 0'F

'~

j T AFTER 32 EFPY:

1/4T, 159'F

~~

3/4T, 135'F l

Thes curves are applicable for heatup rates up to 00*F/hr f or i

the s vice period up to 32 EFPY and contain marg

ns of 10*F and 60- psi for possible instrument errors.

1500

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1 FIGURE 3.4-2a REACTOR COOLANT SYSTEM EEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1) i BRAIDWOOD - UNITS 1 & 2 3/4 4-33 AMENDMENT NO. 53

ib)ik CURVES APPLICABLE FOR HEATUP RATES UP TO 100*F/HR FOR THE 1 EFPY.

CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUN 25-

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FIGURE 3.4-2b REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)

BRAIDUOOD - UNITS 1 & 2 3/4 4-34 AMEN 0 MENT NO. 30

3

{"

MATERIAL PROPERTY BASIS r,

s ';

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CIRCUMFERENTIAL WELD INITIAL RT,,g:

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BRAIDWOOD - UNITS 1 & 2 3/4 4-35 AMENDMENT NO. 53

a l :

M CURVES APPLICABLE FOR C00LDOWN RATES UP TO 100*F/HR FOR.THE SE l

PERIOD UP i

TO 16 EFPY. CONTAINS MARGIN OF 10*F AND 60 PSIG FOR POSSIBLE STRUMENT ERRORS.

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BRAIDWOOD - UNITS 1 & 2 3/4 4-36 AMENDMENT NO. 30 i

i l

i

TA!

4.4-5 REACTOR VESSEL MATERIAL SURVEllLANCE PROGRAM - WITHDRAWAL SCHEDULE I

G S

UNIT-1 8

i I

e e

CAPSULE (ESSEL LEAD 5

DESIGNATION LOCAU ON FACTOR

. WITHDRAWAL TIME (EFPY)*

i U

58.5" 4.00 1.10 (Removed)

~

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4.02 4.50 V

61*

3.75 9.00 t

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3.

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p Z

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I g

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4.02 Standby l

l m

i E

Z 301.5*

4.02 Standby f

I iE 5

z

Withdrawal time spaj be modified to coincide with those refueling outages or reactor shutdowns most closely a roaching the withdrawal schedule.

w t

i

. =

l REACTOR COOLANT SYSTEM I

PRESSURIZER l-LIMITING CONDITION FOR OPERATION l

3.4.9.2 The pressurizer temperature shall be limited to:

~

A m'aximum heatup of 100 F in any 1-hour period, a.

i b.

A maximum cooldown of 200 F in any 1-hour period, and i

A maximum spray water temperature differential of 320 F.

c.

APPLICABILITY:

At all times.

I ACTION:

I With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next b hours and reduce the pressurizer pressure to less'than 600 psig i

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

SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per.30 minutes during system heatup or cooldown.

The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

i BRAIDWOOD - UNITS 1 & 2 3/4 4-38

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION i

)

3.4.9.3 At least two overpressure protection devices shall be OPERABLE, and

.each device shall be either:

3 i

a.

A residual heat removal (RHR) suction relief valve with a lift setting of 1ess than or equal to 450 psig, or e

b.

A power operated relief valve (PORV) with a lift setpoint that i

varies with RCS temperature which does not exceed the limit i

established in F her: ' '- '- Wit 1 (Fi; r: 2.0 ib f;; ". it G.

ht* Mif)

APPLICABILITY: MODES 4, 5, and 5 witTn the reactor vessel head on, i

j 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 l

through at least a 2 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

s b.

With one of the two required overpressure protection devices i

inoperable in MODES 5 or 6, restore two overpressure protection devices to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or vent the RCS through l

at least a 2 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

I c.

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

I 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; etherwise, verify the vent pathway every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

In the event either the PORVs, RHR suction relief valves, or the RCS e.

~

vents are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to i

Specification 6.9.2 within 30 days.

The report shall describe the i

circumstances initiating the transient, the effect of the PORVs, RHR suction relief valves, or RCS vents on the transient, and any i

corrective action necessary to prevent recurrence.

l f.

The provisions of Specification 3.0.4 are not applicable.

1 l

1 i

i i

BRAIDWOOD - UNITS I & 2 3/4 4-39 AMENDMENT NO. 53 i

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6 110 160 210 260 310 60 T(RTD), LOWEST COMS RTD Temperature (OF)

FIGURE'3.4-4a NOMINAL PORY PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE UP TO 16 EFPY (UNIT 1)

BRAIDWOOD - UNITS 1 & 2 3/4 4-40 AMENDMENT NO. 64

~

IfttfirD PolWT! tm Wr curyr -5 g

b k

[i 300 70 505 N

=~

N 505 5

150.

522 240 600

--E 314 700

=

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FIGURE 3.4-4b NOMINAL PORY PRES 8URE RELIEF SETPOINT VERSUS RC8 TEMPERATURE FOR THE COLD OVERPFE880RR PROTECTION 8 STFN APPLICABLE UP TO 16 EFPY (UNIT 2)

.7 BRAIDWOOD - UNITS 1 & 2 3/4 4-40a AMENDMENT NO. 53

1 REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) i 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.

DeleAt..

bpke 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are

. g 1

ited to be consistent with the requirements given in the ASME Boiler and Pres re Vessel Code,Section III, Appendix G:

5 1.

The reactor coolant temperature and pressure and system hea p and oldown rates (with the exception of the pressurizer) sh 1 be 1

ited in accordance with Figures 3.4-2a (3.4-2b) and

.4-3a (3.4-3b) t for e service period specified thereon:

1 a.

Al able combinations of pressure and tempe ture for specific tempe ture. change rates are below and to e right of the J

limit es shown.

Limit lines for cool wn rates between

~

J those pre nted may be obtained by int polation; and b.

Figures 3.4-(3.4-2b) and 3.4-3a 3.4-3b) define limits to assure prevent 1 of non-ductile ailure only.

For normal operation, other herent plant characteristics, e.g., pump 1

heat addition and p ssurize heater capacity, may limit the heatup and cooldown r es at can be achieved over certain pressure-temperature ra s.

2.

These limit lines shall b calcu ted periodically using methods provided below.

3.

The secondary side the steam genera r must not be pressurized above 200 psig if he temperature of the team generator is below i

70*F, 4.

The pressur er heatup and cooldown rates shal not exceed 100*F/hr and 200*

hr respectively.

The spray shall not e used if the tempera re difference between the pressurizer and he spray fluid is gr ter than 320*F, and 5.

S tem preservice hydrotests and in-service leak and hyd tests

' hall be performed at pressures in accordance with the req ' ements c

of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the rea or sel are determined in accordance with the 1973 Summer Addenda to Section f the ASME Boiler and Prest,ure Vessel and Code.

AMENDMENTNO.k BRAIDWOOD - UNITS 1 & 2 B 3/4 4-7

REACTOR COOLANT SYSTEM j

4 BASES PRESSURE / TEMPERATURE LIMITS (Continued) m Heatup and cooldowilimit curves are calculated using the most limiting va of the nil-ductility reference temper ture, RT

, at the end of 32 l

j effe ve full power years for Unit 1.(16 effective ILIll power ye s for. Unit 1

2) of se ice life. The 32 EFPY for Un 1 (16 EFPY for Unit 2 service life l

period is sen such that the limiti RT at the 1/4T loca on in the core 1

region is gr er than the RT of elimItingunirradiate material. The selection of s a limiting k ssures that all compon ts in the Reactor I

Coolant System wi be operated onservatively in accor nce with applicable j

Code requirements.

The reactor vessel terials have been teste to determine their initial

^

RT ; the results of these ests are shown in T le B 3/4.4-la for Unit 1 l

(TEleB3/4.4-bforUnit2 Reactor opera on and resultant fast neutron l

(E greater-th 1 MeV) irradia on can caus an increase in the RT T

'e or.

fore, an ad' sted reference tempe ature, stion, can be predicted usinbased upon thj j

and nicke content of the material que i

Figure

/4.4-1 and the largest valu f ART., computed by either R ulatory i

Guide Y.99, Revision 2, " Radiation E r tiement of Reactor Vessel aterials" or thf Westinghouse Copper Trend ves s wn in Figure B 3/4.4-2 The heatup and/cooldown limit curves of Fig s 3.4-2a 3.4-2b) and 3.4-3a 3.4-3b) i j

jilude predicted adjustments this shift at the en of 32 EFPY for RT,7orpossibl pnit 1 (16 EFPY for Unit'2) well as adjustme errors in the 4

pressure and temperature ing instruments.

Values of ART ed until the results 1

from the material., determined in this manner may be surv.e111ance program, evaluated accord to ASTM E185, are

{

l Capsule !s dill be removed in accordance ith the equirements of available.

ASTM E185-73 and 19 CFR Part 50, Appendix H.

Th surveillanc specimen with-drawal schedule ( shown in Table 4.4-5.

The 1 ad factor repre nts the rela-tionship betwe the fast neutron flux densit at the location o he capsule i

j and the inner all of the reactor vessel.

erefore, the results o tained from i

the surveil) nce specimens can be used to redict the future radiatio damage j

3 to the regtor vessel material by using he lead factor and the withdra 1 time i

of the ca'psule.

The heatup and cooldow curves must be recalculated when he ART,, determined from the surveillanc capsule exceeds the calculated ART.,

j for e equivalent capsule radiation xposure.

i I

t i

4 l

W BRAIDWOOD - UNITS 1 & 2 B 3/4 4-8 AMENDMENT NO. 53

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

Allowable pressure-mpera nte relationships for various heatup and cooldown rates ne cal lated using methpds derived f, rom Appendix G in Section III of the E Boiler.and PreM ure Vessel f

de as requ, red by Appendix G to 10 C Part 50, and th e methods e discussed n detail.in WCAP-7924-A, "B is for Heatup and ooldown Li Curves,"

ril 1975.

The ge ral method for ca ulating he p and cool wn limit rves is j

based upo he principles of he linear astic fract e mechani (LEFM) technolo In the calcul ion proced es a semi e iptical s face defect with a epth of one quar r of the w thickness

, and a 1 ngth of 3/2T is assu d to exist at th inside of t) vessel wa as well af at the outside of th vessel wal'..

.Th dimensions pf this post ated crack / referred to in

/ppendixGofASME ection III af the refer ce flaw, a ly exceed the cut t

t capabilities of ' service in tion techn~ ques.

Ther fore, the reactor tion limit cur developed r this reference crac are conservative a provide era-sufficient s ty margins r protecti against n -ductile failure./To assure that the raf ation embri lement eff ts are act nted for in the calculation of the li ft curves, t most limitring value o the nil-ductility eference tempera re, RTNDT, i used and tVis includes the radiation-ind ed shift, ART

, correspond g to the e of the pe od for which hea p and cooldown N

cu es are gener ted.

The ASM approachfprcalculatpi the allowable 1,imit curves for vari heatupand/ooldownrap4sspecifiepthatthetotalstressintensityfac K, for t combined /hermaland ressure stresses at any time during tup

{

y or cool own cannot be greater an the referent stres's intensity f or, K 7

IR' for e metal te perature a that tinie.

K s obtained from th reference y

fracture toug ess curve, efined in Appe ix G to the ASME C e.

The K

/-

IR curve is gi 'en by the uation:

V,y = 26.78 +. 223 exp [0.0145(T-RTNDT + 160)]

(1)

Where.

K is t reference stress intensity factor s a function of he metal i

temperatur T a the metal n ductility referenc temperature RT

Thus, T.

t e governin equation for e heatup cooldown alysis is defin in Appendix G of the ASM Code as follow :

C g+Kyg i Ky (2)

's co K

/;

IM=thesyressintensityfactorcausedbymemb ne (pressure)

stress,

=the/

KIt stress intensity fac or caused by he thermal gradients, BRAIDWOOD - UNITS 1 & 2 B 3/4 4-9 AMENDMENT NO 30 1

I 4

_ REACTOR COOLANT SYSTEM

. BASES PRESSURE / TEMPERATURE LIMITS (Continued)

~

K IR = to the Rconstant pr vided by the code as a function DT of the materi r, f temperature relative j

C=

2.0 or level A and B ervice limit, and C=

1.5 for.inservic ydrostatic leak tes operatio any time during e heatup or c Idown tra ient, K s determined by

{

th etal temperature t the tip of e postul d flaw, t appropriate value r RTNDT, and thej ference frac e toughn s curve.

ffe thermal stresses f

resulting from t perature gr i nts thro the vess wall are calcul.ated and then the responding mal stre intensity actor, Kyg, for,the reference f) w is comput factors eobtainedan-)e From Eq tion (2) t e pressure stressdntensity

, from the

, the allo ble pressures are' calculated.

COOLDDWN

/

./

For the ca ulation of the allowa e pressure versps/coolanttempe during cooldoyn, the Cod (reference ture the vessel wa11.

Duri ' cooldown, aw is assumed to exist at the inside of always at e controlling lo' cation of the, flaw is stresses,tfie inside the wall because the therroll gradients produce tensile pressupe/at the ins j

which increase with incre'asing cooldowryrates.

e

-temperatdre re,lationy'are generated fir both steady State and fi Allowable cooldown rate situations.

fromtheserelations, composite Imit curves are cors'tructed each cooldown rate of int est.

Theyeofthecopp/osite curve i control,of the cooldown procedure is he cooldown apelysis is necessary bec/ause temperature, whereas'the limiting ased on measurement of reactor coolant temperature at the'tip of the assuined flaw. essure is actually dependent on tb(' material loca' tion is at a' higher tem During cooldown, the 1/4T vessel Tti'is conditior)[of course, peratdre than the flu'id adjacent to ti)e' vessel ID.

i follows thatrat any given reac, dot true for the' steady-state situation.

is f

It cooldown res,ults in a highet value of Ktor coolant temperature, the AT 6eveloped durin f

cooldowng'ates than for

't the 1/4T location for finite j

f exist,Soch that the inpfease in Keady state op,eration.

Furtherm e, if condition IR e eeds kit, the c 'culated allowable pressure during cool own w ll be gre ter than the st p

dy state val

/

I The above precedures are needed because therp is no direct be violated if/the rate of cooling is decreasep'at vari ntrol on y

k cooldown ramp.

The use of thf composite curve eliminates phis problem and I

assures conservative operation of the system for the entire cooldown period.

BRAIDWOOD - UNITS 1 & 2 8 3/4 4-10

m g

G TABLE 8 3/4.4-la REACTOR VESSEL TOUGHNESS O

(UNIT 1)

E Q

\\

Average T

Shelf Energy I

MATERIAL Cu P

NDT NDT

.NMWD*

MWD **

COMPONENT Heat No.

SPEC.

F F"

ft-lbs ft-lbs Closure Head Dome 01398-1 A5338, C1. 1 ~D6

.009

-30

-30 129 N

Closure Head Ring 49C1126-1-1 508, Cl. 3

.02

.009

-20 123 Closure Head Flange 2030-V-1 A

, C1. 2

.11

.009

-20

-20 163 Vessel Flange 122N357VA1

A508,

1. 2

.010

-10

-10 106 Inlet Nozzle 21-3257 A508, C. 2

.09

.00

-20

-20 144 Inlet Nozzle 21-3257 A508, C1.

.09 0

-10

-10 144 Inlet Nozzle 22-3313 A508, C1. 2

.07

.008

-10

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

.0

.010 0

0 115 m

Outlet Nozzle 22-3025 A508, C1. 2

.013

-10

-10 125 m}

Outlet Nozzle 4-3329 A508, C1. 2

.08 009

-20 156 Outlet Nozzle 4-3383 A508, C1

.08

.DO

-20 3

.00(7 20

, 147 4

Outlet Nozzle 11-5226

A508,

-.2

.09

-10 Nozzle Shell SP7016 A5

, C1. 2

.04

.008 10 10 155 10 125 i

Upper Shell***

49D383/

08, C1. 3

.05

.008(.73)

-30

-30 122 173 l

[

49C344-1-1 Lower Shell***

490867/

A508, C1. 3

.03

.007(.73)

-20

-20 135 151 l

[

i 49C81)

-1 Bottom Head Ring 493148-1-1 A508, C1. 3

.05

.008

-50 40 147

~

t Bottom Head Dome FA882-1 A533B, C1. 1.14

.010

-20

-20 123

'l Upper Shell to***

/WF-562

.04

.015(.67) 40 40 80 l

l Lower Shell Girt (eld Weld HAZ 4

-70

<-10 5(

h

Nor to major working direction.

N N

g Jor working direction.

Calculations per Regulatory Guide 1.99 Revision 2 use the Nickel content shown in parentheses.

-4 l

t I

..m

.. m m

~ -..

mg

. TABLE B 3/4.4-1b h

REACTOR VESSEL TOUGHNESS (UNIT 2)

E Average Z

T RT Shelf Energy MATERIAL Cu P

NOT NOT / NMWD" MWD""

COMPONENT HEAT NO.

SPEC.

F F*

/ ft-lbs ft-lbs Closure Head Dome B9754-1 45338, C1. 1 ~15

.D05

-60

-0 151 l

Closure Head Ring 50C478-1-1 A5

, C1. 3

.05

.006

-30 0

128 N

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

.009 20 20 135 i

Vessel Flange 124P455 A508, Ch2

.07

.010 20 20 128 Inlet Nozzle 41-5414 A508,C1.2%07

.008

-10

-10 137 Inlet Nozzle 41-5414 A508, C1. 2

.0

.009

'10

-10 140 Inlet Nozzle 42-5417 A508, C1. 2

.09

.011

-10

-10 122 Inlet Nozzle 42-5417 A508. C1. 2

.09

.009

-10

-10 116 Outlet Nozzle 4-3502 A508, C1. 2

.09

.91

-10

-10 155 Outlet Nozzle 11-5226 A508, C1. 2

.09

.00

-10

-10 116 m

} Outlet Nozzle 4-3481 A508, C1. 2

.07 f.008

-10

-10 163 Outlet Nozzle 11-5266 A508, C1. 2

.09

.010 10 10 117 U'

Nozzle Shell SP7056 A508, C1. 2

. 0.4,/

.005 0

30 115 Upper Shell***

490963/

A508, C1. 3

.63

.007(.71)- -3D

-30 119 147 49C904-1-1 Lower Shell***

500102/

A508, C /3

.06

.006(.75)

-30

-30 144 168 l

50C97-1-1 Bottom Head Ring 4901066-1-1 A50, C1. 3

.07

.008

-30

-30 156 l

Bottom Head Dome 01429-1 A$33B, C1. 1.11

.010

-20

-20 120 l

i

- Upper Shell to***

WF-562

/

.04

.015(.67) 40 40 80 Lower Shell Girth Weld

/

l l

Weld HAZ

-30

-30 145 i

g

=

's i

E N

T

Normal to inajor working direction.

\\

--4

- Major yofking direction.

N L

zP U *** Calculations per Regulatory Guide 1.99 Revision 2 use Nickel content shown in parentheses s

I L

t

~... - -. -

.. ~,.

s

/

1 4

1

-5 :. : :- 5. f.v==:r:_m = L =: ~ 1 : a :: c -= =~.= :.: g --- _ _1.

t----

2

-. - _ l f

I

/

-,- s 018

\\

M J). W.hi-i: 0 + 0.F A -77 /. f-YD1/4T.1.2 x 1018 N/cri,2 e

~

~'

f

,m

~ ~

=- f.

u

. -~

=,: r.

. _. :/ '. **.

m.. _= j..; _ _.

g.-...=.-_=_.

_--. 1_: _

_.-_..f

. +

.m

.-t--

_... f :-- _,

_,.1

> 'Udu=

1:

i-.d.i = tw.tsh

.-%=-ht-M-;=

ia +=L '. : --.

C =+ iM

/ a stas -t

- i-ArN=5 #.

1'; =- 3

. :.:...._v

'.. a

==

gz

-i=:;-f25 r.,-

B

. _ =..-L k:/

=.=_..g=.__.==. _. ~. _ -.

w.-

y

--. fq _5. = - -

, m._ -

w

.--._.-/.,-

2 u

w

- m -

A 3

2

.d

...j J.

3 u.

.r..

3/4T 2.7,x 10 '8,N/cm.

2

~ _.

.- l. __.4 -

2

..f r

. v

_ _. i

/

.i 9

3;

==*

W

-2 1088 ~t.

f/

g l

lrt t, f~#*.

ier he r s 's*rr- 'a.s: 33 t:_---

r e itt*_g-i e

j ' ' I --- J /: 4 M f v3

- M. F 4 i : *M Na-E q-py { -r74 e 8

?

~_-l e !-I : / 4.H i h [.45- : f-W I#M f-I-.Z E-- - 'tW41-1 i r

s y u :... / j...

_ :. s. :.: :

...:.:t._ ;;.. =--

c

::=

O

...~~....._~,_--m..__.:.

g

-...g..

3..

f f E _ - =. --

a l-

fA ';

- s. { & r...t '. E s...1. j. h

' bt.a + K L.. p a:. z f or r N-a.1_;a t

f

.t+

i4: H.2d.=. -..EJ =2L :- ) i M.E E-M i. i=i I - = - -*-:-

h4 } :E!.3.-:=*.~Et+5'.t.=Li * -i--5 45EM~5 ';

D " =i

.=?]. = w..: . a t -

= :. L.= :--

.w.:-,

= : =...._:.=

3lg

=.

=-f g.

2

-.g_

W f

1 I

\\

i 4

aN

~

e g.

10U-

., i.x.. -- *

~

0 10 20 30 40 50 l

i EFFECTIVE FULL POWER (Years) l

)

l FIGURE B 3/4.4 FAST NEUTRON FLUENCE (E > 1 MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE BRAIDWOOD - UNITS 1 & 2 B 3/4 4-13

~.

E 2:

o c5 O

/

l l 11 I l l Ill(llllibilililiihiN Illilill!Illi.l;ll.(,

j I

i i i :]{

~H [ il i1 UPPER LIMIT 10.30% COPPER 8 ASE. 0.25% WELD 4

~

y s.

g g:%

i h.,

- :l

~

'0.25% COPPER Sf. 0.20% WELD %. k t FM i

~

w%

hs Io"TIl[ $

a n

a w

s;. w t

U :

a.'

v,,

2n 1

i ia I "J.

.$.2 1lp _... %,,,

.- r t,

.. as,,,.

s

.77 9

.p e

200 p.

1 g ;

i gad]p

%g

",ili {,,,=f'"

_~E Ei i

g,,,,,

Ui

~

',1'".b%

4 T

I:II.'."

,l Fil r

c

~~

IIEh

},.

g

- g% ]

}hp,j,[ri,t'..h

!',,;;..il o

e"

""I ft i

t:

"di l

.il li

-v 7 g i., i[ig n_.y,,.

i.N LOWER LIMIT a

&T,I t

E

i P r ?

yY'd !:I i

e 2

y;,,,,

,,e i

ii W

E Ii i 33.**fTI !!'i ll t Wd

_j;g_,

a Gi Di liii l li hl I

s" l

,+ hi! i!H '.').*='TI; '!I 'M#M.,, -

. i

!I I IU U!! i.E 'd

, [ j IQQ tu y

hY,I i: UEO"7

!i'.ii[@(-

. ili

' li Ils I5!!If5 !!i$ s:ii Ill!III L

' I' ii i

p"tEi Fij Il! IIi ff$,Nihf M

ii: 0.20% COPPER SE. O.15% WELD ~l!

i 60 E'

f j;

u N:g::

g, :, ;g 1

,,.hd,,,,,-7 3;

7

a

... 3 u

il

,:,,.3 7

t a, t- ::I

. tr 1

0.15% COPPER BAS. 10% WELD :;.

,,, )ct""'E

..Il 1

II I

t 0.10% COPPER BASE. O.05 WELD

...I..h $

! f! N

!;i.iiW

!2 ill !!H!f

.l___1 I h h N'

_ l d

_9 h

!!!! I i W

i

!!I iti i eli I

ii

. li iig ::,i p

't I til

!:- I,pe:,;ij 1

i I

i jp i(d

" [l

'd

1:l i '-

i i

i;

.,, :r 4 i

i

[

i:

t'l

,1

'T i., a.j ap

.i.i I-i 4:

s i

i ($ $ng,,:

i dl!S !!i..i.

i l E nb I

i i

.j

.. l: ill!,

y:

20

,g.

.,,i.

,o=

FAST NEUTRON FLUENCE IN/CM*, E >1 MeV)

FIGURE B 3/4.4-2

/

EFFECT OF FLUENCE AND COPPER,0N SHIFT OF RT FOR HDI REACTORVESSELSTEELSEXPOSEDTbIRRADIATIONAT550'F

1 REACTOR COOLANT SYSTEM 4

BASES PRESSURE / TEMPERATURE LIMITS (Continued)

I O

A notch in the oldown curve of Figure 3.4-3a for Unit 1 (3.4-3b for Unit 2) may be pre nt due to the added constrai t' on the vessel closure flange given.in Appendi G of 10 CFR.50./This constr nt requires that, at pressures greater than 2 of the preservYe system by ostatic test 'ressure, theeflange regions that re highly stres d by the bo preload mus exceed the j

of f

NDT the mater 1 by at least 1 F.

The fla ge RTNDT + 12 F may impi e on the cooldow curves and ther ore the notch is required If no not is present, f

this dicates that th vessel clos e flange reg' n has been etermined to be t limiting.

REATUP

/

Three se for finite tyeatup rates. rate calcula) ons are regtdred to deteymine the limit rves s is done i he cooldowp' analysis, all ble pressure-tpmperature rel ionships ar developed pf r steady-stat nditions as well ps finite heat rate condi ons assuming the presence a 1/4T defect t the inside the vessel wall.

The tfiermal gradientf during heatup

{

produp compressiv duced by ipternal presspr/the wall that/11eviate the tresses at) e inside of tenplestresses e

The meta cfr ck tip lags tp'e coolant typerature; therefore, the K ytemperature a IR j4uring heatup ts lower tha(the K fr he 1/4T crac,k'during steady-state p

IR conditions at the same c lant temper ture.

end'ofthe/ transient,ponditionsmay'existsucht@ttheeffectsoDuringMeatup ompressive fresses and different K

's for steady-state and finite atup rates l

thermal do not ffset eacty ther and th pressure-tempe/rature curve ba on steady-state cond' ions no longer represenpf a lower bound of all similar rves for finite hea up rates w 'n the 1/4T f, law is consideyed.

Therefore, h cases have to b/analyzedi order to asp 0re that at a coolant tenperature the lower value

/ftheallo ble pressure / calculated fo steady-stateanp'finiteheatuprates is obtained'.

/

Tt)/second porti n of the heatup analysis concer/

ns the calculation f pressure-temperatur,e' limitations fdr the case in w (ch a 1/4T deep ou) ide the!'hermal gradi'ents established at the outside surface during he surf ce flaw is assumed.

Unlikefhesituationa he vessel inside surface, t

stfesseswhicharetensilein/atureandthusJ'e'ndtoreinforce f

pressure

!(tresses pre nt.

These thermal stresses, of course, are depe ent on both the rate of eatup and the me (or coolant emperature) alo the heatup ramp.

Fupthermore, since he thermal stre es, at the outs' e are tensile and increase /with increasin eatup rate, a 1 er bound curve annot be defined.

Rather each heatup ra of interest mus be analyzed on n individual basis.

Following the ge ration of pressyre-temperature urves for both the steady-state and fin 1te heatup rate sAtuations, the nal limit ~ curves are produced as follows.

A composite curve is constructed based on a point-by point k tf N

~

BRAIDWOOD - UNITS 1 & 2 B 3/4 4-15 AMENDMENT NO. 30

\\

J i

REACTOR COOLANT SYSTEM 1

BASES j

PRESSURE / TEMPERATURE LIMITS (Continued) comparison of the steady-state and finite heatup rate data.

At any given [

mperature, the allowable pressure is taken to be the lesser of the threef i

es taken from the curves under consideration.

va 4

Th use of the composite curve is necessary to set conservativ 4eatup i

limitatio because it is possible for conditions to exist'such t over the course'of t heatup ramp the controlling condition switches frpef the inside to the outside nd the pressure limit must at all times be based on analysis of the most crit al criterion.

f j.

Finally, the c site curves for the heatup rate ta and the cooldown rate data are adjusted r possible errors in the p sure and temperature sensing instruments by th values indicated on the espective curves.

l Although the pressurizer erstes in temperature ranges above those for which there is reason for concer of nonductile failure, operating limits i

are provided to assure compatibili of ope' ration with the fatigue analysis performed in accordance with the ASM ode requirements.

i 1

i The OPERABILITY of two PORVs,1ve, or a R suction relief valves, or o Ar two i

PORV and one RHR suction relief a CS vent opening of at least i

2 square inches ensures that RCS will be p tected from pressure transients l

which could exceed the limits of Appendix G to 1 CFR Part 50 when one or more l

of the RCS cold legs are }ess than or equal to 350 Either PORV has adequate relieving capability to protect the RCS from overpres rization when the tran-cient is limited to der: (1) the start nf an idla R. with the eacen % ~

j water temperat'ure the st'eam generator le'ss than or equ to 50*F above the j.

PrA cold leg t atures.or (2) +5e start af a centrifuge. rherging pu :p end t

its injection ' o a water solid RCS.

These wo scenarios are analyzed to determine the resulting o shoots assumin single PORV actuation with a stroke time of 2.0 seconds f full j

close o full open.

Figure 3.4-4a (3.4-4b) are based upon this analys and i

rep sents the maximum allowable PORV variable setpoint such that, for t two o

rpressurization transients noted, the resulting pressure will not exceed i

pendix G reactor vessel NDT limits.

2 j

3/4.4.10 STRUCTURAL INTEGRITY f

i The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the i

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 i

i 10 CFR 50.55a(g) except where specific written relief has been granted by the j

Commission pursuant to 10 CFR 50.55a(g)(6)(1).

}

BRAIDWOOD - UNITS 1 & 2 B 3/4 4-16 AMEN 0MENTNO.k.N i.

4 INSERT B

[

All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. LCO 3.4.9.1, i

" Pressure / Temperature Limits," limits the pressure and temperature changes during RCS j

heatup and cooldown to within the design assumptions and the stress limits for cyclic l

operation.

1 The PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) contains pressure and temperature (P/T) limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate ofchange of reactor coolant temperature.

Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to detennine that operation is within the allowable region.

LCO 3.4.9.1 establishes operating limits that provide a margin to non-ductile failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB). The vessel is the component most subject to non-ductile failure, and the LCO limits apply to the entire RCS, except the pressurizer, which has differerit design characteristics and operating functions.

j 10 CFR 50, Appendix G, requires the establishment of P/T limits for specific material fracture toughness requirements of the RCPB materials. Appendix G of 10 CFR 50 requires an adequate margin to non-ductile failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use.of the American Society of Mechanical Engineers (ASME) Code,Section XI, Appendix G.

The neutron embrittlement effect on the material toughness is reflected by increasing the Nil Ductility Reference Temperature (RTer) as exposure to neutron fluence increases.

The actual shift in the RTwor of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 and Appendix H of 10 CFR 50. The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99, Revision 2.

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive.

At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, I

-.. - -. --. -~

y INSERT B (continued) 1:

different locations are more restrictive, and, thus, the curves are composites of the most

{

restrictive regions.,

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal j

- gradient reversal alters the location of the tensile stress between the outer and inner walls l

during heatup and cooldown, respectively.

i a

i The criticality limit curve includes the 10 CFR 50, Appendix G requirement that it be it 40 F l

above the heatup curve or the cooldown curve, and not less than the minimum permissible

[;,

temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.1.1.4, " Minimum Temperature for Criticality."

)

The consequence of violating the LCO limits is that the RCS has been operated under j

conditions that can result in non-ductile failure of the RCPB, possibly leading to a nonisolable j.

leak or loss-of-coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. The i

[

ASME Code,Section XI, Appendix E, provides a recommended methodology for evaluating

{

an operating event that causes an excursion outside the limits.

l OVERPRESSURE PROTECTION SYSTEMS i

The Low Temperature Overpressure Protection (LTOP) System controls the RCS pressure at low temperatures so the integrity of the RCPB is not compromised by violating the P/T limits 3

I of 10 CFR 50, Appendix G. The reactor vesselis the limiting RCPB component for i

demonstrating such protection. The PTLR provides the maximum allowable actuation logic j

setpoints for the pressurizer Power-Operated Relief Valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the 10 CFR 50, Appendix G requirements during the MODES in which the LTOP system is necessary.

i The reactor vessel material is less ductile at low temperatures than at normal operating i~

temperature. As the vessel neutron exposure accumulates, the material toughness decreases i-and the material becomes less resistant to stress at low temperatures..RCS pressure, l

therefore, is maintained low at low temperatures and is increased only within the limits j

specified in the PTLR.

The potential for vessel overpressurization is most acute when the RCS is water solid, i

t occurring only during shutdown; a pressure fluctuation can occur more quickly than an j

operator can react to relieve the condition. Exceeding the RCS P/T limits by a significant j.

.b 1

e r

i INSERT B (continued) i amount could cause non-ductile failure of the reactor vessel. LCO 3.4.9.1,

" Pressure / Temperature Limits," requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the PTLR limits.

l LCO 3.4.9.3, " Overpressure Protection Systems," provides RCS overpressure protection by i

having a minimum coolant input capability and having adequate pressure relief capacity.

j Limiting coolant input capability requires all Safety Injection (SI) pumps and all but one charging pump (a centrifugal charging pump) incapable ofinjection into the RCS and isolation of the SI accumulators. The pressure relief capacity requires either two redundant RCS relief valves, or a depressurized RCS and an RCS vent of sufficient size. One RCS relief valve or the open RCS vent is the overpressure protection device that acts to terminate an increasing pressure event.

1 With minimum coolant input capability, the ability to provide core coolant addition is restricted. The LCO does not require the makeup control system deactivated or the SI actuation circuits blocked. Due to the lower pressures in the LTOP MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the l

makeup control valve. If conditions require the use of more than one centrifugal charging 1

pump for makeup in the event ofloss ofinventory, then pumps can be made available through manual actions.

The LTOP System for pressure relief consists of two PORVs with reduced lift settings, or two Residual Heat Removal (RHR) suction relief valves, or one PORV and one RHR suction relief valve, or a depressurized RCS and a RCS vent ofsufficient size. Two RCS relief valves are l

required for redundancy. One RCS relief valve has adequate relieving capability to prevent j

overpressurization for the required coolant input capability.

PORV Reauirements As designed for the LTOP System, each PORV is signaled to open if the RCS pressure approaches a limit determined by the LTOP actuation logic. The LTOP actuation logic i

monitors both RCS temperature and RCS pressure and determines when a condition is approaching the PTLR limits. The wide range RCS temperature indications are auctioneered j

to select the lowest temperature signal.

The lowest temperature signal is processed through a function generator that calculates a pressure limit for that temperature. The calculated pressure limit is then compared with the indicated RCS pressure from a wide range pressure channel. If the indicated pressure meets or exceeds the calculated value, a PORV is signaled to open.

3

~ -.

l INSERT B (continued) l The PTLR presents the PORV setpoints for the LTOP system. The setpoints are normally staggered so only one valve opens during a low temperature overpressure transient. Having the setpoints ofboth valves within the limits in the PTLR ensures that the 10 CFR 50, Appendix G limits will not be exceeded in any analyzed event.

When a PORV is opened in an increasing pressure transient, the subsequent reliefwill cause

[

l the pressure increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close. The pressure continues to decrease below the reset pressure as the valve closes.

RHR Suction Rt 'ief Valve Reauirements During the LTOP MODES, the RHR System is operated for decay heat removal and low i

- pressure letdown control. Therefore, the RHR suction isolation valves are open in the piping from the RCS hot legs to the inlets of the RHR pumps. While these valves are open, the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS.

The RHR suction isolation valves must be open to make the RHR suction relief valves OPERABLE for RCS overpressure mitigation. The RHR suction relief valves are spring loaded, bellows-type relief valves with pressure tolerances and accumulation limits established by Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for Class 2 relief valves.

RCS Vent Reauirements Once P 'CS is depressurized, a vent exposed to the containment atmosphere will maintain thel

- containment ambient pressure in an RCS overpressure transient, if the relieving requireuents of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass or heat input transient, and maintaining pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.

For an RCS vent to meet the flow capacity requirement, it requires removing a pressurizer safety valye; removing a PORV's intemals, and disabling its block valve in the open position; or similarly establishing any comparable veat. The vent path (s) must be above the level of reactor coolant, so as not to drain the RCS when open.

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ADMINISTRATIVE CONTROLS CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS 6.9.1.10 Fuel enrichment limits for storage shall be established and documented in the CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL l

STORAGE RACKS.

The analytical methods used to determine the maximum fuel enrichments shall be those previously reviewed and approved by the NRC in

" CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS The fuel enrichment limits for storage shall be determined so that all applicable limits (e.g., subcriticality) of the safety analysis are met.

The CRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE i

OSt(t ACKS report shall be provided upon issuance of any changes, to the NRC l

Document Control Desk, with copies to the Regional Administrator and the C

Resident Inspector.

4 SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least 5 years:

Records and logs of unit operation covering time interval at each a.

power level; j

b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety; c.

All REPORTABLE EVENTS; d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications; Records of changes made to the procedures required by e.

Specification 6.8; f.

Records of radioactive shipments; Records of sealed source and fission detector leak tests and results; g.

and h.

Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the unit Operating License:

Records and drawing changes reflecting unit design modifications a.

made to systems and equipment described in the Final Safety Analysis Report; b.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories; BRAIDWOOD - UNITS 1 & 2 6-23 AMENDMENT NO. %

INSERT C Ragetor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.11 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, as well as heatup and cooldown rates and pressurizer power-operated relief valve lift settings, shall be established and documented in the

)

PTLR for the following: '

1)

LCO 3.4.9.1, " Pressure / Temperature Limits," and 2)

LCO 3.4.9.3, " Overpressure Protection Systems."

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP-14040 NP-A, Revision 2, January 1996," Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," as approved by the NRC by letter dated October 16,1995 (TAC M91749); and NRC Safety i

Evaluation Report dated _

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

i i

4 k

i ATTAClIMENT B-2B j

PROPOSED CHANGES TO DRAFT IMPROVED TECHNICAL SPECIFICATIONS, 4

FOR FACILITY OPERATING LICENSES NPF-72 AND NPF-77 BRAIDWOOD NUCLEAR POWER STATION UNITS 1 AND 2 REVISED PAGES 5.0-43 B 3.4-10 B 3.4-15 B 3.4-17 B 3.4-67 B 3.4-75 B 3.4-77 i

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) a.

RCS pressure and temperature limits for heat up. cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates, and Power Operated Relief Valve (PORV) lift settings shall be established and documented in the PTLR for the following:

LC0 3.4.3. "RCS Pressure and Temperature (P/T) Limits"; and LCO 3.4.12. " Low Temperature Overpressure Protection."

b.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC. specifically those described in the following documents:

up-A 1.

WCAP-14040-4 " Methodology used to Develop Cold Overpressure Mitigating System setpoints and RCS Heatua and Cooldown Limit Curves." Revision s.2 Jag 1996 Ccc;.;acr 1004. Approval provided by NRC Safety Evaluation Report, dated October 16,1995;ag 2.

Cc-Ed lette to NRC. " Initial Pressure and Ted crcture ti=1ts Re: Ort." dated fpprov provided :y NRC Safety Evaluation Report, dated c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.7 Post Accident Monitorina Reoort When a report is required by Condition C or H of LC0 3.3.3 " Post Accident Monitoring (PAM) Instrumentation." a report shall be submitted within the following 14 days. The report shall outline the pre)lanned alternate method of monitoring, the cause of the inopera)ility, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

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i (continued)

BRAIDWOOD - UNITS 1 & 2 5.0-43 Revision A

- - =..

RCS P/T Limits.

B 3.4.3'

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l B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 RCS Pressure and Temperature (P/T) Limits i

i l

BASES i

BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.

These loads are introduced by startup (heatup) and shutdown (cooldown) oxrations, power transients. and reactor trips. This

.C0 limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

The PTLR contains P/T limit curves for heatup, cooldown.

Inservice Leak and Hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature (Ref. 1).

Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LC0 establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB).

The vessel is the component most subject to brittle failure, and the LCO limits apply to the entire RCS (except the pressurizer).

The limits do not apply to the pressurizer, which has different design characteristics and operating functions.

10 CFR 50. Appendix G (Ref. 2), requires the establishment of P/T limits for specific material fracture toughness requirements of the RCPB materials.

Reference 2 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.

It mandates the use of the American Society of Mechanical Engineers (ASME) Code. Section

. Appendix G (Ref. 3).

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(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.4-10 Revision A

RCS P/T Limits B 3.4.3' BASES ACTIONS A.1 and A.2 (continued)

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Condition A is modified by a Note requiring Required Action A.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.

Restoration alone per Recuired Action A.1

.is insufficient because higher than analyzec stresses may have occurred and may have affected the RCPB integrity.

B.1 and B.2 If a Required Action and associated Comaletion Time of Condition A are not met, the unit must Je placed in a lower MODE because either the RCS remained in ari unacceptable P/T l

region for an extended period of increased stress or a sufficiently severe event caused entry into an unacceptable region.

Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure. nd t~;;r:ture.

In reduced l

pressure end teier:ture conditions, the possibility of propagation with undetected flaws is decreased.

If the required restoration activity of Required Action A.1 cannot be accom311shed within 30 minutes Required Action B.1 and Required Action B.2 must be implemented to reduce pressure and temperature.

If the required evaluation for continued operation cannot be o

accomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in Required Action B.1' and Required Action B.2.

A favorable engineering evaluation must be completed and documented before returning to j

operating pressure and temperature conditions.

Pressure and temperature are reduced by bringing the unit to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.4-15 Revision A

RCS P/T Limits B 3.4.3' BASES (continued) i SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is required every 30 minutes when RCS aressure and temperature conditions are undergoing planned c1anges. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments. 30 minutes permits assessment and correction for minor deviations within a reasonable time.

This SR 1's modified by a Note that only requires this SR to be performed during system heatup. cooldown, and ISLH testing. This SR is not required during critical operations i

because the combination of LCO 3.4.2 establishing a lower bound and the Safety Limits establishing an upper bound will i

provide adequate controls to prevent a change in excess of 100*F prior to entry into the performance condition of heatup and cooldown operations.

T-A, Rev;. ;on 2) 4 i

REFERENCES 1.

WCAP-14040f 'Metriodology used to Develop Cold

}

Overpressure Mitigating System Set Heatup and Cooldown Limit Curves." points and IM J=c 100

%eg 2.

10 CFR 50. Appendix G.

3.

ASME Boiler and Pressure Vessel Code.Section X I, Appendix G.

j 4.

ASTM E 185-82. July 1982.

j 5.

10 CFR 50. Appendix H.

6.

Regulatory Guide 1.99. Revision 2. May 1988.

i 7.

ASME Boiler and Pressure Vessel Code.Section XI.

l Appendix E.

.:=..e i

BRAIDWOOD - UNITS 1 & 2 B 3.4-17 Revision A

LTOP System B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 Low Temperature Overpressure Protection (LTOP) System BASES BACKGROUND The LTOP System controls RCS pressure at low temperatures so the integrity of the Reactor Coolant Pressure Boundary (RCPB) is not com temperature (P/T) promised by violating the pressure.andlimits The reactor vessel is the limiting RCPB component for demonstrating such protection. The PTLR provides the maximum allowable actuation logic setpoints for the pressurizer Power Operated Relief Valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg.

temperature during cooldown. shutdown, and heatup to meet the Reference 1 requirements during the MODES in which LTOP is necessary.

The reactor vessel material is less ductile at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2).

RCS pressure. therefore, is maintained low at low temperatures and is increased only within the limits specified in the PTLR.

~

The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only while shutdown:

a pressure fluctuation can occur more quickly than an operator can react to relieve the condition.

Exceeding the Aockchle (Arc RCS P/T limits by a significant amount could cause trittle

-creding of the reactor vessel.

LCO 3.4.3 "RCS Pressure and Temperature-(P/T) Limits." requires administrative control of RCS pressure and tem)erature during heatup and cooldown to prevent exceeding tie PTLR limits.

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.4-67 Revision A

LTOP System B 3.4.12 i

BASES LCO The elements of the LCO that provide low temperature (continued) overpressure mitigation through pressure relief are:

a.

Two OPERABLE PORVs:

b.

Two OPERABLE RHR suction relief valves:

c.

One OPERABLE PORV and one OPERABLE RHR suction relief valve; or d.

A depressurized RCS and an OPERABLE RCS vent.

A PORV is OPERABLE for LTOP when its block valve is open.

its lift setpoint is set to the limit required by the PTLR and testing proves its ability to open at this setpgint and motive power is available to the Mvehel and theirits control circuit \\.

Pond An RHR suct' ion relief valve is OPERABLE for LTOP when its RHR suction isolation valves are open, its setpoint is s 450 psig, and testing has proven its ability to open at this setpoint.

An RCS vent is OPERABLE when open with an area of a 2.0 square inches.

Each of these methods of overpressure prevention is capable of mitigating the limiting LTOP transient.

(continued)

BRAIDWOOD - UNITS 1 & 2 B 3.4-75 Revision A

LTOP System B 3.4.12 BASES ACTIONS C.1 and D.1 (continued)

An unisolated accumulator requires isolation within I hour.

This is only required when the accumulator pressure is at or more than the maximum RCS pressure for the existino temperature allowed by the P/T limit curves g k p g.g If the Required Action and associated Com>1etion Time of Condition C are not met. Required Action 3.1 must be performed in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Depressurizing the i

accumulators below the LTOP. limit from the PTLR prevents an accumulator pressure from exceeding the LTOP limits if the accumulators are fully injected.

The Completion Times are based on operating experience..that these activities can be accomplished in these time periods i

and on engineering evaluations indicating that an event requiring LTOP is not likely in the allowed times.

s M

In MODE 4, with one required RCS relief valve inoperable, the RCS relief valve must be restored to OPERABLE status within a Completion Time of 7 days.

Two RCS relief valves in any combination of the PORVS and the RHR suction relief valves are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.

The Completion Time considers that only one of the RCS i

relief valves is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is /ery j

low.

I (continued)

BRAIDWOOD - UNITS 1 & 2 B 3.4-77 Revision A

9 i

i ATTACHMENT C EVALUATION OF SIGNII'ICANT HAZARDS CONSIDERATIONS FOR j

PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, i

OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 1

Comed has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

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

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

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

Comed prcroses to amend Appendix A, Technical Specifications, ofFacility Operating Licenses FPI' 37, NPF-66, NPF-72, and NPF-77 to allow licensee control of the Reactor Coolant Sutem (RCS) Pressure and Temperature (P/T) limits for heatup, cooldown, low temperatum overpressure protection (LTOP) system, and hydrostatic testing. This change is

^

consistent with the guidance provided in Nuclear Regulatory Commission (NRC) Generic Letter (GL) 96-03, " Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits"; NUREG-1431, " Standard Technical Specifications for Westinghouse Plants," Revision 1; and WCAP-14040-NP-A, " Pressure and 1

Temperature Limits Methodology for Heatup, Cooldown, and COMS Analysis."

Additionally, Comed proposes to relocate the reactor vessel material surveillance program capsule withdrawal schedules in accordance with GL 91-01, " Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications."

The determination that the criteria set forth in 10 CFR 50.92 are met for this amendment request is indicated below:

l. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed changes relocate RCS P/T limits, LTOP system setpoints, hydrostatic testing requirements, and the reactor vessel capsule withdrawal schedule, along with supporting information, from the Technical Specifications to a PTLR. Compliance with these limits will continue to be required by the Technical Specifications. However, the limits themselves will be maintained in a Licensee-controlled document. Changes to the 1

limits will be controlled by Section 6.9.1.11 of the Technical Specifications. Changes to the RCS P/T limits can only be made in accordance with the NRC-approved methodologies listed in the Technical Specifications. The limits and the Technical Specifications will continue to assure the function of the reactor vessel as a pressure boundary. Revision to the LTOP limits can only be made in accordance with the approved methodologies listed in the Technical Specifications, and any resulting setpoint changes are made through the provisions of 10 CFR 50.59. Changes to the specimen withdrawal requirements are governed by Appendix H to 10 CFR 50.

The proposed changes do not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. They do not involve the addition or removal of any equipment, or any design changes to the facility. Therefore this proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

1

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes do not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed changes will not impose any new or different requirements or introduce a new accident or malfunction mechanism. There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite, and there is no significant increase in individual or cumulative occupational radiation exposure. In addition, the Byron and Braidwood Technical Specifications will continue to require that the reactor is maintained within acceptable operational limits and ensure that the LTOP system meets operability requirements. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

\\The proposed changes do not result in any reduction in the margin of safety because they have no impact on safety analysis assumptions. The proposed changes have been shown to ensure that the Pff and LTOP limits in the PTLR continue to meet all necessary requirements for reactor vessel integrity. Any future changes to the RCS P/T, LTOP limits, or supponing information must be performed in accordance with NRC-approved methodologies. Technical Specifications continue to require compliance with the limits in the PTLR. Additionally, any revision to the LTOP limits which result in setpoint changes will be evaluated under the provisions of 10 CFR 50.59. The reactor vessel capsule withdrawal schedule will continue to meet the requirements of Appendix H to 10 CFR 50.

Therefore, these changes do not involve a significant reduction in the margin of safety.

2

Comed has concluded that the RCS P/T and LTOP limits are no longer required to be i

located in the Technical Specifications under 10 CFR 50.36 or Section 182a of the Atomic Energy Act, and are not required to obviate the possibility of an abnormal situation or l

event giving rise to an immediate threat to the public health and safety. Additionally, they l

do not fall within any of the four criteria set forth in 10 CFR 50.36 (c)(2)(ii) for defining l

Technical Specification Limiting Condition for Operations.

r Therefore, based upon the above evaluation, Commonwealth Edison has concluded that these changes involve no significant hazards consideration.

i

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l.

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l-t 3

ATTACIIMENT D ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 Commonwealth Edison Company (Comed) has evaluated this proposed license amendment request against the criteria for identification oflicensing and regulatory actions requiring environmental assessment in accordance with Section 5121 of Title 10 to the Code of Federal Regulations, Part 51, (10 CFR 51.21). Comed has determined that this proposed license amendment request meets the criteria for a categorical exclusion set forth in 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 that changes a requirement with respect to installation or use of a facility component located within'the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

(i) the amendment involves no significant hazards considerations As demortrated in Attachment.C, this proposed amendment does not involve any significan'..zards considerations.

(ii) there is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite As documented in Attachment C, there will be no change in the types or significant increase in the amounts of any effluents released offsite (iii) there is no significant increase in individual or cumulative occupational radiation exposure.

The proposed changes will not result in changes in the operation or configuration of the facility. There will be no change in the level of controls or methodology used for processing of t'dioactive efiluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant.

Therefore, there will be no increase in individual or cumulative o :cupational radiation exposure resulting from this change.

l m

li ATTACHMENT E EXCEPTIONS TO WCAP-14040-NP-A METHODOLOGY C_prn.jp_yter Analysis Program for LTCP Setpoints Comed proposes to utilize alternate NRC-approved computer analysis programs for preparation of analysis that supports low temperature overprotection protection (LTOP) setpoints. WCAP-14040-NP-A references the use of Westinghouse's LOFTRAN code for j

determination of LTOP setpoints. Comed requests permission to use the NRC-approved RELAP code for this application. The setpoints and curves presented in this submittal are based on LOFTRAN; however, RELAP was used for analysis of replacement steam generator (RSG) LTOP setpoints to verify that the RSGs are bounded by the original steam generator I

(OSG) analysis. The RELAP analysis determined that the PORV setpoints for the RSG are always above the OSG maximum PORV setpoint curve. Comed intends to use RELAP as the basis for future Byron /Braidwood LTOP setpomts.

The RSG setpoint calculation was based on RELAP5/ MOD 2-B&W, version 20.0 HP. This version of the code carries a full certification and is approved for both LOCA and Non-LOCA applications (BWNT Document BAW-10164P, "RELAP5/ MOD 2, an Advanced Computer Program for Light-Water Reactor LOCA and Non-LOCA Transient Analysis," J. A.

l Klingenfuss, Revision 3, December 11,1992). Additionally, this code has already been accepted by the NRC for this app;ication'for a different licensee (NRC Letter "R E Ginna -

Acceptance for Referencing of Pressure Temperature Limits Report, Revision 1, TAC No.

M94770)," from J. A. Mitchell to R.C. Mectedy, May 2,1996, Docket No. 50-244).

Neutron Transport Cross-Section Library and Dosimeter Reaction Cross-Sections WCAP-14040-NP-A references the use of the BUGLE-93 neutron transport cross-section library, which is based on ENDF/B-VI nuclear data, for neutron transport calculations and i

dosimetry evaluations.

WCAP-14824, " Byron Unit 1 !!catup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron and Braidwood," Revision 1, Appendix C I

provides an assessment of the ENDF/B-IV based neutron transport cross-section libraries j

utilized in WCAPs 13880,14064,14241, and 14228 for Byron Units 1 and 2 and Braidwood Units 1 and 2, respectively, which were performed prior to the release of the ENDF/B-VI l

based cross-section library. The net effect of cross-section upgrades combined with low leakage fuel management on projected vessel fluence is expected to be very small and may resuh in an overall reduction in vessel fluence relative to that reported in WCAPs 13880,

{

14064,14241, and 14228. It is planned that neutron fluence evaluations for Byron Units 1 and 2 and Braidwood Units 1 and 2 will be updated to incorporate the use of ENDF/B-VI cross-section libraries at the time of the next scheduled capsule withdrawal for each of the i

?

l I

1

1 units. Based on the evaluation of WCAP-14824, Appendix C, considering the minimal expected impact of updating the neutron fluence evaluations from those previously reported along with the impact oflow leakage fuel management and the low sensitivity to irradiation damage exhibited by the limiting materials of the Byron Units 1 and 2 and Braidwood Units 1 and 2 reactor vessels, the use of the previously documented fluence values is justified until the update to the ENDF/B-VI methodology is completed for each unit.

The dosimetry evaluations performed with the neutron transport calculations were based on the use of dosimeter reaction cross-sections taken from the ENDF/B-V database. Since the' dosimetry evaluations involve scaling the calculated results to the. measurements and since the change to ENDF/B-VI cross-sections has little effect on the measurements, it is expected that upgrading to the ENDF/B-VI would result in small changes for the respective vessels.

Comed requests permission to use an ENDF/B-IV based cross-section library and ENDF/B-V dosimeter reaction cross-sections, which were the most currently available at the time the calculations were performed, in place of ENDF/B-VI based data.

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Year WCAP-14040-NP-A methodology includes use of ASME Code,Section XI, Appendix G, Article G-2000 methodology applicable through the 1988 Addenda and editions through the 1989 Edition. Comed requested an exemption from 10 CFR 50.60 in a letter from J. Hosmer i

to the NRC dated April 3,1997. The exemption would allow Comed to use a later version of the ASME Code than currently specified in the CFR,i.e., the 1996 Addenda of ASME Section XI, Appendix G, Article G-2000, in the development offuture allowable P/T limits, rather than addenda through the 1988 Addenda and editions through the 1989 Edition. The l

curves provided in the attached Pressure Temperature Limit Repons are based on the earlier edition of the Code, however, future PTLR curves will be developed using the later edition of the Code.

1

)

2 a

ML20141K001

Text

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