Text

Amends 65,65,56 & 55 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively,changing TS to Reflect Reduced Thermal Flow to Compensate for Increased SG Tube Plugging Up to 15 Percent of Total Number of Tubes
	ML20078C223
Person / Time
Site:	Byron, Braidwood
Issue date:	10/21/1994
From:	Dick G; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20078C225	List: ML20078C223; ML20078C226;
References
	NPF-37-A-065, NPF-66-A-065, NPF-72-A-056, NPF-77-A-055 NUDOCS 9410310009
	Download: ML20078C223 (58)
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UNITED STATES

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D.C. 20655-0001 g; J/

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1 COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE i

Amendment No. 65 License No. NPF-37 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated March 23, 1994, as supplemented on July 26, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I, 1

i B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 1

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and l

paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:

9410310009 941021 PDR ADOCK 05000454 P

PDR j

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(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 65 and the Environmental Protection Plan contained in Appendix B, both of which arc attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and i

the Environmental Protection Plan.

e 3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jit)

.l Georg F. Dick, Jr., Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV l

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 21, 1994 i

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y-pk UNITED STATES j.kil

! j NUCLEAR REGULATORY COMMISSION

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WASHINGTON. D.C. 30665-0001 g,,, v y COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-455 BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 65 License No. NPF-66 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated March 23, 1994,- as supplemented on July 26, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1-t B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized

?

by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and j

paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby l

amerded to read as follows:

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. (2)

Technical Soecifications The Technical Specifications contained in Appendix A (NUREG-Ill3),

as revised through Amendment No. 65 and revised by Attachment 2 to NPF-66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, dated February 14, 1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which-is hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Geo F. Dick, Jr., Project Manager Project Directorate III-2 Division ci Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 21, 1994 L

7---

ATTACHMENT TO LICENSE AMENDMENT NOS. 65 AND 65 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Pages indicated by an asterisk are provided for convenience only.

Remove Paoes Insert Paaes III III 2-1 2-1 2-2 2-2 2-2a 2-5 2-5 2-7 2-7 2-8 2-8 2-10 2-10 B 2-1 B 2-1 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12 3/4 2-8 3/4 2-8 3/4 5-1 3/4 5-1 3/4 5-11 3/4 5-11 3/4 9-1 3/4 9-1 3/4 9-2*

3/4 9-2*

5 3/4 1-2 B 3/4 1-2 8 3/4 1-3 B 3/4 1-3 B 3/4 2-3*

B 3/4 2-3*

B 3/4 2-4 B 3/4 2-4 8 3/4 5-4 8 3/4 5-4 8 3/4 6-3 8 3/4 6-3 8 3/4 6-4*

B 3/4 6-4*

B 3/4 9-1 B 3/4 9-1 l

l

SAFETY llMITS AND LIMITING SRFETY SYSTEM SETTINGS SECTION EAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................

2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

2-1 l

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION..

2-2 l

FIGURE 2.1-la REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION..

2-2a 2.2 llMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATI0lt SETP0lNTS...............

2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS....

2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS...............

B 2-3

{

BYRON - UNITS 1 & 2 III AMENDMENT NO. 65

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temper,ature (T,y) loop operation.

l shall not exceed the limits shown in Figure 2.1-1 (Figure 2.1-la) for four APPLICABILITY: MODES I and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within I hour, and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES I and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within this limit within I hour, and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within this limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

- Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 2-1 AMENDMENT NO. 65

680 670 3471. pela.'

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580 O

O.2 0.4 0.6 0.8 1

1.2 Power (Fraction of Nominal)

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION Applicable to Unit 1.

Not applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 2-2 AMENDMENT N0. 65

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.__..:.=.=..l_..._=_,.=.=._..-.,...._._-j_....a_..l._..._.,__._,=._;=_i=_..:.=.J 20 40 60 80 100 120 POWER (PERCENT)

FIGURE 2.1-la REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 2-2a AMENDMENT N0. 65

TABLE 2.2-1 (Continued 1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0fNTS FUNCTIONAL UNIT TRIP SETPOINTS All0WABLE VALUE

12. Reactor Coolant Flow-Low 290% of loop mini-289.3% of loop mini-mum measured flow mum measured flow

13. Steam Generator Water level Low-Low a.

Unit 1 233.0% of narrow 231.0% of narrow L

range instrument range instrument span span b.

Unit 2 236.3% of narrow 234.8% of narrow range instrument range instrument span span

14. Undervoltage - Reactor 25268 volts -

14920 volts -

Coolant Pumps each bus each bus

15. Underfrequency - Reactor 257.0 Hz 256.08 Hz Coolant Pumps

16. Turbine Trip a.

Emergency Trip Header 21000 psig 2815 psig-Pressure b.

Turbine Throttle Valve 21% open 21% open Closure

17. Safety injection Input N.A.

N.A.

from ESF

18. Reactor Coolant Pump N.A.

N.A.

Breaker Position Trip

Minimum measured flow = 92,850 gpm" (97,600 gpm)**

- Not Applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

BYRON - UNITS 1 & 2 2-5 AMENDMENT NO. 65 n-r-u

e wy----

w-

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w

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- =

3----,

TABLE 2.2-1 (Continued) l TABLE NOTATIONS i

NOTE 1: OVERTEMPERATURE AT A T ((1+t S) ( 1+T S) s A T [K -K (1+t S) [T( 1+t S) -T'] +K (P-P') -f ( AI)}

1 1

(1+t S) 1 4

o 3

2 3

1 1+t S) 2 3

3 s

Where:

AT Measured AT by RTD Manifold Instrumentation, 1+T S t

L ad-lag compensator on measured AT, 1+t S 2

Time constants utilized in lead-lag compensator for AT, r, - 8 s, 7,

r 3

2 72-3s, 1

Lag compensator on measured AT, 1+T s 3

3 Time constants utilized in the lag compensator for AT, r3 - O s, r

4 AT, Indicated AT at RATED THERMAL POWER,

=

1.325*, (1.164)**

K, 0.0297/*F*, (0.0265/*F)**

l K

2 1 + T'S The function generated by the lead-lag compensator for T, pys5 dynamic compensation, Time constants utilized in the lead-lag compensator for T,, r4 - 33 s, 7,

r 4

3 r3-4s, T

Average temperature,

F,

,,' Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 7-7 AMENOMENT NO. 65

~

TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

NOTE 1:

(Continued) i 1

i Lag compensator on measured T,,,

1+t S

=

J Time constant utilized in the measured T,,' lag compensator, T6 - O s, 7

6 T'

5; 588.4*F (Nominal T,,at RATED THERMAL POWER),

0.00181*, (0.00134)"

K 3

Pressurizer pressure, psig, P

=

2235 psig (Nominal RCS operating pressure),

P' Laplace transform operator, s,

S and f,(AI) is a function of the indicated difference between top and bottom detectors.of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:

4 for q - q, between -24%* (-32%)" and +10%* (+13%)". f AI 0 where q and q, are percent RATED, THERMAL POWER in the top and bottom halves of th(e c)or=e r,espective*1y, and q, +

(i) total THERMAL POWER in percent of RATED THERMAL POWER; (1.74%)3, f its value at RATED THERMAL POWER.q, exce (ii) for~eachpercentthatthemagnjtudeof automatically reduced by 4.11%

o (1.67%)3, f its value at RATED THERMAL POWER.q, exce (iii) for each percent that the magnjtude of l

automatically reduced by 3.35%

o L

NOTE 2:

The channel's gaximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.16% (3.71%)

of AT span.

j

,,'Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

RYRON - IINTTS 1 A 7 2-8 AMEN 0 MENT

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1 TABLE 2.2-1 (Continued)

't TABLE NOTATIONS (Continued)

NOTE 3:

(Continued) 0.00245/*F* (0.00170/*F) *, for T > T" and K, = 0 for T s T",

K 6

T As defined in Note 1,

=

T" Indicated T at RATED THERMAL POWER (Calibration temperature for 61

=

instrumental, ion,s588.4*F),

S As defined in Note 1, and

=

f (AI) 0 for all AI.

=

2 NOTE 4:

The channel's p,aximum Trip Setpoint shall not exceed its computed Trip-Setpoint by more than 3.08% (2.31%)

of oT span.

4 l

t 1

l l

,, Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

i Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 P-10 AMFNOMFNT No. 65 j

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB.

This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The DNBR thermal design criterion is that the probability that DNB will not occur on the most limiting rod is at least 95% (at a 95% confidence level) for any Condition I or 11 event.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered. As described in the UFSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty.

Design limit DNBR values have been determined that satisfy the DNB design criterion.

The design DNBR values are 1.25 for the typical and thimble cells (1.34 and 1.32 for a typical cell and a thimble cell, respectively for 0FA** fuel, and 1.33 for a typical cell and 1.32 for a thimble cell for the VANTAGE 5 fuel *).

In addition, margin has been maintained in both designs by meeting safety analysis DNBR limits of 1.50 for the typical and thimble cells (1.49 for a typical cell and 1.47 for a thimble cell for 0FA fuel, and 1.67 and 1.65 for a typical cell and a thimble cell, respectively for the VANTAGE 5 fuel *)

in performing safety analyses.

j The curves of Figure 2.1-1 (Figure 2.1-la*) show the loci of points of l

THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum design DNBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

l

Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

- 0ptimized fuel assembly.

BYRON - UNITS 1 & 2 B 2-1 Amendment No. 65

t REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

T a.

A Boric Acid Storage System with:

1)

A minimum contained borated water level of 7.0%,

L 2)

A minimum boron concentration of 7000 ppm, and i

3)

A minimum solution temperature of 65*F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water level of 9.0%,

2) a)* A boron concentration between 2300 and 2500 ppm, b)** A minimum boron concentration of 2000 ppm, and 3)

A minimum solution temperature of 35*F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involvi' g CORE n

ALTERATIONS or positive reactivity changes.

SURVEILLANCE RE0VIREMENTS f

4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

j 1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water level, and r

3)

Verifying the boric acid storage tank solution temperature when it is the source of borated water.

I b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 35*F.

I i

i

Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

,,Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 3/4 1-11 AMENDMENT NO. 65 4

~

REACTIVITY CONTROL SYSTEMS B0 RATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by. Specification 3.1.2.2 for MODES 1, 2 and 3 and one of the following borated water sources shall be OPERABLE as required by Specification 3.1.2.1 for MODE 4:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water level of 40%,

2)

A minimum boron concentration of 7000 ppm, and 3)

A minimum solution temperature of 65*F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water level of 89%,

2) a)* A boron concentration between 2300 and 2500 ppm, b)** A minimum boron concentration of 2000 ppm, 3)

A minimum solution temperature of 35*F, and 4)

A maximum solution temperature of 100*F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With the Boric Acid Storage System inoperable and being used as one a.

of the above required borated water sources in MODE 1, 2, or 3, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the RWST inoperable in MODE 1, 2, or 3, restore the tank to-OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at-least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With no borated water source OPERABLE in MODE 4, restore one borated water source to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

,Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 3/4 1-12 AMENDMENT NO. 65 l

POWER DISTRIBUTION LIMITS 3 /4. 2. 3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 Indicated Reactor Coolant System (RCS) total flow rate and Fl, shall be maintained as follows for four loop operation.

1)* RCS Total Flowrate 2371,400 gpm, a.

2)** RCS Total Flowrate 2390,400 gpm, and b.

Fl, 51.55 [1.0 + 0.3 (1.0-P)] for 0FA fuel Fl, s1.65 [1.0 + 0.3 (1.0-P)] for VANTAGE 5 fuel where:

Measured values of Fl, are obtained by using the movable incore detectors. An appropriate uncertainty of 4% (nominal) or greater shall then be applied to the measured value of Fl, before it is compared to the requirements, and P=

THERMAL POWER RATED THERMAL POWER APPLICABILITY:

MODE 1.

AC'10N:

With RCS total flow rate or Fl, outside the region of acceptable operation:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1.

Restore RCS total flow rate and Fl, to within the above limits, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l

Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

,,Not applicable to Unit 1.

Applicable to Unit 2 until the completion of cycle 5.

BYRON - UNITS 1 & 2 3/4 2-8 AMENDMENT NO. 65

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:

a.

The isolation valve open and power removed, b.

A contained borated water level of between 31% and 63%,

c.

1)" A boron concentration between 2200 and 2400 ppm, 2)** A boron concentration between 1900 and 2100 ppm, and d.

A nitrogen cover-pressure of between 602 and 647 psig.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE0VIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1)

Verifying the contained borated water level and nitrogen cover-pressure in the tanks, and 2)

Verifying that each accumulator isolation valve is open.

Pressurizer pressure above 1000 psig.

Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

- Not applicable to Unit 1.

Applicable to Unit 2 until comp'etion of cycle 5.

BYRON - UNITS 1 & 2 3/4 5-1 AMENDMENT NO. 65

EMERGENCY CORE Ck)0 LING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK

~

LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) and the heat traced portion of the RWST vent path shall be OPERABLE with:

a.

A minimum contained borated water level of 89%,

b.

1)* A boron concentration between 2300 and 2500 ppm, 2)** A minimum boron concentration 'of 2000 ppm, c.

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

A maximum water temperature of 100*F.

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

ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within I hour or be in at least H0T STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

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

-Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when j

the outside air temperature is either less than 35*F or greater l

than 100*F, and j

c.

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

Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

,,Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 3/4 5-11 AMENDMENT NO. 65

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:

a.

A K,,, of 0.95 or less, or b.

1)" A boron concentration of greater than or equal to 2000 ppm.

2)" A boron concentration of greater than or equal to 2300 ppm.

APPLICABILITY: MODE 6*.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until K is reduced to less than or equal to 0.95 or the boro concentration is r,e,s,tored to greater than or equal to 2000 ppm" (2300 ppm,n),

whichever is the more restrictive.

SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a.

Removing or unbolting the reactor vessel head, and b.

Withdrawal of any full-length control rod in excess of 57 steps (approximately 3 feet) from its fully inserted position within the 3

reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.9.1.3 Valves CVlllB, CV8428, CV8441, CV8435, and CV8439 shall be verified i

closed and secured in position by mechanical stops or by removal of air or electrical power at least once per 31 days.

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

Not applicable to Unit 1.

Applicable to Unit 2 until the completion of cycle 5.

- Applicable to Unit 1.

Not applicable to Unit 2 until after cycle 5.

BYRON - UNITS 1 & 2 3/4 9-1 AMENDMENT NO. 65 y

I

I REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two Source Range Neutron Flux Monitors shall be OPERABLE each with continuous visual indication in the control room and one with audible I

indication in the containment and control room.

APPLICABILITY:

MODE 6.

ACTION:

l l

With one of the above required monitors inoperable or not operating, a.

immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

b.

With both of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:

a.

A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , i

b.

An ANALOG CHANNEL OPERATIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS, and c.

An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.

i l

1 l

l

\\

l BYRON - UNITS 1 & 2 3/4 9-2 l

l l

REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions.

These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.

Th transformed into th,e limiting MTC value -4.1 x 10',is value of the MDC was then for burnup and soluble /k/*F represents a conservative /k/*F.

M The MTC value of -3.2 x 10' M value (with corrections boron at a core condition of 300 ppm equilibrium boron value of -4.1 x 10'f obtained) by making these corrections to the lim concentration and i M/k/*F.

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC can be maintained within its limits.

The BOL MTC n asurement, combined with the predicted MTC throughout core life, will be used to impose administrative limits on rod withdrawal, as required during core life to ensure that MTC will always be less positive than 0 M /k/*F.

This coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 550*F.

This limitation is required to ensure: (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the OPERABLE status with a steam bubble, pressurizer is capable of being in an temperature, and (5) the(4) the reactor vessel is above its minimum,RT,P-12, plant is above the cooldown steam dump permissive, 3/4.1.2 BORAT10N SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each MODE of facility operation.

The components required to perform this function include:

(3) separate flow paths, (4) bor(1) borated water sources, (2) charging pumps, ic acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 350*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUT 00WN MARGIN from expected operating conditions of 1.3% M/k after xenon decay and cooldown to 20p)*F.The maximum expected boration capability requirement is 13,487 tanks or(15,780 gallong of 7000-ppm borated water from the boric acid storage 54,014 (70,450 ) gallons of 2300-ppm (2000-ppm,) borated water from the refueling water storage tank. A Boric Acid Storage System level of 49%

ensures that there is a volume of greater than or equal to 13,487 (15,780 )

l gallons available.

A RWST level of 89% ensures that there is a vo' ume of greater than or equal to 395,000 gallons available.

Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 B 3/4 1-2 AMENDMENT NO. 65

REACTIVITY CONTROL SYSTEMS BASES B0 RATION SYSTEMS (Continued)

With the RCS temperature below 350*F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 330*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or an RHR Suction relief valve.

The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% Ak/k after xenon decay and cooldown from 200*F to 140*F.

This condition requires either 740 (2,652*) gallons of 7000-ppm borated water from the boric acid storage tanks or 2264 (11,840 ) gallons of 2300-ppm (2000-ppm ) borated water from the refueling water storage tank (RWST).

A Boric Acid Storage System level of 7% ensures there is a volume of greater than or equal to 740 (2652*) gallons available. An RWST level of 9% ensures there is a volume of greater than or equal to 38,740 gallons available.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.0 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

The OPERABILITY of the automatic Boron Dilution Protection System ensures adequate capability for negative reactivity insertion to prevent a transient caused by the uncontrolled dilution of the RCS in MODES 3,4, and 5.

The functioning of the system precludes the necessity of operator action to prevent further dilution by terminating flow to the charging pump (s) from possible unborated water sources and initiating flow from the RWST. The most restrictive condition occurs shortly after beginning of life when the critical boron concentration is highest, and a 205 gpm dilution flowrate provides the maximum positive reactivity addition rate. One reactor coolar t pump in operation with all reactor coolant loop stop isolation valves open reduces the reactivity addition rate by mixing the dilution through all four reactor coolant loops. A minimum count rate of ten counts per second minimizes the impact of the uncertainties associated with the source range nuclear instrumentation.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3 Ak/k is required to control the reactivity transient. Actions taken by

Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 B 3/4 1-3 AMENDMENT NO. 65

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INDICATED AXIAL FLUX DIFFERENCE FIGURE B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER BYRON - UNITS 1 & 2 B 3/4 2-3 1

POWER DISTRIBUTION LIMITS BASES HEAT F.UX HOT CHANNEL FACTOR. and RCS FLOW RATE AMD NUCLEAR ENTHALPY RISE HOT CHANNE. FACTOR (Continued) c.

The control rod insertion limits of Specification 3.1.3.6 are maintained, and d.

The axial ower distribution, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the limits.

F"E d. above are main ained.will be maintained within its limits provided the Conditions a.

throuk00gpm(390,400 gpm})] and the requirement on F"the RCS flow requirement l

The combination of

1371, guarantee that the DNBR used in the safety analysis will be met.

Margin between the safety analysis limit DNBRs [i and thimble cells 1.50 for the typical and' thimble cells (d 1.67 and 1.65 for the VANTAGE 5 typical and thimble cells,)]

1.49 and 1.47 for the OFA fuel typica l

respectively an and the design limit DNBRs [l and thimble cells, and 1,)33 and 1.3 1.25 for the typical and thimble cells (1.34 and 1.32 for the 0FA fuel typica VANTAGE 5 fuel typical and thimble cells, respectively ] is maintained.

A fraction of this margin is utilized to accommodate the transition core DNBR penalty (maximum of 12.5%) Revision 1).and the appropriate fuel rod bow DNBR penalty (less than 1.5% per WCAP-8691, The rest of the margin between design and safety analysis DNBR limits can be used for plant design flexibility.

1 The RCS flow requireglent is based on the loop minimum measured flow rate of 92,850 gpm (97,600 gpm ) which is used in the Revised Thermal Design Procedpre (Improved Thermal Design Procedure described in UFSAR 4.4.1 and 15.0.3 to cali)brate the RCS flow rate indicators.A precision heat balance is performed once each cycle and is used Potential fouling of the feedwater venturi, which might not be detected, could bias the results from the precision heat balance in a non-conservative manner.

Therefore a penalty of 0.1% is assessed for potential feepwater venturi fouling.

Amaximum measured flow rate to account for(2.2% ) has been included in the loop minimum l

measurement uncertainty of 3.5%.

potential undetected feedwater venturi l

fouling.and the use of the RCS flow indicators for flow rate verification.

Any fouling which might bias the RCS flow rate measurement greater than 0.1%

l can be detected by monitoring and trending various plant performance parameters.

If detected, action shall be taken, before performing subsequent precision heat balance measurements, i.e., either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate the fouling.

l Surveillance Requirement 4.2.3.4 provides adequate monitoring to detect possible flow reductions due to any rapid core crud buildup.

Surveillance Requirement 4.2.3.5 specifies that the measurement instrumentation shall be calibrated within seven days prior to the performance of the calorimetric flow measurement. This requirement is due to the fact that the drift effects of this instrumentation are not included in the flow measurement uncertainty analysis. This requirement does not apply for the instrumentation whose drift effects have been included in the uncertainty analysis.

I

Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

j i

BYRON - UNITS 1 & 2 B 3/4 2-4 Amendment No. 65 j

EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.5 REFVELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

The contained water volume limit includes an allowance for water not usable because 'of tank discharge line location or other physical characteristics. A minimum contained borated water level of 89% ensures a volume of greater than or equal to 395,000 <;allons.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.0 and 11.0 for the solution recirculated within containment after a LOCA.

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

BYRON - UNITS 1 & 2 B 3/4 5-4 AMENDMENT N0. 65

CONTAINMENT SYSTEMS BASES CONTAINMENT PURGE VENTILATION SYSTEM (Continued) be exceeded in the event of an accident during containment purging operation.

Operation with one line open will be limited to 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months during a calendar year.

Leakage integrity tests with a maximum allowable leakage rate for containment purge supply and exhaust supply valves will provide early indica-tion of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop.

The 0.60 L leakage limit of Specification 3.6.1.2.b. shall not be exceeded when the leak, age rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment depressurization and cooling capability will be available in the event of a LOCA or steam line break.

The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses.

The Containment Spray System and the Containment Cooling System are redundant to each o!'.er in providing post-accident cooling of the containment atmosphere.

However, the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient Na0H is added to the containment spray in the event of a LOCA.

The limits on Na0H volume and concentration ensure a pH value of between 8.0 and 11.0 for the l

solution recirculated within containment after a LOCA.

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

The contair,ed solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics.

These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses.

A spray additive tank level of between 78.6% and 90.3% onsures a volume of greater than or equal to 4000 gallons but less than or equal to 4540 gallons.

BYRON - UNITS I & 2 B 3/4 6-3 Amendment No. 65

CONTAINMENT SYSTEMS BASES 3/4.6.2.3 CONTAINMENT COOLING SYSTEM The OPERABILITY of the Containment Cooling System ensures that: (1) the containment air temperature will be maintained within limits during normal operation, and (2) adequate heat removal capacity is available when operated in conjunction with the Containment Spray Systems during post-LOCA conditions.

l The Containment Cooling System and the Containment Spray System are l

redundant to each other in providing post accident cooling of the containment atmosphere.

As a result of this redundancy in cooling capability, the allowable out-of-service time requirements for the Containment Cooling System have been appropriately adjusted.

However, the allowable out-of-service time require-ments for the Containment Spray System have been maintained consistent with that assigned other inoperable ESF equipment since the Containment Spray i

System also provides a mechanism for removing iodine from the containment atmosphere.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of i

GDC 54 thru 57 of Appendix A to 10 CFR Part 50.

Containment isolation within I

the time limits specified for those isolation valves designed to close auto-matically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

3.'4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to raintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions.

Either recombiner unit (or the Purge System) is capable of controlling the expected hydrogen generation associated with: (1) 2irconium water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment.

These Hydrogen Control Systems are consistent with t.ie recommendations of Requiatory Guide 1.7, " Control of Comoustible Gas Concentrations in Containment Following a LOCA " March 1971.

The Hydrogen Mixing Systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA.

This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

BYRON - UNITS 1 & 2 B 3/4 6-4

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain subtritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on Keff of no greater than 0.95 is sufficient to prevent reactor criticality during refueling operations and includes a 1% ok/k conservative allowance for uncertainties.

Similarly, the boron concentration value of 2300 ppm (2000 ppm ) or greater includes a conservative uncertainty allowance of 50 ppm. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The locking closed of the required valves during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portions of the RCS.

This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subtriticality prior to movement of irradiated fuel assenblies in the reactor vessel ensures that sufficient time has elapsed to allot the radioactive decay of the short-lived fission products.

l This decay time is :onsistent with the assumptions used in the safety analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS I

The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment.

The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

The Byron Station is designed such that the containment opens into the t

fuel building through the personnel hatch or equipment hatch.

In the event of i

a fuel drop accident in the containment, any gaseous radioactivity escaping from the containment building will be filtered through the Fuel Handling Building Exhaust Ventilation System.

+

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

Not applicable to Unit 1.

Applicable to Unit 2 until completion of cycle 5.

BYRON - UNITS 1 & 2 B 3/4 9-1 Amendment No. 65

f>oa atog%

4.-

?

UNITED STATES

{(C [ g IhW S

NUCLEAR REGULATORY COMMISSION i

WASHINGTON, D.C. 30666 4001

,,e ss, COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO. 1 AMENDMENT TO FACIllTY OPERATING LICENSE Amendment No. 56 License No. NPF-72 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated March 23, 1994, as supplemented on July 26, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendmant is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-l cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:

?

i i

~

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 56 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license.

The licensee shall operate-the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Ramin R. Assa, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation t

Attachment:

Changes to the Technical Specifications Date of Issuance: October 21, 1994 i

a

)

QC8cg

?g UNITED STATES y

2 S

NUCLEAR REGULATORY COMMISSION

'f WASHINGTON, o.C. 20555-0001

%...../

COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-457 E_RAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 55 License No. NPF-77 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated March 23, 1994, as supplemented on July 26, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, the-provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:

i

. 1

.(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 55 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date if its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/

Ramin R. Assa, Project Manager i

Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation i

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 21, 1994 1

1

ATTACHMENT T0 LICENSE AMENDMENT NOS. 56 AND 55 FACILITY OPERATING LICENSE N05. NPF-72 AND NPF-77 j

DOCKET NOS. STN 50-456 AND STN 50-457-Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Pages indicated by an asterisk are provided for convenience only.

Remove Paoes Insert Paaes III III 2-1 2-1 2-2 2-2 2-2a 2-5 2-5~

2-7 2-7 2-8 2-8 2-10 2-10 B 2-1 B 2-1 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12 3/4 2-8 3/4 2-8 3/4 5-1 3/4 5-1 i

3/4 5-11 3/4 5-11 3/4 9-1 3/4 9-1 3/4 9-2*

3/4 9-2*

B 3/4 1-2 B 3/4 1-2 B 3/4 1-3 B 3/4 1-3 B 3/4 2-3*

B 3/4 2-3*

8 3/4 2-4 B 3/4 2-4 8 3/4 5-4 B 3/4 5-4 8 3/4 6-3 8 3/4 6-3 B 3/4 6-4*

B 3/4 6-4*

B 3/4 9-1 B 3/4 9-1 l

i a

b f

4

SAFETY LIMfTS AND LIMfTING SAFETY SYSTEM SETTINGS SECTION E6SE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................

2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION..

2-2 l

FIGURE 2.1-la REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.

2-2a l

2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS...............

2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0lNTS...

2-4 I

t BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE.................................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS...............

B 2-3 l

UNIT 1 -

AMENDMENT NO. 56 BRAIDWOOD - UNITS 1 & 2 III UNIT 2 -

AMENDMENT N0. 55

2.0 SAFETY LIMITS AND LfMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORf 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temper,,ature (T,y) loop operation.

shall not exceed the limits shown in Figure 2.1-1 (Figure 2.1-la)

for four APPLICABILITY: MODES I and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within I hour, and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within this limit within I hour, and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within this limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

,,' Applicable to Unit I and Unit 2 starting with cycle 6.

Applicable to Unit I and Unit 2 until completion of cycle 5.

Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 2-1 Unit 2 - Amendment No. 55

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FIGURE 2.1-1 REACTOR CORE SAFETY LlHIT - FOUR LOOPS IN OPERATION Applicable to Unit I and Unit 2 until completion of cycle 5 Unit 1 - Amendment No. 56 be

"'9 - UNI *lS 1 & 2 2-2 Unit 2 - Amendment No. 55

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Figure 2.1-la Reactor Core Safety Limit - Four Loops in Operation Applicable to Unit I and Unit 2 starting with cycle 6 Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 2-2a Unit 2 -

Amendment No. 55

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

12. Reactor Coolant Flow-low 290% of loop mini-289.3% of loop mini-mum measured flow mum measured flow

13. Steam Generator Water Level Low-Low a.

Unit 1 233.0% of narrow 231.0% of narrow range instrument range instrument span span b.

Unit 2 217% (Cycle 3);

216.3% (Cycle 3);

236.3% (Cycle 4 234.8% (Cycle 4 and and after) of after) of narrow narrow range range instrument instrument span span

14. Undervoltage - Reactor 25268 volts -

24920 volts -

Coolant Pumps each bus each bus

15. Underfrequency - Reactor 257.0 Hz 256.08 Hz Coolant Pumps

16. Turbine Trip a.

Emergency Trip Header 21000 psig 1815 psig Pressure b.

Turbine Throttle Valve 21% open 21% open Closure

17. Safety Injection Input N.A.

N.A.

from ESF

18. Reactor Coolant Pdmp N.A.

N.A.

Breaker Position Trip

Minimum measured flow - 97,600 gpm" (92,850 gpm)'

- Applicable to Unit 1 and Unit 2 until completion of cycle 5.

Applicable to Unit I and Unit 2 starting with cycle 6.

UNIT 1 -

AMENDMENT NO. 56 BRAIDWOOD - UNITS 1 & 2 2-5 UNIT 2 -

AMENDMENT NO. 55

TABLE 2.2-1 (Continued)

~

TABLE NOTATIONS NOTE 1: OVERTEMPERATURE AT A T ((1 +t S) ( 1+T S ) s A TfK -K (1 +T S) [ T( 1 +T,S ) -T'] +K (P-P') -f ( AI) 1 1

4 1

1 2

1+T S)

(1 +T S) 3 1

2 3

s r

Where:

'AT Measured AT by RTD Manifold Instrumentation, 1+t s i

Lead-lag c mpensator on measured AT, 1+t S 2

Time constants utilized in lead-lag compensator for AT, r, - 8 s, r,r i

2 r2 = 3 s.

1 1+t s Lag compensator on measured AT, 3

Time constants utilized in the lag compensator for AT, r3 - O s, r

3 AT, Indicated AT at RATED THERMAL POWER, 1.164,* 1.325**

K, 0.0265/*F,* 0.0297/*F**

K 2

1 + T*S The function generated by the lead-lag compensator for T, g

dynamic compensation, s

Time constants utilized in the lead-lag compensator for T

, 74 - 33 s, 7,

7 4

3 r3-4s, 1

T Average temperature, "F,

~ Applicable to Unit I and Unit 2 until completion of cycle 5.

,, Applicable to Unit I and Unit 2 starting with cycle 6.

UNIT 1 -

AMENDMENT NO. 56 BRAIDWOOD - UNITS 1 & 2 2-7 UNIT 2 -

AMEN 0 MENT NO. 55

TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

NOTE 1:

(Continued) 1 Lag compensator on measured T,y,,

1+T s Time constant utilized in the measured T,y lag compensator, 76-O s, 7

=

6 T'

s 588.4*F (Nominal T,,, at RATED THERMAL POWER),

0.00134*, 0.00181**

K 3

P Pressurizer pressure, psig, P'

2235 psig (Nominal RCS operating pressure),

Laplace transform operator, s",

S and f,(AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:

(i) for q - a between -32%, -24%" and +13%', +10%** f (AI) - 0. where q, and q are percent RATED *THERNAL POWER in the top and bottom halves of the core respectively, a,nd q, + q3 3

is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnjtude of,,q, f its value at RATED THERMAL POWER-q, exceed automatically reduced by 1.74%, 4.11%

o (iii) for each percent that the magnjtude of,,q, f its value at RATED THERMAL POWER

- q, exceeds -32%', -24%

the AT trip setpoint shall be automatically reduced by 1.67%, 3.35%

o NOTE 2:

Thechannel's,,maximumTripSetpointshallnotexceeditscomputedTripSetpointbymorethan 3.71%, 1.16%

of AT span.

l

,,' Applicable to Unit I and Unit 2 until completion of cycle 5.

Applicable to Unit I and Unit 2 starting with cycle 6.

UNIT 1 - AMENDMENT NO. 56 BRAIDWOOD - UNITS 1 & 2 2-8 UNIT 2 - AMENDMENT NO. 55

TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

NOTE 3:

(Continued) 0.00170/*F*, 0.00245/*F** for T > T" and K6 - O for T s T",

K 6

T As defined in Note 1,

=

T" Indicated T a

instrumentaI. Ton,t RATED THERMAL POWER (Calibration temperature for AT s 588.4*F),

S As defined in Note 1, and f (AI) 0 for all AI.

=

2 NOTE 4:

The channel's,, maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.31%, 3.08%

of AT span.

,,' Applicable to Unit I and Unit 2 until completion of cycle 5.

Applicable to Unit I and Unit 2 starting with cycle 6.

UNIT 1 - AMENDMENT NO. 56 BRAIDWOOD - UNITS 1 & 2 2-10 tlNIT ?. - AMFNDMFNT NO. 55

2.1 SAFETY LIMITS BASES 2,1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB.

This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distri-butions.

The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The DNBR thermal design criterion is that the probability that DNB will not occur on the most limiting rod is at least 95% (at a 95% confidence level) for any Condition I or II event.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered.

As described in the UFSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty. Design limit DNBR values have been determined that satisfy the DNB design criterion.

The design DNBR values are 1.34 and 1.32 for a typical cell and a thimble cell, respectively for 0FA* fuel, and 1.33 for a typical cell and 1.32 for a,,

thimble cell for the VANTAGE 5 fuel (1.25 for the typical and thimble cells).

I In addition, margin has been maintained in both designs by meeting safety analysis DNBR limits of 1.49 for a typical cell and 1.47 for a thimble cell for 0FA fuel, and 1.67 and 1.65 for a typical cell and a thimble cell, respectively for the VANTAGE 5 fuel (1.50 for the typical and thimble cells)" in performing l

safety analyses.

The curves of Figure 2.1-1 (Figure 2.1-la)" show the loci of points of l

THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum design DNBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

' Optimized Fuel Assemblies

" Applicable to Unit I and Unit 2 starting with cycle 6.

Unit 1 -

Amendment No.

56 BRA]DWOOD - UNITS 1 & 2 B 2-1 Unit 2 -

Amendment No. 55

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION I

3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water level of 7.0%,

2)

A minimum boron concentration of 7000 ppm, and 3)

A minimum solution temperature of 65*F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water level of 9.0%,

2) a)

A minimum boron concentration of 2000 ppm, b)

- A boron concentration between 2300 and 2500 ppm, and 3)

A minimum solution temperature of 35*F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water level, and 3)

Verifying the boric acid storage tank solution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 35*F.

,,' Applicable to Unit I and Unit 2 until completion of cycle 5.

Applicable to Unit I and Unit 2 starting with cycle 6.

Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 3/4 1-11 Unit 2 - Amendment No. 55

REACTIVITY CONTROL SYSTEMS E0 RATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification-3.1.2.2 for MODES 1, 2 and 3 and one of the following borated water sources shall be OPERABLE as required by Specification 3.1.2.1 for MODE 4:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water level of 40%,

2)

A minimum baron concentration of 7000 ppm, and 3)

A minimum solution temperature of 65*F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water level of 89%,

2) a)

A minimum boron concentration of 2000 ppm, b)

"A boron concentration between 2300 and 2500 ppm, 3)

A minimum solution temperature of 35*F, and 4)

A maximum solution temperature of 100*F.

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

ACTION:

With the Boric Acid Storage System inoperable and being used as one a.

of the above required borated water sources in MODE 1, 2, or 3, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the RWST inoperable in MODE 1, 2, or 3, restore the tank to i

OPERABLE status within I hour or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With no borated water source OPERABLE in MODE 4, restore one borated c.

water source to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

,,' Applicable to Unit I and Unit 2 until completion of cycle 5.

Applicable to Unit I and Unit 2 starting with cycle 6.

{

Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 3/4 1-12 Unit 2 - Amendment No. 55

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 Indicated Reactor Coolant System (RCS) total flow rate and Fl, shall be maintained as follows for four loop operation.

a.

1)

RCS Total Flowrate 2 390,400 gpm, and 2)

- RCS Total Flowrate 2 371,400 gpm, and b.

Fl, s 1.55 [1.0 + 0.3 (1.0-P)] for OFA fuel Fl, s 1.65 [1.0 + 0.3 (1.0-P)] for VANTAGE 5 fuel where:

Measured values of Fl, are obtained by using the movable incore detectors. An appropriate uncertainty of 4% (nominal) or greater shall then be applied to the measured value of Fl, before it is compared to the requirements, and P-THERMAL POWER RATED THERMAL POWER

. APPLICABILITY: MODE 1.

ACTION:

With RCS total flow rate or Fl, outside the region of acceptable operation:

a.

Withi 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1.

Restore RCS total flow rate and Fl, to within the above limits, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

,,' Applicable to Unit I and Unit 2 starting with cycle 6. Applicable to Unit I an Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 3/4 2-8 Unit 2 - Amendment No. 55

A 4

.. ~,

a L.

n s.,

s--_

a 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:

1 a.

The isolation valve open and power removed, b.

A contained borated water level of between 31% and 63%,

c.

1)

'A boron concentration between 1900 and 2100 ppm,

2) **A boron concentration between 2200 and 2400 ppm, and d.

A nitrogen cover-pressure of between 602 and 647 psig.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within I hour 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 in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE0VIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1)

Verifying the contained borated water level and nitrogen cover-pressure in the tanks, and 2)

Verifying that each accumulator isolation valve is open.

i

~ Pressurizer pressure above 1000 psig.

' Applicable to Unit I and Unit 2 until completion of cycle 5.

,, Applicable to Unit I and Unit 2 starting with cycle 6.

Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 3/4 5-1 Unit 2 - Amendment No. 55

EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) and the heat traced portion of the RWST vent path shall be OPERABLE with:

a.

A minimum contained borated water level of 89%,

b.

1)

A minimum boron concentration of 2000 ppm,

2) **A boron concentration between 2300 and 2500 ppm, c.

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

A maximum water temperature of 100*F.

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

ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

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

Verifying the boron concentration of the water.

b.

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

c.

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

,lApplicabletoUnitIandUnit2startingwithcycle6.

Applicable to Unit I and Unit 2 until completion of cycle 5.

UNIT 1 -AMENDMENT NO. 56 BRAIDWOOD - UNITS 1 & 2 3/4 5-11 UNIT 2 -AMENDMENT N0. 55

3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 1

3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:

a.

A K,,, of 0.95 or less, or b.

1)

A boron concentration of greater than or equal to 2000 ppm.

2)

- A boron concentration of greater than or equal to 2300 ppm.

APPLICABILITY: MODE 6*.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equiv-alent until K concentration,'i's restored to greater than or equal to 2000 ppmis reduced to less t whichever is the more restrictive.

SURVEILLANCE RE0VIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a.

Removing or unbolting the reactor vessel head, and b.

Withdrawal of any full-length control rod in excess of 57 steps (approximately 3 feet) from its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.9.1.3 Valves CV1118, CV8428, CV8441, CV8435, and CV8439 shall be verified closed and secured in position by mechanical stops or by removal of air or electrical power at least once per 31 days.

'The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

,,' Applicable to Unit I and Unit 2 until completion of cycle 5.

Applicable to Unit I and Unit 2 starting with cycle 6.

I Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 3/4 9-1 Unit 2 - Amendment No. 55

REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 1

3.9.2 As a minimum, two Source Range Neutron Flux Monitors shall be OPERABLE, each with continuous visual indication in the control room and one with audible indication in the containment and control room.

APPLICABILITY:

MODE 6.

ACTION:

With one of the above required monitors inoperable or not operating, a.

immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

b.

With both of the above required monitors inoperable or not operating, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.9.2 Each Source Range Neutron Flux Monitor shall be demonstrated OPERABLE by performance of:

a.

.A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

An ANALOG CHANNEL OPERATIONAL TEST within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the initial start of CORE ALTERATIONS, and An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.

c.

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

REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions.

These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. Th transformed into the limiting MTC value -4.1 x 10',is value of the MDC was then M/k/*F. The MTC value of -3.2 x 10 M/k/'F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10 M/k/*F.

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC can be maintained within its limits.

The BOL MTC measurement combined with the predicted MTC with core burnup can be used to impose administrative limits on rod withdrawal to ensure that MTC will always be less positive then 0 M /k/*F.

This coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICAllTY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 550*F.

This limitation is required to ensure: (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, (4) the reactor vessel is above its minimum RT temperature, and (5) the plant is above the cooldown steam dump m

permissive, P-12.

3/4.1.2 BORAT10N SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each MODE of facility operation.

The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 350*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% M/k after xenon decay and cooldown to 200,*F.

The maximum expected boration capability requirement is 15,780 (13,487) gallons,of 7000-ppm borated water from the boric acid storage tanks or 70,450 (54,014) gallons of 2000-ppm (2300-ppm) borated water from the refueling water storage tank.

' Applicable to Unit I and Unit 2 starting with cycle 6.

Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 B 3/4 1-2 Unit 2 -Amendment No.

55

9 REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued)

A Boric Acid Storage System level of 40%, ensures that there is a volume of greater than or equal to 15,780 (13,487) gallons available. A RWST level of l

89% ensures that there is a volume of greater than or equal to 395,000 gallons available.

With the RCS temperature below 350'F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 330*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or an RHR Suction relief valve.

l The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% Ak This condition requires e/k after xenon decpy and cooldown from 200'F to 140*F.

ither 2,652 (740) gallons f 7000-ppm borated water ppm),the boric acid storage tanks or 1,840 (2,264),o from gallons of 2000-ppm (2300-borated water from the refueling water storage tank RWST). A Boric Acid Storage Systep level of 7% ensures there is a volume of gr(eater than or equal to 2652 (740) gallons available. An RWST level of 9% ensures there is a l

volume of greater than or equal to 38,740 gallons available.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.0 and 11.0 for the solution recirculated l

within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Baron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

The OPERABILITY of the automatic Boron Dilution Protection System ensures adequate capability for negative reactivity insertion to prevent a transient caused by the uncontrolled dilution of the RCS in MODES 3,4, and 5.

The func-tioning of the system precludes the necessity of operator action to prevent further dilution by terminating flow to the charging

)umo unborated water sources and initiating flow from the RWST.(s)The mostfrom possible restrictive condition occurs shortly after beginning of life when the critical boron concentration is highest, and a 205 gpm dilution flowrate provides the maximum positive reactivity addition rate. One reactor coolant pump in operation with all reactor coolant loop stop isolation valves open reduces the reactivity addition rate by mixing the dilution through all four reactor coolant loops. A minimum count rate of ten counts per second minimizes the impact of the uncertainties associated with the source range nuclear instrumentation.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3 Ak/k is required to control the reactivity transient. Actions taken by the microprocessor if the neutron count rate is doubled will prevent return to criticality in these MODES.

j

~ Applicable to Unit I and Unit 2 starting with cycle 6.

l Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 B 3/4 1-3 Unit 2 - Amendment No.

55

1 1

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INDICATED AX1AL FLUX DIFFERENCE FIGURE B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER BRAIDWOOD - UNITS 1 & 2 B 3/4 2-3

4 POWER DISTRIBUTION LIMITS BASES REAT FLUX HOT CHANNEL FACTOR. and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) c.

The control rod insertion limits of Specification 3.1.3.6 are maintained, and d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FL will be maintained within its limits provided the Conditions a.

througTfd. above are main 390,400 gym (371,400 gpm)}ained. The combination of the RCS flow requirement and the requirement on F,, guaradee Gat %e MM l

used in tie safety analysis will be met.

Margin between the safety analysis limit DNBRs [1.49 and 1.47 for the OFA l

fuel typical and thimble cells, respectively and 1.67 and 1.65 for the VANTAGE 5 typical and thimble cells (1.50 for the typical and thimble cells),]

and the design limit DNBRs [1.34 and 1.32 for the OFA fuel ty)ical and thimble cells, and 1.33 and 1.32 for the VANTAGE 5 fuel typicp]l and tiimble cells, respectively (1.25 for the typical and thimble cells) is maintained.

A fraction of this margin is utilized to accommodate the transition core DNBR penalty (maximum of 12.5%) and the appropriate fuel rod bow DNBR penalty (less than 1.5% oer WCAP-8691, Revision 1). The rest of the margin between design and safety analysis DNBR limits can be used for plant design flexibility.

The RCS flow requirem minimum measured flow rate of 97,600 gpm (92,850 gpm)pnt is based on the loo) which is used in Procedure (Revised Thermal Design Procedure),the ;mproved Thermal Design described in UFSAR 4.4.1 and 15.0.3.

A precision heat balance is performed once each cycle and is used to calibrate the RCS flow rate indicators.

Potential fouling of the feedwater venturi which might not be detected, could bias the results from the precision heatbalanceinanon-conservativemanner. Therefore, a penalty of 0.1% is assessed for potential fee A maximum measurement uncertainty of 2.2% (3.5%)pwater venturi fouling.has been included in the loop minimum rate to account for potential undetected feedwater venturi fouling and the use of the RCS flow indicators for flow rate verification. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters.

If detected, action shall be taken, before performing subsequent precision heat balance measurements, i.e., either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate the fouling.

Surveillance Requirement 4.2.3.4 provides adecuate monitoring to detect possible flow reductions due to any rapid core cruc buildup.

Surveillance Requirement 4.2.3.5 specifies that the measurement instrumen-tation shall be calibrated within seven days prior to the performance of the calorimetric flow measurement. This requirement is due to the fact that the drift effects of this instrumentation are not included in the flow measurement uncertainty an9'ysis.

This requirement does not apply for the instrumentation whose drift efLcts have been included in the uncertainty analysis.

Applicable to Unit I and Unit 2 starting with cycle 6.

Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 B 3/4 2-4 Unit 2 - Amendment No. 55

EMERGENCY CORE COOLING SYSTEMS i

BASES 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

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

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

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.0 and 11.0 for the solution recirculated within containment after a LOCA.

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

p UNIT 1 - AMENDMENT NO. 56 BRAIDWOOD - UNITS 1 & 2 B 3/4 5-4 UNIT 2 - AMENDMENT NO.

55

O CONTAINMENT SYSTEMS BASES CONTAINMENT PURGE VENTILATION SYSTEM (Continued) be exceeded in the event of an accident during containment purging operation.

Operation with one line open will be limited to 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months during a calendar year.

Leakage integrity tests with a maximum allowable leakage rate for containment purge supply and exhaust supply valves will provide early indica-tion of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop. The 0.60 L leakage limit of Specification 3.6.1.2.b. shall not be exceeded when the leak,ge rates a

determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment depressurization and cooling capability will be available in the event of a LOCA or steam line break. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses.

The Containment Spray System and the Containment Cooling System are redundant to each other in providing post-accident cooling of the containment atmosphere. However, the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient Na0H is added to the containment spray in the event of a LOCA.

The limits on Na0H volume and concentration ensure a pH value of between 8.0 and 11.0 for the l

solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses.

A spray additive tank level of between 78.6% and 90.3% ensures a voluma of greater than or equal to 4000 gallons but less than or equal to 4540 in hns.

UNIT 1 - AMENDMENT N0. 56 BRAIDWOOD - UNITS I & 2 B 3/4 6-3 UNIT 2 - AMENDMENT N0. 55

[DTAINMENTSYSTEMS BASES 3/4.6.2.3 CONTAINMENT COOLING SYSTEM The OPERABILITY of the Containment Cooling System ensures that: (1) the containment air temperature will be maintained within limits during normal operation, and (2) adequate heat removal capacity is available when operated in conjunction with the Containment Spray Systems during post-LOCA conditions.

The Containment Cooling System and the Containment Spray System are redundant to each other in providing post-accident cooling of the containment atmosphere.

As a result of this redundancy in cooling capability, the allowable out-of-service time requirements for the Containment Cooling System have been appropriately adjusted.

However, the allowable out of-service time require-me '

f or the Containment Spray System have been maintained consistent with t' a tigned other inoperable ESF equipment since the Containment Spray S.,.

also provides a mechanism for removing iodine from the containment atmosphere.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atinosphere or surizatior, of the containment and is consistent with the requirements of L

04 thru 57 of Appendix A to 10 CFR Part 50.

Containment isolation within the time limits specified for those isolation valves designed to close auto-matically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable lirrit during post-LOCA conditions.

Either recombiner unit (or the Purge System) is capable of controlling the expected hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment.

These Hydrogen Control Systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment following a LOCA," March 1971.

The Hydrogen Mixing Systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA.

This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

BRAIDWOOD - UNITS 1 & 2 B 3/4 6-4

Q 3/4.9 REFUELING OPERATIONS J

BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on Keff of no greater than 0.95 is sufficient to prevent reactor criticality during refueling operations and includes a 1% ok/k conservative allowance for uncertainties.

Similarly, the boron concentration value of 2000 ppm (2300 ppm) or greater includes a conservative uncertainty allowance of 50 ppm.

These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The locking closed of the required valves during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portions of the RCS. This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water.

3/4.9.2 INST.lVMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products.

This decay time is consistent with the assumptions used in the safety analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

The Braidwood Station is designed such that the containment opens into the fuel building through the personnel hatch or equipment hatch.

In the event of a fuel drop accident in the containment, any gaseous radioactivity escaping from the containment building will be filtered through the Fuel Handling Building Exhaust Ventilation System.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

' Applicable to Unit I and Unit 2 starting with cycle 6.

UNIT 1 - AMENDMENT NO. 56 BRAIDWOOD - UNITS 1 & 2 B 3/4 9-1 UNIT 2 -

AMENDMENT NO.

55

ML20078C223

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