Text

Amends 51 to Licenses NPF-37 & NPF-66,respectively & 40 to Licenses NPF-72 & NPF-77,respectively,revising TS to Reflect Changes to Current Boron Dilution Analyses
	ML20114F798
Person / Time
Site:	Byron, Braidwood
Issue date:	10/05/1992
From:	Barrett R; Office of Nuclear Reactor Regulation
To:	; Commonwealth Edison Co
Shared Package
ML20114F801	List: ML20114F798; ML20114F807;
References
	NPF-37-A-051, NPF-66-A-051, NPF-72-A-040, NPF-77-A-040 NUDOCS 9210130313
	Download: ML20114F798 (35)
	v • d • e

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UNITED STATES S

NUCLEAR REGULATORY COMMISSION

'f WASHINGTON, D.C 20655 o%.v f COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 51 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 July 28, 1992, complies with tha standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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 huthorized by tnis 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-37 is hereby amended to road as follows:

NkO3N j

P L

.~-

. (2)

Technical Specifications The Technical Specifications contained in Appendix A as re ised through Amendment No. 51 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 FEGULATORY COMMISSION

/'

(

f.N Richr4J.Barrett, Director Project Directorate III-2 Division of Reactor Projects - Ill/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Sper.i ficat ions Date of Issuance:

October 5, 1992 i

Q..

p ascq C9 y " 3,7 i.7, UNITED STATES i

NUCLEAR REGULATORY COMMISSION

'k*I[f o

WASHINGTON, D.C. 20566

%g v....f g)_MMONWEAL1H EDIS0N COMPANY E0EU. NO STN 50-455 BYRON STATION. UNIT NO. 2 6MP10 MENT TO FAClllTY OPERATING LICEN:1 Amendment No. 51 License No. NPF-66 1.

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

A.

The application for amendment by Commonwealth Edison Compsny (the licensee) dated July 28, 1992, 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 condutted 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 indicateJ in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follos.s:

! (2) lechnical Specifications The Technical Specifications contained in Appendix (NUREG-lll3),

a> revised through Amendment No. 51 and revised by 8 tachment 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 f acility 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 4-(,/

Richard arrett, Director g

Project Directorate 111-2 Division of Reactor Projects - lil/lV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 5, 1992

ATTACHMENT TO LICENSE AMENDMENT NOS. 51 AND 51 FAClllTY OPERATING LICENSE NOS. NPF-37 AND NPF-66 QOCKET NOS, STN 50-454 AND STN 50-4M 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 nu:nber and contain marginal lines indicating the area of change.

Pages identified by an asterisk are provided' for convenience.

Remove Pages Insert Pages

l11

111 IV IV 3/4 1-3 3/4 1-3/4 1-13a 3/4 1-13b 3/4 3-2 3/4 3-2 3/4 3-5 3/4 3-5 3/4 3-9 3/4 3-9 3/4 3-12a 3/4 3-12a B 3/4 1-1 B 3/4 1 -

B 3/4 1-3 B 3/4 1-3 8 3/4 1-4 8 3/4 1-4

SAFETY-LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 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 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...............

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

2-4 l

t i

l l

l BASES l

t i

SECTICN PAGE l

l 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 L

l l

BYRON - UNITS 1 & 2 III l

LIMITING CONDITIONS FOR OPERATION-AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY......................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS i

3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg > 200*F...........................

3/4 1-1 Shutdown Margin - T,yg < 200 F...........................

3/4 1-3 Moderator Temperature Coefficient........................

3/4 1-4 Minimum Temperature for Criticali ty......................

3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Flow Path - Shutdown.....................................

3/4 1-7 Flow Paths - Operating...................................

3/4 1-8 Charging Pump -_ Shutdown................................

3/4 1-9 Charging Pumps - Operating...............................

_3/4 1-10 Borated Water Source - Shutdown..........................

3/4 1-11 l

Boraied Water Sources - Operating........................

3/4 1-12 Boron Dilution Protection System.........................

3/4 1-13a l

l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height.............................................

3/4 1-14 l

TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE l'

EVENT OF AN IN0PERABLE FULL-LENGTH R00..............

3/4 1-16 l

Position Indication Systems - Operating..................

3/4 1-17 Position Indication System -

Shutdown....................

3/4 1-18 Rod Drop Time............................................

3/4 1-19 Shutdown Rod Insertion Limit.............................

3/4 1-20 Control Rod Insertion Limits.........................._...

3/4 1-21 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION...........................

3/4 1-2E l-e l-BYRON - UNITS-1 & 2 IV Amendment No. 51

~

REACTIVI'iY CONTROL SYSTEMS SHUTOOWN MARGIN - T

< 200*F 3yg LIMITING CONDITJON FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k.

l APPLICABILITY:

MODE 5.

ACTION:

a.

With the SHUTDOWN MARGIN less than 1.3% Ak/k declare both Boron Dilution Protection System subsystems inoperable and apply Specifi-cation 3.1.2.7.b.

b.

With the SHUTOOWN MARGIN less than 1% Ak/k, immediately 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 equivalent until the SHUTOOWN MARGIN is restored to greater than or equal to 1% Ak/k.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1% Ak/k:

a.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod (s) is immovable'or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1)

Reactor Coolant System boron concentration, 2)

Control rod position, 3)

Reactor Coolant System average temperature, 4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

BYRON - UNITS 1 & 2 3/4 1-3 Amendment No. 51

REACTIVITY CONTROL SYSTEMS BORON DILUTION PROTECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.1.2.7 Two independent Boron Dilution Protection System (BDPS) subsystems shall be OPERABLE.*

APPLICABILITY:

MODES 3, 4, and 5.

ACTION:

a.

With one BDPS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or within the next hour, and at least once every 31 days thereaf ter, verify valves CV-111B, CV-8428, CV-8439, CV-8441, and CV-8435 are closed and secured in position.**

b.

With both BDPS subsystems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter:

1.

Verify valves CV-111B, CV-8428, CV-8439, CV-8441, and CV-8435 are closed and secured in position **, and 2.

Verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable.

(

i l

The BDPS Flux Doubling signals may be blocked daring reactor startup.

- These valves may be opened on an intermittent basis under administrative I

control when required to support plant evolutions.

l BYRON - UNITS 1 & 2 3/4 1-13a AMENDMENT NO. 51 l

o REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.2.7 Each BDPS subsystem shall be demonsts ted OPERA 8LE:

a.

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

1.

Verifying that its associated nuclear instrumentation source range detector is OPERABLE and indicating greater than or equal to 10 counts per second, 2.

Verifying that all reactor coolant loop stop isolation valves are open, and 3.

Verifying that at ~aast one reactor coolant pump is in operation, b.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position, c.

At least once per 92 days by verifying that the BDPS Alarm Setpoint is less than or equal to an increase of twice the count rate within a 10-minute period.

d.

At least once per 18 months when shutdown by verifying that on a simulated BDPS Flux Doubling test signal valves CV-1120 and CV-112E open and valves CV-1128 and CV-112C close in less than or equal to 30 seconds, l

BYRON - UNITS 1 & 2 3/4 1-13b AMENDMENT N0. 51 1

TABLE 3.3-1 Yg REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM b

TOTAL NO.

"HANNELS CHANNELS APPLICABLE d

FUN 2TICNAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1.

Mr. sal Reactor Trip 2

1 2

1, 2 1

w 2

1 2

3*,

4*, 5*

10 g.

m 2.

Fower Range, Nr 'ron Flux a.

High Set-4 2

3 1, 2 2

b.

Lew Setw.nt 4

2 3

1###, 2 2

3.

Power Range, Neutron Flux 4

2 3

1, 2 2

High Positive Rate 2

4.

Power Range Neatron Flux, 4

2 3

1, 2 2

w High Negative Rate m

5.

Intermediate Range, Neutron Flux 1

2 1###, 2 3

6.

Source Range, Neutron Flux a.

Startup 2

1 2

2##

4 I

b.

Shutdor 2

1 2

3,4,5 5

7.

Overtegerature AT 4

2 3

1, 2 6

8.

Overpower AT 4

?

3 1, 2 6

2, 3

9.

Pressurizer Pressure-Low g

(Above P-7) 4 2

3 1

6***

E 5

h

TABLE 3.3-1 (Continued)

TABLE NOTATIONS

With the Reactor Trip System breakers in the closed position and the Control Rod Drive System capable of rod withdrawal.

l

- - These channels also provide inputs to ESFAS.

The Action Statement for the channels in Teble 3.3-3 is more conservative and, therefore, controlling.

1. Below the P-6 (Intermediate _ Range Neutron Flux Interlock) Setpoint.

1. 1. Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

@Whenever the Reactor Trip Bypass Breakers are racked in and closed for bypass-ing a Reactor Trip Breaker.

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimem Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT-STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inopeirable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1; and c.

Either, THERMAL POVER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATJO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.

ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

Below the P-6 (Intermediate Range Neutron Flux Interlock)-

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint; and b.

Above the P-6 (Intermediate Range Neutron Flux irlock)

Setpoint but below 10% of RATED THERMAL POWER

ore the inoperable channel to OPERABLE'$tatus prior i reasing THERMAL POWER above 10% of RATED THERMAL POWEk.

BYRON - UNITS I & 2 3/4 3-5 AMENDMENT NO. 51

TABLE 4.3-1 e%3 REACTOR "If a.a TEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS I

TRIP c-25 ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL C'%NNEL OPERATIONAL OPERATIONAL AC TUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED o=

~

1.

Manual Reactor Trip N.A.

N.A.

R(14)

N.A.

1, 2, 3 *, 4 *, 5

2.

Power Range, Neutron Flux a.

High Setpoint S

D(2, 4),

Q N A.

N.A.

1, 2 M(3, 4),

Q(4, 6),

R(4,g5a)#

b.

Low Setpoint 5

R(4)

Q N.A.

N.A.

1### 2 3.

Power Range, Neutron Flux, fi. A.

R(4)#,

Q N.A.

N.A.

1, 2 s"

[

High Positive Rate 4.

Power Range, Neutron Flux, N.A.

R(4)#

Q N.A.

N.A.

1, 2 High Negative Rate S.

Intermediate Range, S

R(4, Sa)"

Q N.A.

N.A.

I###, 2 Neutron Flux 6.

Source Range, Nc1 tron Flux 5

R(4, Sb)#

Q(9)

N.A.

2##, 3, 4, 5 7.

Overtemperatt're AT S

R(13)#

Q N.A.

N. A.

1, 2 g

m h

8.

Overpower AT S

R Q

N.A.

N. A.

1, 2 o

5 9.

Pressurizer Pressure-Low S

R Q**

N.A.

I g

(Above P-7) 10.

Pressurizer Pressure-High 5

R Q

N.A.

1, 2 11.

Pressurizer Water Level-High 5

R Q

N.A.

1 (Above P-7)

-_______.um.

~ $

l l

TABLE 4.3-1 (Continued)

TABLE NOTATIONS (9) Surveillance in H0 DES 3*, 4", and 5* shall also include verifi.*ation that permissives P-6 and P-10 are in their required state for exist rg plant conditions by observation of the permissive annunciator window.

(10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall be performed such that each train is tested at least every 62 days on a STA.3GERED TEST BASIS and following maintenance or adjustment of the Reactor Trip Breakers and shall include independent verification of the OPERABILITY of the Undervoltage and Shunt Trip Attachments of the Peactor Trip Break'rs.

(12) Not Used.

(13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

(14) Verify that the appropriate signals reach the Undervoltage and Shunt Trip Relays, for both the Reactor Trip and Bypass Breakers from the Manual Trip Switches.

(15) Manual Shunt Trip prior to the Reactor Trip Bypass Breaker being racked in and closed for bypassing a Reactor Trip Breaker.

(16) Automat s Undervoltage trip.

e BYRON - UNITS 1 & 2 3/4 3-12a AMENDMENT NO. 51

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUT 00WN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated e:cident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subtritical to preclude inadvertent criticality in the shutdown condition.

SHUTOOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg.

The most restrictive condition occurs u EOL, with T,yg at no load operating temperature, and is associated with e postulated steam line break accident and resulting uncon-trolled RCS cooldown.

In the analysis of this accident, a minimum SHUT 00VN MARGIN of 1.3% ak/k is required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions, With T,yg less than 200 F, the reactivity transients resulting from a postulated steam line break cooldown are minimsl and a 1% Ak/k SHUTDOWN MARGIN provides adequate protection provided that boration dilution paths are isolated.

A 1.3% ok/k SHUTDOWN MARGIN is required to ensure the OPERABILITY of the automatic Boron Dilution Protection System.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC value:, at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

BYRON - UNITS 1 & 2 B 3/4 1-1 Amendment No. 51

RE;A,CTIVITY CONTROL SYSTEMS BASES BORAT10N SYSTEMS (Continued)

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

A RWST level of 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 floron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional resttictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injec' lon System becomes incperable.

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 tn be inoperable balow 330'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or an RHR Suction valve.

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

This condition requires either 2,652 gallons of 7000 ppm borated water from the boric acid storage tanks or 11,840 gallons of 2000 ppm borated eter 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 2652 gallons available.

An PWST level of 9% ensures there is a volume of greater than nr equal to 38,740 gallons available.

The contained water volume limits include allowance for water not a'711able 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 cf between 8.5 and 11.0 for the solution recirculated within containment after s LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical syste.es 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 autcmatic Boron Dilution Protection System ensures adequate capability for negative reactivity insertion to prevent a transient caused by the uncontrolled dilution of tl RCS in MODES 3,4, and 5.

The func-tioning of the system precludes the necessity of operator action to prevent fur-ther dilution by terminating flow to the charging pump (s) from possible unborated water sources and initiating flow from the RWST.

The most restrictive conCition occurs tihortly 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 thr3 ugh 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 ok/k is required to control the reacthity tran-sient.

Actions taken by the microprocessor if the neutron count rate is doubled will prevent return to criticality in these MODES.

BYRCN - UNITS 1 & 2 B 3/4 1-3 AMENDMENT NO. 51

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution limits are maintained, (2) the minieum SHUTDOWN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated accident analyses are limited.

OPERABILITY-of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

Verification that the Digital Rod Position Indicator agrees with the demanded position within

.t12 steps at 24, 48, 120 and 228 steps withdrawn for the Control Banks and 18,210,and228stepswIthdrawnfortheShutdownBanksprovidesassurances that the Digital Rod Position Indicator is operating correctly over the full range of indication.

Since the Digital Rod )osition System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points 1

in the indicated ranges are picked for verification of agreement with demanded position.

The ACTION statements which permit limited vailations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met.

Inoperability or misalignment of a single, trippable rod requires measurement of peaking factors and a restriction in THERMAL POWER.

These restrictions provide ai,surance of fuel rod integrity during continued operation.

In addition, those safety analyses affected by an inoperable or misaligned rod are reevaluated to confirm that the results remain valid during future operation. With multiple inoperable or misaligned, but trippable, rods; alignment of the remaining rods in the bank (s) to within

+ 12 steps of the inoperable rods, and restriction in THERMAL POWER assures Tuel rod integrity during continued operation.

For Specification 3.1.3.1 ACTIONS b. and c., it is incumbent un the plant to confirm tripability of the inoperable rod (s).

This confirmation may be, for example, by serification of a control system failure, usually electrical in nature (such is an Urgent Failure Alarm), or that the failure is associated with the cont'ol rod stepping mechanism.

In the event the plant is unable to verify the rrd(s) trippability, it must be assumed to be untrippable and thus falls under he requirements of ACTION a.

The rnximum rod drop time restriction is consistent with the assumed rod drop timr used in the safety analyses.

Measurement with T,yg greater than or equal t9 550 F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced C :ng a Reactor trip at operating conditions.

Control rod positions ar.d OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if a rod position deviation monitor is inoper-able.

These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.

BYRON - UNITS 1 & 2 B 3/4 1-4 AMENDMENT NO. 51

pa erog.

t UNITED SYATES c,

Y i

NUCLEAR REGULATORY COMMISSION f

WASHINGTON. D.C. 20565 c,

%,4..e*'/

QMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 40 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 July 28, 1992, complies with the ;tandards 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 Coinmi ss ion; C.

There is reasonable assurance (i) that the activitics authorized by this amendment can be conducted without endangerir.g 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 amendme-+ will not be inimical to the common defense and security or to tne health and safet.' 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 changet 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-72 is hereby i

amended to read as follows:

l

.. _ - _ _. ~. _ _ _ _ _ - - _ _ _..

.-.__m-__

i (2)

It.chnical Sytcifications The Technical Specifications contained in Appendix A as revised through Amendment No. 40 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 Y

)//ph{f,j

/

Richu J Barrett, Director Project Directorate 111-2 Division of Reactor Projects - !!!/IV/V Office of kuclear Reactor Regulation

Attachment:

Changes to the lechnical Specifications Date of issuance:

October 5, 1992 l

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, y, c UNITED STATES n

i NUCLEAR REGULATORY COMMISSION

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[0MMONWEALTH EDISSN COMPANY 00CKET 10. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACillTY OPERATING LICENSE Amendment No. 40 License No. NPF-77 1.

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

A.

The application for amendment by Connonwealth Edison Company (the licensee) dated July 28, 1992, 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 i

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 authori:ed by this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in compliance with the Commission's regulations; D.

The '.ssuance 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. NPT-77 is hereby amended to read as follows:

1 I

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. (2) lechnical Specifications The Technical Specifications contained in Appendix A as revised

+

through Amendment No. 40 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 lictnse.

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 i

(

JA.n t Richar J.

Project Dir,Rarrett, Director ectorate 111-2 Division of Reactor Projects - lil/lV/V v

office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 5, 1992 l-

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---_ _ - _.._..____. _.. _. _ _ _. _. _ _. _ _ -... _. ~ _

ATTACHMENT TO LICENSE AMENDMENT NOS. 40 Afl0 40 s

i FAtlllTY OPERATING LICENSE NOS. NPF-72 AND NPf-77 DQ{ JET NQE. STN 50-456 AND STN 50-457 t

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 identified by an asterisk are provided for convenience.

~

Eg. move Paaes insert Paaes Ill 111 IV IV t

3/4 1-3 3/4 1-3 3/4 1-13a 3/4 1-13b 3/4 3-2 3/4 3-2 3/4 3-5 3/4 3-5 3/4 3-9 3/4 3-9 3/4 3-12 3/4 3-12 3/4 3-12a 3/4 3-12a B 3/4 1-1 B 3/4 1-1 B 3/4 1-3 B 3/4 1-3 B 3/4 1-4 B 3/4 1-4 i

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SAFETV LIMITS AND LIMITING SAFETY SYTTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMIT,5

?.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 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 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 SETF0lNTS...........

B 2-3 BRAIDWOOO - UNITS 1 & 2 III

9 LIMITINGCONDITIONS-FOROPERATIONAN;,SURVEILLANCEREQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - Tavg > 200 F.......................

3/4 1-1 Shutdown Margin - T,yg i 200*F...........................

3/4 1-3 Moderator Temperature Coefficient........................

3/4 1-4 Minimum Temperature for Criticality......................

3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path -

Shutdown.....................................

3/4 1-7 Flow Paths - Operating...................................

3/4 1-8 Charging Pump - Shutdown.................................

3/4 1-9 Charging Pumps - Operating...............................

3/4 1-10 Borated Water Source - Shutdnwn..........................

3/4 1-11 Borated Water Sources - 0perating........................

3/4 1-12 Boron Dilution Protection System.........................

3/4 1-13a 3/4.1. 3 MOVABLE CONTROL ASSEMBLIES Group Height.............................................

3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENi 0F AN INOPERABLE FULL-LENGTH R00..............

3/4 1-16 Position Indication Systems - Operating..................

3/4 1-17 Position Indication System - Shutdown.............

3/4 1-18 Rod Drop Time...................

3/4 1-19 Shutdown Rod Insertion Limit.............................

3/4 1-20 Control Rod Insertion Limits.............................

3/4 1-21 FIGURE 3.1-1 R0D BANK INSERTION LIMIIS VERSUS THERMAL POWER FOUR LOOP 0PERATION...........................

3/4 1-22 BRAIDWOOD - UNITS 1 & 2 IV Amendment No. 40

REACTIVITY CONTROL SYSTEMS SHUTD0 'N MARGIN - T,yg < 200*F LIMITING CONDITION FOR OPERAT*0N 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% Ak/k.

l APPLICABILITY:

MODE 5.

ACTION:

a.

With the SHUTDOWN MARGIN less than 1.3% Ak/k declare both Boron Dilutinn Protection System subsystems inoperable and apply Specifi-cation 3.1.2.7.b.

b.

With the SHUTDOWN MARGIN less than 1% Ak/k, immediately 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 equivalent until the SHUTDOWN MARGIN is restored to greater than er equal to 1% Ak/k.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1% ak/k:

a.

Within I hour after detection of an inoperable control rod (s) and at l

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod (s) is immovable or unttippable, the SHUTOOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and l

l b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1)

Reactor Coolant System boron concentration, 2)

Control rod position, 3)

Reactor Coolant System average temperature, 4)

Fuel burnup based on gross thermal energy generation, 5)

Xenon concentration, and 6)

Samarium concentration.

BRAIDWOOD - UNITS 1 & 2 3/4 1-3 Amendment No. 40

REACTIVITY CONTROL SYSTEMS BORON DILUTION PROTECTION SYSTEM LIMITING CONDITION FOR OPERATION 3.1.2.7 Two independent Boron Dilution Protection System (BDPS) subsystems shall be OPERABLE.*

APPLICABILITY:

MODES 3, 4, and 5.

ACTION:

a.

With one BDPS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or within the next hour, and at least once every 31 days thereafter, verify valves CV-1118, CV-8428, CV-8439, CV-8441, and CV-8435 are closed and secured in position.**

b.

With both BDPS subsystems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter:

1.

Verify valves CV-1118 CV-8428, CV-8439, CV-8441, and CV-8435 are closed and secured in position **, and i

2.

Verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable.

+

The BDPS Flux Doubling signals may be blocked during reactor startup.

- These valves may be opened sn an intermittent basis under cdministrative control when required to support plant evolutions.

BRAIDWOOD - UNITS 1 & 2-3/4 1-13a Amendment No. 40

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.. __ -. ~ - - _

a REACTIVITY __ CONTROL SYSTEMS SURVC; iANCE REQUIREMENTS 4.1.2.7 Each BDPS subsystem 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 that its associated nuclear in;trumentation source range detector is OPERABLE and indicating greater than or equal to 10 counts per second.

2.

Verifying that all reactor coolant loop stop isolation valves are open, and

3..

Verifying that at least one reactor coolant pump is in operation, b.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position, c.

At least once per 92 days by verifying that the BDPS Alarm Setpoint is less than or equal to an increase of twice the count rate within a 10-minute period, d.

At least once per 18 months when shutdown by verifying that on a simulated BDPS Flux Doubling test si0nal valves CV-1120 and CV-112E open and valves CV-1128 and CV-112C close in less than or equal to 30 seconds.

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BRAIDWOOD - UNITS 1 & 2 3/4 1-13b Amendment No. 40

i TABLE 3.3-1

'g REACTOR TRIP SYSTEM INSTRUMENTATION is8 MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE

' FUNCTIONAL UNIT OF CHANNELS TO TRIP OPEP.ABLE MODES ACTION c_5g 1.

Manual Reactor Trip 2

1 2

1, 2 1

~

2 1

2 3*, 4*, 5*

10 w

2..

Power Range, Neutron Flux m

a.

High Setpoint 4

2 3

1, 2 2

b.

Low Setpoint 4

2 3

1###, 2 2

3.

Power Range, Neutter. Flux 4

2 3

1, 2 2

High Positive Rate 4.

Power Range, Neutron

'ua, 4

2 3

1, 2 2

High Negative Rate w

E 5.

.retermediate Range, Neutron Flux 2

1 2

1###, 2 3

6.

Source Range, Neutron Flux a.

Startup 2

1 2

2#F 4

l b.

Shutdown 2

1 2

3,4,5 5

7.

Overtersperature AT 4

2 3

1, 2 6

8.

Overpower AT 4

2 3

1, 2 6

i

>g 9.

Pressurizer Pressure-Lcw g

(Above P-7) 4 2

3 1

6***

3a i

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TABLE 3.3-1 (Continued)

TABLE NOTATIONS

With the Reactor Trip System breakers in the closed position and the Control Rod Drive System capable ef rod withdrawal.

l

- - These channels also provide inputs to ESFAS.

The Action Statement for the channels in Table 3.3-3 is more conservative and, therefore, controlling.

1. Below the P-6 (Intermediate Rangt Neutron Flux Interlock) Setpoint.

1. 1. Below the P-10 (Low Setpoint Power Range Neut on Flux Ir.tcrlock) Setpoint.

@Whenever the Reactor Trip Bypass Breakers are racked in and closed for bypass-ing a Reactor Trip Preaker.

ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement,' restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following cond:Uons are satisfied:

a.

The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be byparsed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months fur surveillance testing of other channels per Specification 4.3.1.1; and c.

Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and tb3 Power Range Neutron Flux Trip Setooint is redL.ed to it;s than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADHANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.

ACTION 3 - With the number of channels OPERABLE one less than the Minimum l

Channels OPERABLE requirement and with the THERMAL POWER level:

I a.

Below the P-6 (Intermediate Range Neutron Flux Interlock) l Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P Setpoint; and b.

Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

BRAIDWOOD - UNITS 1 & 2 3/4 3-5 AMENOMENT NO. 40

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G REACTOR TRIP SYSTEH INSTRUMENTATION SURVEILLANCE REQUIREHENTS

'8

. '8 TRIP ANALOG ACTUATTNG MODES FOR 4

.c-CHANNEL DEVICE WHICH

S CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE y

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED 1.

Manual Reactor Trip N.A.

N.A.

R(14)

N.A.

1, 2, 3*, 4*, 5*

ru 2.

Power Range, Neutron Flux a.

High Setpoint S

D(2, 4),

Q N.A.

M.A.

1, 2 M(3, 4)

Q(4, 6),

R(4, Sa)#

b.

Low Setpoint S

R(4)#

Q N.A.

N.A.

1###, 2 l

l

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Power Range, Neutron Flux, N.A.

R(4)#

Q N.A.

N.A.

1, 2 High Positive Rate

.w

.E 4.

Power Range, Neutron Flux, H.A.

R(4)#

Q N.A.

N.A.

1, 2 High Negative Rate 5.

Intermediate Range, S

R(4, Sa)#

Q N.A.

N.A.

1###, 2 Neutron Flux.

6.

Source Range, Neutron Flux 5

R(4, 5b)#

0(9)

N.A.

2##, 3, 4, 5 l

7.

Overtemperature AT S

R(13)#

Q N.A.

N.A.

1, 2

E 8.

Overpower AT S

R#

Q N.A.

N.A.

1, 2 9.

Pressurizer Pressure-Low S

R#

Q**

N.A.

ii. A.

1 (Above P-7)

.g 10.

Pressurizer Pressure-High S

Rf Q

N.A.

1, 2 i

11.

Pressurizer Water Level-High 5

R#

Q N.A.

N.A.

1 1

(Above P-7) i

1 m.

_s.,

TABLE 4.3-1 (Conting:ij TABLE NOTATIONS

With the Reactor Trip System breakers closed and the Control Rod Drive System capable of rod withdrawal.

- These channels also provide inputs to ESFAS.

The Operatiool Test Frequen:y for these channels in Table 4.3 2 is more conservative and, therefore, controlling.

H Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

M Below P-10 (Low Setpoint Power Range Neutror Flux Interlock) Setpoint.

(1) If not performed in previous 7 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERPAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%.

The provisions of Spect-fication 4.0.4 are not applicable for entry into H0DE 2 or 1.

(3) The initial single point comparison of incore to excore AXIAL FLUX DIFFERENCE following a refueling outage shall be performed prior to exceeding 75% of RATED THERMAL Power.

Otherwise the single point com-parison of incore to excore AXIAL FLUX DIFFERENCE shall be performed above 15% of RATED TiiERHAL POWER.

Recalibrate if the absolute differen:e is greater than ot equal to 3%.

The provisions of Specification 4.0.4 are not applicable for entry into H0DE 2 or 1.

For the purposes of this surveillance, monthly shall mean at least once per 31 EFPD.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time provisions of Specification 4.0.3 are not applicable.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(Sa) Initial plateau curves shall be measured for eacil detector.

Subsequent plateau curves shall be obtained, evaluated and compared to the initial curves.

For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(5b) With the high voltage setting varied as recommended by the manufacturer, an initial discriminator bias curve shall be measured for each detector.

Subsequent discriminator bias curves shall be obtained, evaluated and compared to the initial curves.

(6)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

The provi-sions of Specification 4.0.4 are not applicable for entry into HDDL 2 or 1.

For the purposes of this surveillance, quarterly shall mean at least once per 92 EFPD.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(8) With power greater than or equal to the interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the inter-lock is in the required state by observing the permissive annunciator window.

(9) Surveillance in H0 DES 3*, 4*, and $* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

BRAIDWOOD - UNITS 1 & 2 3/4 3-12 AMENDMENT NO. 40

)

TABLE 4.3-1 (Continued),

~iABLE NOTATIONS (10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall be performed suct* that each train is tested at least everv 62 days on a STAGGERED TEST BASIS and following maintenance or adjustment of the Reactor Trip Breakers and shall include independent verification of the OPERABILITY of the Undervoltage and Shut Trip Attachments of the Reactor Trip Breakers.

(12) Not Used.

(13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

(14) Verify that the appropr: ate signals reach the Undervoltage and Shunt Trip relays, for both the Reactor Trip and Bypass Breakers from the Manual Trip Switches.

(15) Manual Shunt Trip prior to the Reactor Trip Bypass Breaker being racked in and closed by bypassing a Reactor Trip Creaker.

(16) Automatic undcrvoltage trip.

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BRAIDW9OD - UNI 15 1 & 2 3/4 3-12a Amendment No. 40 T

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?/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made I

suberitical from all operating conditions, (2) the reactivity transients asso-4 ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactur will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boran concentration, and RCS T,yg.

The most restrictive.

condition occurs at EOL, with T,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown.

In the analvsis of this accident, a min! mum SHUTDOWN MARGIN of 1.3% ok/k is required to ;ontrol the reactivity transient.

Accordingly, the SHUTOOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,yg less than 200*F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% Ak/k SHUTDOWN MARGIN provides adequate protection provided that boration dilution paths are isolated.

A 1.3% ak/k SHUTDOWN MARGIN is required to ensure the OPERABILITY of t'r automatic Boron Dilution Protection System.

3/4.1.1.3 MODERAIOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and *.nsient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accoro.ngly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit en accurate comparison.

BP.AIDWOOD - UNITS 1 & 2 B 3/4 1-1 Amendment No. 40

9 REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTCH_S (Continued) i A Boric Acid Storage System level of 40% ensures that there is a volume of i

greater than or equal to 15,780 gallons available.

A RWST level of 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 restrictio..a prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximt.m 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 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 2,652 gallons of 7000-ppm borated water trom tne boric acid storage tanks or 11,840 gallons of 2000 ppm borated water from the refueling water storage tank (RWST).

A Boric Acid Storage System level of 7% ensures there is a volute of greater than or equal to 2652 gallons availeble.

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 8.5 and 11.0 for the solution recirculated within containment after a LOCA, This pH band minimizes the evolution of iodine and einimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of cae Baron Injection S/ stem during REFUELING ensures that this system is available for reactivity control whilt 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-tion %g of the system precludes the necessity of operator action to prevent fur-ther,ilution by terminating flow to the charging pump (s) from possible unborated water sources and initiating flow from the RWST.

Tne 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 coolant pump in ope ction with all reactor coolant loop stop isolation valves open reduces the reacthity addition rate by mixing the dilution through all four reactor coolant loops.

A minimum count rate of ten counts per se ond minimizes the impact of the uncertainties associated with the source rangi nuclear instrumentation.

In the analysis of this accident, a minimum SHUTDOWN MDQN of 1.3 Ak/k is required to control the reactivity tran-sient.

Actions taken by the microprocessor if the neutron count rate is doubled will prevent return to criticality in these MODES.

BRAIDWOOD - UNITS 1 & 2 B 3/4 1-3 Amendment No. 40 i

n-_.,.

3 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distribution Ifmits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated accident analyses are limited.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rnd alignment and insertion limits.

Verification that the Digital Rod Position Indicator agrees with the demanded position within 112 steps at 24, 48, 120, and 228 steps withdrawn for the Control Banks and 18, 210, and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication.

Since the Digital Rod Position System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for verification of agreement with demanded position.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met.

Inoperability or misalignment of a single, trippeble rod requires measurement of peaking factors and a restriction in THERMAL POWER.

These ir.trictions provide assurance of fuel rod integrity during continued operation.

In addition, those safety analyses affected by an inoperable or misaligned rod are reevaluated to confirm that the results remain valid during future operation.

With multiple inoperable or misaligned, but trippable, rods; alignment of the remaining rods in the bank (s) to within

+ 12 steps of the inoperable rods, and restriction in THERMAL POWER assures Tuel rod integrity during continued operation.

For Specification 3.1.3.1 ACTIONS b. and c., it is incumbent on the plant to confirm trippability of the inoperable rod (s).

This confirmation may be, for example, by verification of a contrel system failure, usually electrical in nature (such as an Urgent failure Alarm), or that the failure is associated with the control rod stepping mechanism.

In the event the plant is unable to verify the rod (s) trippability, it must be assumed to be untrippable and thus falls under the requirements of ACTION a.

fhe maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses.

Measurement with T,yg greater then or equs1 to 550 F and with all reactor coolant pumps-operating ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if a rod position deviation monitor is inoper-able.

These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.

BRAIDWOOD - UNITS 1 & 2 8 3/4 1-4 Amendment No. 40

ML20114F798

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