Text

Amends 80 to Licenses NPF-72 & NPF-77,respectively, Relocating Certain cycle-specific Parameter Limits from TSs to Operating Limits Rept
	ML20137Y066
Person / Time
Site:	Byron, Braidwood
Issue date:	04/16/1997
From:	Dick G; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20137X084	List: ML20137X078; ML20137X116; ML20137Y066;
References
	NUDOCS 9704220306
	Download: ML20137Y066 (23)
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$i UNITED STATES

_g Ij NUCLEAR REGULATORY COMMISSION 3*

WASHINGTON, D.C. enmas %

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COMONWEALTH EDISON COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO. 1

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AMENDMENT T0 1ACILITY OPERATING LICENSE Amendment No. 80 f

License No. NPF-72 i

1.

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

i c

i A.

The application for amendment by Commonwealth Edison Company (the licensee) dated December 21, 1995, as supalemented on October 24, 4

1996, and March 24, 1997, complies with tie standards and requirements of the Atomic Energy Act'of 1954, as amended (the i

Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; l

B.

The facility will operate in conformity with the application, the i

provisions of the Act, and the rules and regulations of the i

Commission; C.

There is reasonable assurance (1) 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 j

i conducted in compliance with the Commission's regulations D.

The issuance of this amendment will not be inimical to the common i

defense and security or to the health and safety of the public-l and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable L

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-72 is hereby l

amended to read as follows:

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1-I 1

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,q 7 9704220306 970416 ADOCK050004y4 PDR P

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. (2)

Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 80 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.

1 3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION O

Geo F. Dick, Senior Project Manager Project Directort.te III-2 Division of Reactor Projects - III/IV l

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 16, 1997

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UNITED STATES g

4 g

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 001

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COMONWEALTH EDISON COMPANY DQ(KET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 80 j

License No. NPF-77 1.

The Nuclear Regulatory Commission (the Comission) has found that.

4 A.

The application for amendment by Commonwealth Edison Company (the licensee) dated December 21, 1995, as supplemented on October 24, 1996, and March 24, 1997, 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 l

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 Comission; 1

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 i

conducted in compliance with the Commission's regulations; D.

The issuance of this amenoment 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-1 cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2)ofFacilityOperatingLicenseNo.NPF-77ishereby amended to read as follows:

+

. )

i (2)

Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 80 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 4

Plan.

3.

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

shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION i

Geo Dick, Senior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 16, 1997

)

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ATTACHMENT TO LICENSE AMENDMENT NOS. 80 AND 80 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-71 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.

Remove Paaes Insert Paaes IV IV V

V l-4 1-4 3/4 1-14 3/4 1-14 3/4 1-15 3/4 1-15 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 3/4 1-22 3/4 1-22 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 1

3/4 2-4 3/4 2-4 i

3/4 2-5 3/4 2-5 3/4 2-8 3/4 2-8 B 3/4 2-2 B 3/4 2-2 6-22 6-22 6-22a 6-22a l

6-22b e

t LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

l l

SECTION PAGE i

1 l

'3/4.0 APPLICABILITY................................................

3/4 0-1 i

i 3/4.1 REACTIVITY CONTROL SYSTEMS

]

l 3/4.1.1 BORATION CONTROL Shutdown Margin - T,,>

200*F............................

3/4 1-1

)

Shutdown Margin - T,,s 200*F............................

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

3/4 1-4 FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT VERSUS POWER LEVEL.........................................

3/4 1-Sa Mi nimum Temperature for Cri tical i 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 Borated Water Sources - Operating....................

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 EVENT OF AN INOPERABLE FULL-LENGTH R0D..............

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 l

FIGURE 3.1-1 (THIS FIGURE IS NOT USED).............................

3/4 1-22 l

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BRAIDWOOD - UNITS 12 IV Amendment No. 80 l

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e I-l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS l

i 4;

SECTION E8SI 4

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3/4.2 POWER DISTRIBUTION LIMITS

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3/4.2.1 AXIAL FLUX DIFFERENCE....................................

3/4 2-1 l

FIGURE 3.2-1 (THIS FIGURE IS NOT USED)............................

3/4 2-3 1

3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R.............................

3/4 2-4 FIGURE 3.2-2 (THIS FIGURE IS NOT USED)............................

3/4 2-5 4

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R.................................................

3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATI0...............................

3/4 2-10 3/4.2.5 DNB PARAMETERS...........................................

3/4 2-13 TABLE 3.2-1 DNB PARAMETERS........................................

3/4 2-14 j

3/4.3 INSTRUMENTATION l

3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

3/4 3-1 l

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION...................

3/4 3-2 TABLE 3.3-2 (THIS TABLE IS NOT USED)..............................

3/4 3-7 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE l

REQUIREMENTS........................................

3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTL'% TION SYSTEM INSTRUMENTATION........................................

3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.....................................

3/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0!NTS......................

3/4 3-23

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TABLE 3.3-5 (THIS TABLE IS NOT USED)..............................

3/4 3-30 i

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........

3/4 3-34 i

1 BRAIDWOOD - UNITS 1 & 2 V

Amendment No. 80 j

4 4

4

.s 4 -

DEFINITIONS j

i 0FFSITE DOSE CALCULATION MANUAL i

l.18 The 0FFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous. and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-l-

mental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Sections 6.8.4.e and f, and (2) descriptions of the i

information that should be included in the Annual Radiological Environmental i

Operating and Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7.

L OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or I

have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, i.

cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATING LIMITS REPORT 1.19.a The'0PERATING LIMITS REPORT (OLR) is the unit-specific document that l

provides operating limits for the current operating reload cycle'. These

{

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

Plant operation within these operating i

limits is addressed in individual specifications.

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE.(i.e., MODE) shall correspond to any one inclusive

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combination of core reactivity condition, power level, and average reactor 1

coolant temperature specified in Table 1.2.

l i

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' E.

I 1.20.a P shall be the maximum calculated primary containment pressure (44.4 psi,) for the design basis loss of coolant accident.

g PHYSICS TESTS i

1.21. PHYSICS TESTS shall be those tests performed to measure the fundamental j

nuclear characteristics of the core and related instrumentation: (1) described i

in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR

{

50.59, or (3) otherwise approved by the Commission.

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

BRAIDWOOD - UNITS 1 & 2 1-4 AMENDMENT NO. 80 a

_..... _._ _ _ _.._ _ _. ~

i REACTIVITY CONTROL SYSTEMS

j' 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT l

LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and

]

positioned within i 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: NODES 1* and 2*.

ACTION:

a.

With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or

}

known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within I hour and be in i

l

' HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

r i

b.

With one full-length rod trippe.ble but inoperable due to causes other than addressed by ACTION a. above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within I hour:

1.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to witnin i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Specification 3.1.3.6.

The l

THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.

The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

a)

The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.

3 b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; 6

L

' See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

BRAIDWOOD - UNITS 1 & 2 3/4 1-14 AMENDMENT NO. 80 i

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1 I

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2 l

REACTIVITY CONTROL SYSTEMS l

LIMITING CONDITION FOR MERATION ACTION (Continued) c)

A power distribution map is obtained from the movable within their limits wil(hin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; andincore detectors and F Z) a l

d)

A reevaluation of each accident analysis of Table 3.1-1 is i

performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; i

c.

With more than one full-length rod trippable but. inoperable due to causes other than addressed by ACTION a. above, or misaligned from its group step counter demand height by more than 12 steps l

(indicated position), POWER OPERATION may continue provided that:

1.

Within I hour, the remainder of the rods in the group (s) with the inoperable rods are aligned to within i 12 steps of-the 2

l inoperable rads while maintaining the rod sequence and insertion limits of Specification 3.1.3.6.

The THERMAL POWER l

1evel shall be restricted pursuant to Specification 3.1.3.6 j'

during subsequent operation, and 2.

The inoperable rods shall be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Otherwise, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

}

e i

SURVEILLANCE RE0VIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position j

deviation monitor is inoperable, then verify the group positions at least once per.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined OPERABLE by movement of at least 10 steps in any one direction at l

1 east once per 92 days.

i t

BRAIDWOOD - UNITS 1 & 2 3/4 1-15 AMENDMENT NO. 80 4

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REACTIVITY CONTROL SYSTEMS 4

l SHUTDOWN ROD INSERTION LIMIT i

LIMITING CONDITION FOR OPERATION t

l 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified j

in-the OPERATING LIMITS REPORT.

l' APPLICABILITY: MODES 1* and'2**.

8LIIQH:

With a maximum of one shutdown rod inserted beyond the insertion limit, except l

for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 4

either:

i a.

Restore the rod to within the insertion limit specified in the 4'

OPERATING LIMITS REPORT, or i

l-b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

4 SURVEILLANCE REQUIREMENTS I

4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit:

)

a.

Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and b.

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

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With K,,, greater than or equal to.l.

BRAIDWOOD - UNITS 1 & 2 3/4 1-20 AMENDMENT NO. 80 y

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REACTIVIIY CONTROL SYSTEMS l

CONTROL R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the OPERATING LIMITS REPORT.

APPLICABILITY: MODES 1* and 2**.

i ACTION:

i With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:

a.

Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL F0WER which is allowed by the bank position using the insertion limits specified in the OPERATING LIMITS REPORT, or i

i c.

Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l f

SURVEILLANCE RE0VIREMENTS i

4.1.3.6 The position of each control bank shall be determined to be within i

the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Alarm is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4 See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With K,,, greater than or equal to 1.

BRAICWOOD - UNITS 1 & 2 3/4 1-21 AMENDMENT NO. 80

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FIGURE 3.1-1 (THIS FIGURE IS NOT USED) t l

BRAIDWOOD - UNITS 1 & 2 3/4 1-22 AMENDMENT NO. 80.

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3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band (flux difference units) about the target flux difference. The target band is specified in the OPERATING LIMITS REPORT.

The indicated AFD may deviate outside the required target band at greater than or equai to 50% but less than 90% of RATED THERMAL POWER provided the indicated AFD is within the Acceptable Operation Limits specifed in the OPERATING LIMITS REPORT and the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated AFD may deviate outside the required target band at greater than l

15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILITY: MODE 1 above 15% of RATED THERMAL POWER *.

ACTION:

a.

With the indicated AFD outside of the required target band and with l

THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes, either:

1.

Restore the indicated AFD to within the required target band l

limits, or 2.

Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

b.

With the indicated AFD outside of the required target band for more j

than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits specified in the OPERATING LIMITS REPORT and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce:

1.

THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and 2.

The Power Range Neutron Flux - High' Setpoints 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 .

See Special Test Exceptions Specification 3.10.2.

8 Surveillance testing of the Power Range Neutron Flux channel may

e performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits specified in the OPERATING LIMITS REPORT.

A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.

3RAIDWOOD - UNITS 1 & 2 3/4 2-1 AMENDMENT NO. 80 7

l

1 LIMITING CONDITION FOR OPERATION ACTION (Continued) c.

With the indicated AFD outside of the required target band for more l

~

i than I hour of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD l

is within the required target band.

l l

SURVEILLANCE REOUIREMENTS l

4.2.1.1 The indicated AFD shall be determined to be within its limits during l

POWER OPERATION above 15% of RATED THERMAL POWER by:

)

a.

Monitoring the indicated AFD for each OPERABLE excore channel:

t 1)

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, i

and

?.

l 2)

At least once per hour for the first.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status.

b.

Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least j

once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed l

to exist during the interval preceding each logging.

i_

4.2.1.2 The indicated AFD shall be considered outside of its target band when j

two or more OPERABLE excore channels are indicating the AFD to be outside the j

target band.

Penalty deviation outside of the required target band shall be l

accumulated on a time basis of:

a.

One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and b.

One-half minute penalty deviation for each I minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

4.2.1.3 The initial determination of target flux difference following a refueling outage shall be based on design predictions. Otherwise, the target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the predicted value at the end of the cycle life.

BRAIDWOOD - UNITS 1 & 2 3/4 2-2 AMENDMENT NO. 80,

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FIGURE 3.2-1 (THIS FIGURE IS NOT USED) 1 I

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T BRAIDWOOD - UNITS 1 & 2 3/4 2-3 AMENDMENT NO. 80

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POWER DISTRIBUTION LIMITS

.3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z) e 4

j LIMITING CONDITION FOR OPERATION l

j 3.2.2 F,(Z) shall be limited by the following relationships:

F,(Z) s [

] [K(Z)] for P > 0.5, and i

F,(Z) $ [

] [K(Z)] for P s 0.5.

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Where:

i THERMAL POWER j

P RATED THERMAL POWER

=

(

F," -

the F limit (s) at RATED THERMAL POWER (RTP) specifiedintheOPERATINGLIMITSREPORT, i

i and l

K(Z) is the function specified in the OPERATING LIMITS i

REPORT for a given core height location.

APPLICABILITY: MODE 1.

4 ACTION:

)

With F,(Z) exceeding its limit:

a.

Reduce THERMAL POWER at least 1% for each 1% F,(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at 5

least 1% for each 1% F,(Z) exceeds the limit; and b.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F,(Z) 1 is demonstrated through incore mapping to be within its limit.

BRAIDWOOD - UNITS 1 & 2 3/4 2-4 AMENDMENT NO. 80

2 4

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.a au x -.,

,a s.

u- - - -.

>s*-

s n

s.-

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' O 4

t 1

4 1

i FIGURE 3.2-2 (THIS FIGURE IS NOT USED) 4 h

5 i

E BRAIDWOOD - UNITS 1 & 2 3/4 2-5 AMENDMENT NO. 80

i e

l I

I 1

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR 4'

LINITING 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.

j a.

1)

RCS Total Flowrate 2 390,400 gpm, and 1

2)

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

Fl,sFj'[ [1.0 + PF, (1.0-P)]

3 1

l where:

i P-THERMAL POWER RATED THERMAL POWER l

Fj -

the F"h1 RATING LIMITS REPORT, and limit (s) at RATED THERMAL POWER (RTP) specified in the O j

PF,,-

the Power Factor Multiplier (s) for Fj,specified in the OPERATING LIMITS REPORT.

~

Measured values of F",iate uncertainty of 4% (nominal) or greater are obtained by using the movable incore detectors. An appropr j

shall then be applied to the measured value of F", before it is compared to the requiremnts.

l FtllCAEU.JII: MODE 1.

l ACTION:

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

)

j 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 F", 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 .

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

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

BRAIDWOOD - UNITS 1.& 2 3/4 2-8 AMENDMENT NO. 80

POWER DISTRIBUTION LIMITS 1

BASES l

AXIAL FLUX DIFFERENCE (Continued)

Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.

This deviation will-not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a sub:;equent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a j

1-hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits 3-1-

specified in the OPERATING LIMITS REPORT while at THERMAL POWER levels between l l

50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% and i

50% of RATED THERMAL POWER, deviations of the AFD outside of the target band i

are less significant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced d

significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines tha 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. and RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2).in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod position differing by more than 12 steps, indicated, from the group demand position, b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6, BRAIDWOOD - UNITS 1 & 2 B 3/4 2-2 AMENDMENT NO. 80

m.. _ _._ _ _. _ _ _. _. _ _ _ _ _ _ _ _ _ _ _ _ _ ~ _ _ _ _.. _. _.... _ _.. _

ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued)

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT

6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the facility.during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the 00CM and (2) Sections i

IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

l M FFLUENT RELEASE REPORl**

i 6.9.1.7 A Radioactive Effluent Release Report covering the operation of the facility during the previous year shall be submitted prior to May 1 of each

. year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility. The e

material provided shall be (1) consistent with the objectives outlined in the i

ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

MONTHLY OPERATING REPORT 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or RCS safety valves, shall be submitted on a monthly basis to the Director, Office of. Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report.

]

OPERATING LIMITS REPORT 6.9.1.9 Operating limits shall be established and documented in the OPERATING LIMITS REPORT (OLR) before each reload cycle or any remaining part of a reload cycle for the following:

1.

Moderator Temperature Coefficient for Specification 3.1.1.3, 2.

Shutdown Bank Insertion Limit for Specification 3.1.3.5, 3.

Control Bank Insertion Limit for Specification 3.1.3.6, 4.

Axial Flux Difference Limits, Target Band for Specification 3.2.1, 5.

Heat Flux Hot Channel Factor and K(Z) for Specification 3.2.2, 6.

Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for

)

Specification 3.2.3, and 7.

F Radial Peaking Factor for Specifi' cation 4.2.2.2.

y I

, A single submittal may be made for a multi-unit station.

A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

BRAIDWOOD - UNITS 1 & 2 6-22 AMENDMENT NO. 80

l M RATIVE CONTROLS

}

REPORTING REQUIREMENTS (Continued) i

{

OPERATING LIMITS REPORT (Continuedi r'

The analytical methods used to determine the operating limits shall be those i

previously reviewed and approved by the NRC in the following documents:

1.

WCAP-9272.P-A, " Westinghouse Reload Safety Evaluations Methodology,"

dated July 1985 (Westinghouse Proprietary).

(Methodology for Specification: Shutdown Bank Insertion Limit, Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear i

Enthalpy Rise Hot Channel Factor).

]

1

.2.

WCAP-8385, " Power Distribution Control and Load Following Procedures-Topical Report," dated September 1974 (Westinghouse Proprietary).

(Methodology for Specification: Axial Flux Difference, Constant Axial Offset Control).

3.

WCAP-9220-P-A, " Westinghouse ECCS Evaluation Model-1981 Version,"

Revision 1, dated February 1982 (Westinghouse Proprietary).

(Methodology l

for Specification: Heat Flux Hot Channel Factor).

I 4.

WCAP-9561-P-A, "BART A-1: A Computer Code for the Best Estimate Analysis of Reflood Transients," including Addendum 3, "Special Report - Thimble j

Modeling in Westinghouse ECCS Evaluation Model," Revision 1, dated July j

1986 (Westinghouse Proprietary).

(Methodology for Specification: Heat Flux Hot Channel Factor).

5.

WCAP-10266-P-A, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," Revision 2, dated March 1987, including Addendum 1 " Power Shape Sensitivity Studies," Revision 2-P-A, dated December 15, 1987, and Addendum 2 " BASH Methodology Improvements and Reliability Enhancements," Revision 2, dated May 1988 (Westinghouse Proprietary).

(Methodology for Specification: Heat Flux Hot Channel Factor).

i 6.

NFSR-0016, " Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods," dated July 1983.

(Methodology for Specification: Shutdown Bank Insertion Limit, Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Fr.ctor, and Nuclear Enthalpy Rise Hot Channel Factor).

7.

NFSR-0081, " Commonwealth Edison Company Topical Report on Benchmark of PWR Nuclear Design Methods Using the Phoenix-P and ANC Computer Codes,"

dated July 1990.

(Methodology for Specification: Shutdown Bank Insertion Limit, Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, Nuclear Enthalpy Rise Hot Channel Factor, and Moderator Temperature Coefficient).

8.

WCAP-10079-P-A, "NOTRUMP, A Nodal transient Small Break and General Network Code," dated August 1985 (Westinghouse Proprietary).

(Methodology for Specification: Heat Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor).

BRAIDWOOD - UNITS 1 & 2 6-22a AMENDMENT N0 80

-. -... -. ~.

i 4

ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued)

OPERATING LIMITS REPORT (Continued) 9.

WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," dated August 1985 (Westinghouse Proprietary).

(Methodology for Specification: Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor).

10. Comed letter from D. Saccomando to the Office of Nuclear Reactor Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-Il992/11993 and Comed application of the UET methodology addressed in " Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity Controls Systems."

The operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.

]

I The OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

4

)

BRAIDWOOD - UNITS 1 & 2 6-22 b AMENDMENT NO. 80

ML20137Y066

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