Text

Amends 88 to Licenses NPF-37 & NPF-66,respectively, Relocating Certain cycle-specific Parameter Limits from TSs to Operating Limits Rept
	ML20137X078
Person / Time
Site:	Byron
Issue date:	04/16/1997
From:	Dick G; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20137X084	List: ML20137X078; ML20137X116; ML20137Y066;
References
	NPF-37-A-088, NPF-66-A-088 NUDOCS 9704210157
	Download: ML20137X078 (23)
	v • d • e

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p ur UNITED STATES p

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NUCLEAR REGULATORY COMMISSION t

WASHINGTON, D.C. 20616 4001 1

COMMONWEALTH EDISON COMPANY DQCKET NO. STN 50-454 BYRON STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 88 License No. NPF-37 l

1.

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

A.

The application for amendment by Comonwealth 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 4

Act) and the Comission'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 Comission; C.

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

The issuance of this amendment will not be inimical to the comon 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 l

Part 51 of the Comission'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 read as follows:

9704210157 970416 PDR ADOCK 05000454 i

P PDR

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(2)

Technical Snecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 88 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 and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Ut i

Georg 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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$n ner oqk UNITED STATES p

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NUCLEAR REGULATORY COMMISSION i

2 WASHINGTON, D.C. 2008dHm01 i

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i COM)NWEALTH EDISON COMPANY f

DOCKET NO. STN 50_(51 l

i

. BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE i

i Amendment No. 88 i

License No. NPF-66 1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated 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' Comnission's rules and regulations set forth in 10 CFR Chapter 1; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations,as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows:

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(2)

Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113),

as revised through Amendment No. E8 and revised by Attachment 2 to NPF-66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, dated February 14, 1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental' Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION i

.Uf George F. Dick,i,enior Project Manager r

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 i

1!

i l

ATTACHMENT TO LICENSE AMENDMENT N05. 88 AND 88 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPf.-11 DOCKET NOS. STN 50-454 AND STN 50-455 Revise the Appendix A Technica1' Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

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 j

3/4 2-1 3/4 2-1 i

3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-8 3/4 2-8 8 3/4 2-2 B 3/4 2-2 i

6-22 6-22 6-22a 6-22a 6-22b j

i 4

l

a f

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0UIREMENTS i

3 l

SECTION Pf_GE b

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

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

3/4.1.1-BORATION CONTROL l

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 Minimum Temperature for Cri ticality......................

3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Fl ow Pa t h -' Shu tdown.....................................

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 Boratea Water Source - Shutdown..........................

3/4 1-11 i

Borated Water Sources - Operating........................

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

3/41-13a 3/4.1.3 M0VABLE CONTROL ASSEMBLIES Group' Height.............................................

3/4 1-14

)

TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT'0F 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 Insertinn Limits.............................

3/4 1-21 FIGURE 3.1-1 (THIS FIGURE IS NOT USED)..........................

3/4 1-22 BYRON - UNITS 1 & 2 IV AMENDMENT N0. 88

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION fiqE 3/4.2 POWER DISTRIBUTION LIMITS j

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

3/4 2-1 FIGURE 3.2-1 (THIS FIGURE IS NOT USED)............................

3/4 2-3 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 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT0R.................................................

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

3/42-10 3/4.2.5 DNB PARAMETERS...........................................

3/42-13 TABLE 3.2-1 DNB PARAMETERS........................................

3/42-14 3/4.3 INSTRUMENTATION j

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

3/4 3-1 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 REQUIREMENTS........................................

3/4 3-9

)

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................................

3/43-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATI0W.....................................

3/43-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i

INSTRUMENTATION TRIP SETP0!NTS......................

3/43-23 TABLE 3.3-5 (THIS TABLE IS NOT USED)..............................

3/43-30 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUhENTATION SURVEILLANCE REQUIREMENTS...........

3/43-34 l

BYRON

. UNITS 1 & 2 V

AMENDMENT NO. 88

DEFINITIONS

\\

4 0FFSITE DOSE CALCULATION MANUAL J

1.18 The OFFSITE DOSE CALCULATION MANUAL (0DCM) 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 1

effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-

{

mental Radiological Monitoring Program. The ODCM shall also contain (1) the 1

Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Sections 6.8.4e and f, and (2) descriptions of the information that should be included in the Annual Radiological Enviror.;nental Operating and Radioactive Effluent Release Reports required by Specifications j

.6.9.1.6 and 6.9.1.7.

OPERABLE - OPERABILITY

\\

1.19 A system, subsystem, train, component or device shall be OPERABLE or j.

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

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

i l

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

OPERATING LIMITS REPORT 1.19.a The OPERATING LIMITS REPORT (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 limits is addressed in individual specifications.

l QEERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor i

coolant. temperature specified in Table 1.2.

E.

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

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

l PRESSURE BOUNDARY LEAKAGE l

1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component j

body,. pipe wall, or vessel wall.

I BYRON - UNITS 1 & 2 1-4 AMENDMENT NO. 88 f

1

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REACTIVITY CONTROL SYSTEMS 3/4.I.3 MOVABLE (DM ROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION I

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 i

demand position.

APPLICABILITY: MODES 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, detemine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within I hour and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one full-length rod trippable but inoperable due to causes other i

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:

4 1.

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

2.

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

The i

THERMAL POWER level shall be restricted pursuant to l

Specification 3.1.3.6 during subsequent operation, or 3.

The rod is declared inoperable and the SHUTDOWN HARGIN 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 j

is reduced to less than or equal to 85% of RATED THERMAL POWER.

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 ; 5.ee Special Test Exceptions Specifications 3.10.2 and 3.10.3.

BYRCN - UNITS 1 & 2 3/4 1-14 AMENDMENT NO. 88 I

l l

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) 4 c)

A power distribution map is obtained from the movable l

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) t d)

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

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 i

its group step counter demand height by more than i 12 steps (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 12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Specification 3.1'.3.6.

The THERMAL POWER level shall l

be restricted pursuant to Specification 3.1.3.6 during i

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 .

SURVEILLANCE REQUIREMENTS 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 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 least once per 92 days.

l BYRON - UNITS 1 & 2 3/4 1-15 AMENDMENT N0.88 l

l

REACTIVITY CONTROL SYSTEMS SHOTDOWN ROD INSERTION LIMI LIMITING C0ffDITION FOR OFERATION 3.1.3,5 All shutdown rods shall be limited in physical insertion as specified in the OPERATING LIMITS REPORT.

APPLICABILITY: MODES 1* and 2*f.

8011Q!i; 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 I hour

)

either:

a.

Restore the rod to within the insertion limit specified in the l

OPERATING LIMITS REPORT, or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE RE0VIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion l

limit:

a.

Within 15 ' minutes prior to withdrawal of any rods in Control 1

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 1.

BYRON - UNITS 1 & 2 3/4 1-20 AMENDMENT NO. 88

_R_EACTIVITY CONTROL SYSTEMS CONTROL ROD 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*#.

ACTION:

l With the control banks inserted beyond the insertion limits, except for l

surveillance testing p rsuant 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 POWER which is allowed by the bank position using the insertion limits specified in the OPERATING LIMITS REPORT, l

or c.

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

}URVEILLANCEREQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within 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 Roo 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 .

See Special Test Exceptions Specifica.tions 3.10.2 and 3.10.3.

With K,, greater than or equal to 1.

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i BYRON - UNITS 1 & 2 3/4 1-21 AMENDMENT NO. 88

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FIGURE 3.1-1 i

(THIS FIGURE IS NOT USED)

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BYRON - UNITS 1 & 2 3/4 1-22 AMENDMENT NO. 88

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3/4.2 POWER OlSTRIBUTION 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.

l The indicated AFD may deviate outside the required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indicated l

AFD is within the Acceptable Operation Limits specified 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 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 .

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

ACTION:

a.

With the indicated AFD outside of the required target band and with THERriAL 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 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 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 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.

Surveillance testing of the Power Range Neutron Flux channel may be performed l

pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits specified in the OPERATING LIMITS l

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 required target band during testing without penalty deviation.

I BYRON - UNITS 1 & 2 3/4 2-1 AMENDMENT N0. 88

LIMITING' CONDITION FOR OPERATION ACTION (Continued)

[

c.

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

4 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 1

is within the required target band.

l SURVEILLANCE REQUIREMENTS i

4.2.1.1 The indicated AFD shall be datermined to be within its limits during POWER OPERATION above 15% of RATED THEIG14L POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel:

1)

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, 4

and i

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.

s i

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 once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the 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 I minute of POWER OPERATION 4

outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and 4

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 re-fueling 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 betwe'en the most recently measured value and the predicted value at the end of the cycle life.

BYRON - UNITS 1 & 2 3/4 2-2 AMENDMENT NO. 88

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l Figure 3.2-1 (THIS FIGURE IS NOT USED) i BYRON - UNITS 1 & 2 3/4 2-3 AMENDMENT NO. 88 l

2 l

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POWER DISTRIBUTION LIMITS 3/4'2.2 HEAT FLUX HOT CHANNEL FACTOR - F f7)

LIMITING CONDITION FOR OPERATION l

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

F,(Z) s [

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

F,(Z) s [

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

i i.

Where:

THERMAL POWER RATED THERMAL POWER P

j

=

l F," -

the F.ied in the OPERATING LIMITS REPORT, limit (s) at RAT l

j specif and i

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

REPORT for a given core height location.

APPLICABILITY: MODE 1.

ACTION:

l With F,(Z) exceeding its limit:

I a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit within 15 minutes and similarly reduce the Pow,er 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 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 demonstrated through incore mapping to be within its limik(.Z) is a.,

above; THERMAL POWER may then be increased provided F l

j BYRON - UNITS 1 & 2 3/4 2-4 AMENDMENT NO. 88

1 4

1 d

i a

4 l

4 FIGURE 3.2-2 (THIS FIGURE IS NOT USED) l l

)

l BYRON - UNITS I & 2 3/4 2-5 AMENDMENT NO. 88 1

)

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

a.

1)* RCS Total Flowrate 1371,400 gpm, 2)** RCS Total Flowrate 2390,400 gps, and b.

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

3 where:

2 P

THERMAL POWER RATED THERMAL POWER Fl'[ =

the F"1 RATING LIMITS REPORT, and limit (s) at RATED THERMAL POWER (RTP) specified in theOh PF,, -

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

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

then be applied to the measured value of F", before it is compared to 1

the requirements.

j APPLICABILITY: MODE 1.

ACTION-With RCS total flow rate or F", outside the region of acceptable operation:

i a.

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

l.

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

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

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Applicable to Unit 1.

Applicable to Unit 2 after cycle 5.

, Not applicable to Unit 1.

Applicable to Unit 2 until the completion of cycle 5.

BYRON - UNITS 1 & 2 3/4 2-8 AMENDMENT NO. 88

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POWER DISTRIBUTION LIMITS l

BASES 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 subsequent return to RATED THERMAL POWER (with the AFD within the target j

band) provided the time duration of the deviation is limited. Accordingly, a 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 3

provided for operation outside of the target band but within the limits i

- specified in the OPERATING LIMITS REPORT while at THERMAL POWER levels between l 50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% and 4

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

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 i

reduced significance.

1 Provisions for monitoring the AFD on an automatic basis are derived from j

the plant process computer through the AFD Monitor Alarm.- The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and

.i provides an alarm message immediately if the AFD for two or more OPERABLE L

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 J

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 l

limits of I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

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

l 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. and RCS FLOWRATE AND l

NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR l

The limits on heat flux bet channel factor, RCS flowrate, and nuclear l:

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 i

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, BYRON - UNITS 1 & 2 B 3/4 2-2 Amendment No. 88

l ADMINISTRATIVE CONTROLS-i 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-i tions, and analysis:of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be l

~ consistent with the objectives outlined in (1) the ODCM and (2) Sections l

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

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L ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT **

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6.9.1.7.A Radioactive Effluent Release Report covering the operation of the l

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

-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.

L OPERATING LIMITS REPORT 6.9.1.9 Operating limits shall be established and documented in the OPERATING

- LIMITS REPORT (0LR) 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 Hat Channel Factor, and Power Factor Multiplier for Specification 3.2.3, and 7.

F Radial Peaking Factor for Specification 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 j

combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the i

releases of radioactive material from each unit.

l BYRON - UNITS 1 & 2 6-22 AMENDMENT NO. 88

ADMINISTRATIVE CONTROLS REPORTING REOUIREMENTS (Continued)'

OPERATING LIMITS REPORT (Continued)

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

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 Specificaticn: Shutdown Bank Insertion Limit,: Control Bank Insertion Limit, Axial Flux Difference, Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel: Factor).

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 for Specification: Heat Flux Hot Channel Factor).

)

4..

WCAP-9561-P-A, "BART A-1: A Computer Code for the Best_ Estimate Analysis of Reflood Transients," includk.;; Addendum 3, "Special Report - Thimble Modeling in Westinghouse ECCS Evaluation Model," Revision 1, dated July 1986 (Westinghouse Proprietary).

(Methodology for Specification: Heat Flux Hot Channe) 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 i

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).

6.

NFSR-0016, " Commonwealth Edison Company Topical Report on Benchmark of l

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 Factor, 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 t

Insertion Limit, Control Bank Insertion Limit, Axial Flux Difference, Heat F. lux 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).

BYRON - UNITS 1 & 2 6-22a AMENDMENT N0 88

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

ADMINISTRATIVE' CONTROLS 3

REPORTING' REQUIREMENTS (Continued)

OPERATING LIMITS REPORT (Continued)

I

. 9.

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

(Methodolcgy 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 j

Regulation dated December 21, 1994, transmitting an attachment that documents applicable sections of WCAP-11992/11993 and Comed application j

i; '

of.the UET methodology addressed in " Additional Information Regarding Application for Amendment to Facility Operating Licenses-Reactivity J

4 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 l

Inspector.

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BYRON - UNITS 1 & 2 6-22b AMENDMENT N0. 88

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ML20137X078

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