Text

Amend 144 to License NPF-3,relocating Values of cycle-specific Limits from Tech Specs to Core Operating Limits Rept Per Generic Ltr 88-16
	ML20005G042
Person / Time
Site:	Davis Besse
Issue date:	01/11/1990
From:	Hannon J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20005G043	List: ML20005G042; ML20005G045;
References
	GL-88-16, NUDOCS 9001180034
	Download: ML20005G042 (26)
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L-UNITED STATES

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NUCLEAR REGULATORY COMMISSION 7,

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l TOLEDO.ED150N COMPANY p

$Ng THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DuCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. 1 AMENDMENI TO FACILITY OPERATING LICENSE L

Amendment No. 144 License No. NPT-3

1. _

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

A.

The application for amendment by the Toledo Edison Company and The Cleveland Electric Illuminating Company (the licensees) dated June 16, 1989 as revised August 21, 1989, complies with the standards and recuirements of the Atomic Energy Act of 1954, as amended (the Act),

anc 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.

Tnere 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 conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense ano 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 Comission's regulations and all applicable requirements have been satisfied.

2.

- Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:

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(a) Technical Specifications L

The Technical Specifications contained in Appendix A, as revised i

through Amendment No.144, are hereby incorporated in the license.

The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

3.

' This license amendment is effective as of its date of issuance and shall be implemented not later than 45 days after issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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1 John N. Hannon, Director Project Directorate III-3 Division of Reactor Projects - III, IV, J

OffIc&SpecialProjects V

e of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications

-Date of Issuance:

January 11, 1990 s -e4g-Fc'-e w-

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vne ATTACHMENT TO LICENSE AMENDMENT NO. 144 FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 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. The corresponding overleaf pages are also provided to maintain document completeness.

Remove Insert L

la la 1-6c 3/4 1-20 3/4 1-?0 3/4 1-26 3/4 1-26 3/4 1-27 3/4 1-27 3/C 1-28 through 3/4 1-?8d 3/4 1-29 through 3/4 1-29d 3/4 1-30 3/4 1-30 3/4 1-31 3/4 1-32

'3/4 1-33 3/4 1-33 3/4 1-34 3/4 1-34 3/4 1-35 through 3/4 1-43 3/4 2-1 3/4 2-1 3/4 2-2 through 3/4 2 4a 3/4 2-9 3/4 2-9 3/4 2-10 3/4 1-10 3/4 2-11 3/4 2-11 3/4 2-12 3/4 8-4 3/48-4(nochange.)

B 3/4 2-1 B 3/4 2-1 6-16 6-16 6-17 6-17 s

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INDEX DEFINITIONS SECTION'

.PAGE 1.0 DEFINITIONS DEFINED TERMS................................................

1-1 THERMAL P0WER................................................ 1 RATED THERMAL P0WER..........................................

1-1 0

OPERATIONAL M0DE.............................................

1-1 ACTI0N.......................................................

1-1 OPERABLE - OPERABILITY.......................................

1-1 REPORTABLE EVENT.............................................

1-2 CONTAINMENT INTEGRITY........................................

1-2 CHANNEL CALIBRATION..........................................

1-2 CHANNEL CHECK................................................

1-2 CHANNEL FUNCTIONAL TEST......................................

1-3 CORE ALTERATION..............................................

1-3 SHUTDOWN MARGIN..............................................

1-3

' IDENTIFIED LEAKAGE...........................................

1-3 UNIDENTIFIED LEAKAGE.........................................

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

1-4 CONTROLLED LEAKAGE...........................................

1-4 QUADRANT POWER TILT..........................................

1-4 i.

DOS E EQU I V AL ENT I-131........................................

1-4 Y-AVERAGE DISINTEGRATION ENERGY..............................

1-4 STAGGERED TEST BASIS.........................................

1-5 FREQUENCY N0TATION...........................................

1-5 AXIAL P0WER' IMBALANCE........................................

1-5 SHI EL D BUILDING INTEGRITY.................................... 1-5

. REACTOR PP.0TECTION SYSTEM RESPONSE TIME......................

1-5 SAFETY FEATURE RESPONSE TIME.................................

1-6 PHYSICS TESTS................................................

1-6 STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM RESPONSE TIME.....

1-6 SOURCE CHECK.................................................

1-6a PROCESS CONTROL PR0 GRAM......................................

1-6a l

DAVIS-BESSE, UNIT 1 I

Amendment No. 38, S6, 93

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t-INDEX j

DEFINITIONS SECTION PAGE 1.0' DEFINITIONS (Continued)

SOLIDIFICATION...............................................

1-6a OFFSITE DOSE CALCULATION MANUAL (0DCM).......................

1-6a GASEOUS RADWASTE TREATMENT SYSTEM............................

1-6a VENTILATION EXHAUST TREATMENT SYSTEM.........................

1-6a p

PURGE-PURGING................................................

1-6b VENTING......................................................

1-6b MEMBER (S ) 0F THE PUBL I C......................................

1-6b S I T E B 0VN DARY................................................

1 - 6 b UN RE ST R I CTE D ARE A............................................

1 - 6b DEWATERING...................................................

1-6b CORE OPERATING L IMITS REP 0RT.................................

1-6c OPE RATIONAL MODES (TABL E 1.1 )................................

1 -7 FREQUENCY NOTATION (TABL E 1. 2)...............................

1-8 i

I DAVIS-BESSE, UNIT 1 la Amendment No. P6, 144 l

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DEFINITIONS CORE OPERATING LIMITS REPORT 1.41 The CORE OPERATIt:G LIMITS REPORT.is the unit-specific document that provides core operating limits for the current reload cycle. These cycle-specific core operating limitt shall be determined for each reload cycle in 4

accordance with Specification 6.9.1.7.

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

DAVIS-BESSE, UNIT 1 1-6c Amendment No.144

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3/4.1.3 MOVABLE CONTROL ASSEMBLIES

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GROUP HEIGHT - SAFETY AND REGULATING ROD CROUPS i

LIMITING CONDITION FOR OPERATIONS J

3.1. 3.1 All control (safety and regulating) rods shall be OPERABLE and positioned within + 6.5% (indicated position) of their group average height.

APPLICABILITY: MODES 1* and 2*,

ACTION:

a.

With one or more control 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 requirement of Specification 3.1.1.1 is satisfied within one hour and be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With more than one control rod inoperable or misdligr.ed from

.o its group average height by more than + 6.5% (indicated I

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

c.

With one control rod inoperable due to causes other than addressed by ACTION a, above, or misaligned from its group average height by more than + 6.5% (indicated position), POWER OPERATION may continue proviEed that within one hour either:

1.

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

The control 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)

An analysis of the potential ejected rod worth is performed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the rod worth is deter-mined to be < 1.0% ak at zero power and < 0.65%

ak at RATED THERMAL POWER for the remainFer of the fuel cycle, 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 .

5ee Special Test Exceptions 3.10.1 and 3.10.2.

m DAVIS-BESSE, UNIT 1 3/4 1-19

REACTIVITY CONTROL SYSTEMS i

GROUP HEIGHT - SAFETY AND REGULATING ROD GROUPS LIMITING CONDITION FOR OPERATIONS t

ACTION:

(Continued)

_h,[e$ o# f a d yppis btained from the incore r

are verified to be within their limitswithin7hhoursgH d)

Either the THERMAL POWER level is reduced to < 60% of the THERMAL POWER allowable for the reactor cliolant pump combination within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months I

the High Flux Trip Setpoint is reduced to 1 70% of the THERMAL POWER allowable for the reactor coolant pump combination, or e)

The remainder of the rods in the group with the inoperable rod are aligned to within + 6.5% of the inoperable rod within one hour while maintaining the position of the rods within the limits provided in the CORE OPERATING LIMITS REPORT; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

1 SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each control rod shall be detennined to be within the group average height 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 Asyrmetric Rod Fault Circuitry is inoperable, then verify the individual rod position (s) of the rod (s), with inoperable Fault Circuitry 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 control rod not fully inserted shall be determined to be OPERABLE by movement of at least 2% in any one direction at least once every 31 days.

DAVIS-BESSE, UNIT 1 3/4 1-20 Amendment No. J)5,144 I

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REACTIVITY CONTROL SYSTEMS l

SAFETY-R0D INSERTION LIMIT i

, LIMITING CONDITION FOR OPERATION i.

3.1.3.5 All safety rods'shall be fully withdrawn.

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APPLICABILITY:

1* and 2*#,

l ACTION:

l With a maximum of one safety rod not fully withdrawn, except for sur-veillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

4 1

a.

Fully withdraw the rod or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

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L SURVEILLANCE REQUIREMENTS j

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4.1.3.5 Each safety rod shall be determined to be fully withdrawn:

a.

Within 15 minutes prior to withdrawal of any regulating rod during an approach to reactor criticality, 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 Exception 3.10.1 and 3.10.2.

With K,ff 1 1.0, l

i DAVIS-BESSE, UNIT 1 3/4 1-25 1

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REACTIVITY CONTROL SYSTEMS

_ REGULATING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION-3.1.3.6 The regulating rod groups shall be positioned within the acceptable operating limits for regulating rod position provided in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: MODES 1* and 2*#.

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ACTION:

With the regulating rod groups inserted beyond the operating limits (in a region other than acceptable operation), or with any group sequence or overlap outside the limits provided ir the CORE OPERATING LIMITS REPORT except for l

surveillance testing pursuant to Specification 4.1.3.1.2 either:

a.

Restore the regulating groups to within the limits provided in the CORE OPERATING LIMITS REPORT within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or l

b.

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position limits provided in the CORE OPERATING LIMITS REPORT within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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 .

NOTE:

If in unacceptable region, also see Section 3/4.1.1.1.

See Special Test Exception 3.10.1 and 3.10.2.

With K,ff 1 1.0.

DAVIS-BESSE UNIT 1 3/4 1-26 Amendment No. JJ,33,$J.AZ,M,$J.

H,80,123.144

REACTIVITY CONTROL SYSTEMS c

REGULATING ROD INSERTION LIMITS SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating group shall be determined to be within the limits provided in the CORE OPERATING LIMITS REPORT at least once j

every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except when:

I a.

The regulating rod insertion limit alarm is inoperable, then verify the groups to be within the insertion limits at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; b.

The control rod drive sequence alarm is inoperable, then verify.the groups to be within the sequence and overlap limits at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

DAVIS-BESSE, UNIT 1 3/4 1-27 Amendment No.144 (next page is 3/4 1-30) l

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REACTIVITY CONTROL SYSTEMS ROD PROGRAM LIMITING CONDITION FOR OPERATION I

I 3.1.3.7 Each control rod assembly (safety, regulating and APSR) shall be l

programed to operate in the core location and rod group specified in the CORE OPERATING LIMITS REPORT.

t APPLICABILITY: MODES 1* and 2*.

ACTION:

With any control rod assembly not programed to operate as specified above, l

be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

SURVEILLANCE REQUIREMENTS 4.1.3.7 a.

Each control rod assembly shall be demonstrated to be program ed to operate in the specified core location and rod group by:

1.

Selection and actuation from the control room and verification of movement of the proper rod as indicated by both the absolute and relative position indicators:

a)

For all control rod assemblies after the control rod I

drive patches are locked subsequent to test, reprograming

.or maintenance within the panels, b)

For specifically affected individual rod assemblies i

following maintenance, test, reconnection or modification of power or instrumentation cables from the control rod drive control system to the control rod drive.

2.

Verifying that each cable that has been disconnected has been properly matched and reconnected to the specified control rod drive.

b.

At least once each 7 days, verify that the control rod drive patch panels are locked.

See Special Test Exceptions 3.10.1 and 3.10.2.

DAVIS-BESSE, UNIT 1 3/4 1-30 Amendment No. JJ,144 (Nextpageis3/41-33)

p REACTIVITY CONTROL' SYSTEMS XENON REACTIVITY LIMITING CONDITION FOR OPERATION 3.1.3.8 THERMAL POWER shall not be increased above the power level cutoff specified in the acceptable operating limits for regulating rod position provided in the CORE OPERATING LIMITS REPORT unless one of the following conditions is satisfied:

a.

Xenon reactivity is within 10 percent of the equilibrium value for RATED THERMAL POWER and is approaching stability, or b.

THERMAL POWER has been within a range of 87 to 92 percent of RATED THERMAL POWER for a period exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in the soluble poison control mode, excluding xenon free start-ups.

APPLICABILITY: MODE 1.

ACTION:

With the requirements of the above specification not satisfied, reduce THERMAL POWER to less than or equal to the power level cutoff within 15 minutes.

SURVEILLANCE REQUIREMENTS 4.1.3.8 Xenon reactivity shall be determined to be within 10% of the equilibrium value for RATED THERMAL POWER and to be approaching stability or it shall be determined that the THERMAL POWER has been in the range of 87 to 92% of RATED THERMAL POWER for > 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , prior to increasing THERMAL

' POWER above the power level cutoff.

DAVIS-BESSE UNIT 1 3/4 1-33 Amendment No.144

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REACTIVITY CONTROL SYSTEMS i

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AXIAL POWER SHAPING ROD INSERTION LIMITS l

.IMITING CONDITION FOR OPERATION

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3.1. 3. 9 The axial power shaping rod group shall be within the acceptable operating limits for axial power shaping rod position specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: MODES 1 ad 2*.

ACTION:

i With the axial power shaping rod group outside the above insertion limits, either:

a.

Restore the axial power shaping rod group 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 to less than or equal to that fraction of RATED THERMAL. POWER which is allowed by the rod group position using the acceptable operating limits provided in the CORE OPERATING LIMITS REPORT within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 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 .

SURVEILLANCE REQUIREMENTS 4.1.3.9 The position of the axial power shaping rod group shall be determined to be within the limits provided in the CORE OPERATING LIMITS REPORT at least l

once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except when the axial power shaping rod insertion limit alarm is inoperable, then verify the group to be within the limit provided in the CORE OPERATING LIMITS REPORT at least once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

W'ith k,ff > 1.0.

DAVIS-BESSE, UNIT 1 3/4 1-34 Amendment No. 33. A2.4{ 467,67, (Next page is 3/4 2-1) 89 J13,

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LIMITING' CONDITION FOR OPERATION 3.2.1 l AXIAL, POWER IMBALANCE shall be maintained within the acceptable AXIAL

. POWER IMBALANCE operating limits provided in the CORE OPERATING LIMITS REPORT.

APPLICABILITYi~ MODE 1 above 40% of RATED THERMAL POWER.*

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ACTION!

Wii.h 'AXIh NWER IMBALANCE exceeding the limits specified above, either:

i a.

Restore the AXIAL POWER IMBALANCE to with the limits provided in the CORE OPERATING LIMITS REPORT within 15 minutes. or b.

Within one hour' reduce power until imbalance limits provided in-the CORE OPERATING LIMITS REPORT are met or to 40% of RATED THERMAL POWER or less.

' SURVEILLANCE REQUIREMENTS 4.2.1 The' AXIAL POWER IMBALANCE shall be determined to be within the limits t

.provided'in the CORE OPERATING LIMITS REPORT at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when above 40% of RATED THERMAL POWER except when.the AXIAL POWER IMBALANCE i

' alarm is inoperable, then calculate the AXIAL POWER IMBALANCE at least once per hour.-

See Special. Test Exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 2-1 Amendment No. 33,A2,46,67,69, (Next page is 3/4 2-5) 80,J23,144

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QUADRANT POWER TILT g

b LIMITING CONDITION FOR OPERATION 3.2.4 THE QUADRANT POWER TILT shall not exceed the Steady State Limit for QUADRANT POWER TILT provided in the CORE OPERATING LIMITS REPORT.

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

ACTION:

a.

With the QUADRANT POWER TILT determined to exceed the Steady State Limit but less than or equal to the Transient Limit provided in the CORE OPERATING LIMITS REPORT:

l.

Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

a)

Either reduce the QUADRANT POWER TILT to within its Steady State Limit, or b)

Reduce THERMAL POWER so as not to exceed THERMAL POWER, including power level cutoff, allowable for the reactor coolant pump combination less at least 2% for each 1% of QUADRANT POWER TILT in excess of the Steady State Limit and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , reduce the High Flux Trip Setpoint and the Flux-a Flux-Flow Trip Setpoint at least 2% for each 1% of QUADRANT POWER TILT in excess of the Steady State Limit.

2.

Verify that the QUADRANT POWER TILT is within its Steady State Limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the Steady State Limit or reduce THERMAL POWER to less than 60% of THERMAL POWER allowable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to < 65.5%

of THERMAL POWER allowable for the reactor coolant pump-combination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 60% of THERMAL POWER allowable for the reactor coolant pump combination may proceed provided that the QUADRANT POWER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% or greater RATED THERMAL POWER.

See Special Test Exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 2-9 Amendment No. J23,144

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POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued) b.

With the QUADRANT POWER TILT determined to exceed the Transient Limit but less than the Maximum Limit provided in the CORE OPERATING LIMITS REPORT, due to misalignment of either a safety, regulating or axial power shaping rod:

1.

Reduce THERMAL POWER at least 2% for each 1% of indicated QUADRANT POWER TILT in excess of the Steady State Limit within 30 minutes.

2.

Verify that the QUADRANT POWER TILT is within its Transient Limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the Transient Limit or reduce THERMAL POWER to less than 60% of THERMAL POWER allowable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to 1 65.5% of THERMAL POWER allowable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 60%

of THERMAL POWER allowable for the reactor coolant pump combina4 1

tion may proceed provided that the QUADRANT POWER TILT is verified within its Steady State Limit at least once per hour for 12 ho.c; or until verified acceptable at 95% or greater RATED THERMAL POWER.

c.

With the QUADRANT POWER TILT determined to exceed the Transient Limit but less than the Maximum Limit provided in the CORE OPERATING LIMITS REPORT, due to causes other than the misalignment of either a safety, regulating or axial power shaping rod:

1.

Reduce THERMAL POWER to less than 60% of THERMAL POWER allowable for the reactor coolant pump combination within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the High Flux Trip Setpoint to 1 65.5% of THERMAL POWER allowable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 60%

of THERMAL POWER allowable for the reactor coolant pump combina-tion may proceed provided that the QUADRANT POWER TILT is verified within its Steady State Limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.

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DAVIS-BESSE, UNIT 1 3/4 2-10 Amendment No. J23,J35,144 l

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POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) r ACTION:

(Continued) d.

With the QUADRANT POWER TILT detennined to exceed the Maximum Limit provided in the CORE OPERATING LIMITS REPORT. reduce THERMAL POWER to 1 15% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

SURVEILLANCE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT shall be determined to be i the Steady State Limits provided in the CORE OPERATING LIMITS REPORT at least once every 7 days during operation above 15% of RATED THERMAL POWER except when the QUADRANT POWER TILT alarm is inoperable, then the QUADRANT POWER TILT shall be calculated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

DAVIS-BESSE, UNIT 1 3/4 2-11 Amendment No. J23,144 (Next page is 3/4 2-13)

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,s X

4 l,

3 ELECTRICAL p0WER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) r 1.

Verifying the fuel level in the day fuel tank.

2.

Verifying the fuel level in the fuel storage tank.

3.

Verifying the fuel transfer pump can be started and trans-fers fuel from the storage system to the day tank.

4.

Verifying the diesel starts and accelerates up to 900 rpm, preceded by an engine prelube and/or appropriate other warmup procedures.

5.

Verifying the generator is synchronized, loaded to 1 1000 kw, and operates for 1 60 minutes.

6.

Verifying the diesel generator is aligned to provide standby power to the associated essential busses.

7.

Verifying that the automatic load sequence timer is OPERABLE with each load sequence time within i 10% of its required value.

b.

At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank is within the acceptable limits specified in Table 1 of ASTM D975-68 when checked for viscosity, water and sediment.

c.

At least once per 184 days on a STAGGERED TEST BASIS by:

1.

Verifying the fuel level in the day fuel tank.

2.

Verifying the fuel level in the fuel storage tank.

3.

Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the day tank.

4.

Verifying the diesel starts from ambient condition and accelerates to at least 900 rpm in 1 10 seconds.

5.

Verifying the generator is synchronized, loaded to t 1000 kw, and operates for 1 60 minutes.

6.

Verifying the diesel generator is aligned to provide standby power to the associated essential busses.

7.

Verifying that the automatic load sequence timer is OPERABLE with each load sequence time within i 10% of its required value.

DAVIS-BESSE, UNIT 1 3/4 8-3 Amendment No. 75,97,105

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ELECTRICAL' POWER SYSTEMS

~

SURVEILLANCE' REQUIREMENTS- (Continued) d.-

At least once per 18 months during' shutdown by:

1.

Verifying the generator ' capability to reject a load. equal

- to the' largest single emergency load supplied by the generator without tripping..

2.

Simulating a loss of offsite power in conjunction with a i

safety features actuation system (SFAS) test signal, and:

.- ( a ) Verifying de-energization of the essential busses and load' shedding from the essential-busses.

(b) Verifying the diesel starts from ambient condition on the auto-start signal, energizes the-essential busses

-with permanently connected loads, energizes the auto-

. connected essential loads through the load sequencer and operates for > 5 minutes while its generator is loaded with the essential ~ loads.

(c). Verifying that all diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the essential bus and/or an SFAS test signal.

~ 3. -

Verifying the diesel generator operates for > 60 minutes i

while loaded to > 2000 kw.

3 i

'4.

Verifying that the auto-connected loads to each diesel i

generator do not exceed the 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months rating of 2838 kw.

e.

At least once per 30 months by subjecting the diesels to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby i

. service.*

i i

The provisions of Specification 4.0.2 are not applicable.

DAVIS-BESSE, UNIT 1 3/4 8-4 Amendment No. 97,J95,141 l

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3/4.2-POWER DISTRIBUTION LIMITS BASES The.specificationsofthissectionprovideassuranceoffuelintegrity)during Condition I-(normal operation) and II (incidents of moderate frequency events by:- (a) maintaining the minimum DNRB in the core > 1.30 during normal operation and during short term transients.-(b) maintaining the peak linear power density 118.4 kW/ft during nonnalioperation, and (c) maintaining the peak power density

.less than the limits.given in the bases to specification 2.1 during short term transients.

In addition, the above criteria must be met in order to meet the assumptions used for the loss-of-coolant accidents,

.The power imbalance envelope and the insertion limit curves defined in the CORE' o

OPERATING LIMITS REPORT are based on LOCA analyses which have defined the maximum linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance Criteria of 2200'F following a LOCA. Operation outside of the power imbalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur.

The power. imbalance envelope represents the boundary of operation limited.by the Final Acceptance Criteria only if the control rods are at the insertion limits, as defined in the CORE OPERATING LIMITS REPORTiand if the steady-state limit QUADRANT POWER TILT exists.

Additional conservatism is introduced by application of:

a.

Nuclear uncertainty factors.

b.

Thermal-calibration un;ertainty.

c.

Fuel densification effects.

d.-

Hot rod manufacturing tolerance factors.

e.

1 Potential fuel rod bow effects.

The ACTION statements which permit limited variations from the basic require-ments are accompanied by additional restrictions which ensures that the original criteria are met.

The definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:

F Nuclear heat flux hot channel factor, is defined as the maximum local q fuel rod linear power density divided by the average fuel rod linear power density, cssuming nominal fuel pellet and rod dimensions.

3 DAVIS-BESSE, UNIT 1 B 3/4 2-1 Amendment No. JJ,33,M,144

4 POWER DISTRIBUTION LIMITS q

BASES N

F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio AH of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power.

It has been determined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking and on minimum DNBR at full power are met, provided:

g i.93; F3g i 1.71 2

F Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking.

It has been determined that the above hot channel factor lim-its will be met provided the following conditions are maintained.

1.

Control rods in a single group move together with no individual rod in-sertion differing by more than +6.5% (indicated position) from the group average height.

?

Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6.

3.

The regulating rod insertion limits of Specification 3.1.3.6 are main-tained.

4.

AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE is a measure of the difference in power between the top and bottom halves of the core.

Calculations of core average axial peaking factors for many plants and measurements from operating plants under a variety of operat-ing conditions have been correlated with AXIAL POWER IMBALANCE. The cor-relation shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCE is maintained between the limits specified in Specifica-tion 3.2.1.

The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum al-lowable control rod insertion and are the core DNBR design basis.

Therefore, for operation at a fraction of RATED THERMAL POWER, the design limits are met.

Whenusgngincoredetectorstomakepowerdistributionmapstodeter-mine F and F q

AH Meas The measurement of total peaking factor F shall be increased by 1.4 a.

percenttoaccountformanufacturingtolehance,sandfurtherincreasedby i

7.5 percent to account for measurement error.

DAVIS-BESSE, UNIT 1 B 3/4 2-2 Amendment No. JJ, 61 i _..

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ADMINISTRATIVE CONTROLS power operation), supplementary reports shall_ be submitted at least every three months until all three events have been completed.

ANNUAL OPERATING REPORT 6.9.1.4_ Annual reports covering the activities of the unit during the previous calendar year shall be submitted prior to March 31 of each year.

l 6.9.1.5 Reports. required on an annual basis shall include:

e. : A tabulation on an annual basis of the number of station, utility and other personnel (including contractors). receiving exposures greater than 100 mrem /yr and their associated man ran exposure according to work and job functionsl/, e.g., reactor l

operations and surveillance, inservice inspection, routine maintenance,specialmaintenance(describedmaintenance), waste processing,,and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20% of the individual total dose need not be accounted-for.- In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work

-functions.

b.

The complete results of steam generator tube inservice inspections (Specification 4.4.5.5.b).

c.

The results of specific activity analysis in which-the primary coolant exceeded the limits of' Specification 3.4.8.

The following infonnation shall-be included:

(1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in.which the limit was exceeded; (4) Graph of the I-131 concen-tration and one other radiciodine isotope concentration in J/ This tabulation supplements the requirements of 620.407 of 10 CFR l

Part 20.

DAVIS-BESSE, UNIT 1 6-15 Amendment No. 9.J2,4J,52,77, 87,706,135

o ADMINISTRATIVE CONTROLS microcuries per gram as'a function of time for the duration of the specific betivity above the steady-state level; and (5) The time duration when the specific sctivity of the primary coolant exceeded A

the radioiodine limit.

MONTHLY OPERATING REPORT 6.9.1.6. Routine reports of operating sts.tistics, shutdown experience and challenges to the Pressurizer Pilot Operated Relief Valve (PORV) and the.

Pressurizer Code Safety Valves shall be submitted on a monthly basis to arrive no later _than the 15th of each month following the calendar month covered by the report, as follows: The signed original to the Nuclear Regulatory Commission, Document Control Desk, Washington, D. C. 20555, and one copy each to the Region

.III-Adninistrator and the Davis-Besse Resident Inspector.

CORE OPERATING LIMITS REPORT 6.9.1.7 ' Core operating limits 'shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle and any remaining part of a reload cycle for-the following:

3.1.3.6 Regulating Rod' Insertion Limits 3.1.3.7 Rod Program 3.1.3.8 Xenon Reactivity-3.1.3.9 Axial Pewer Shaping Rod Insertion Limits 3.2.1

. AXIAL P0>'ER IMBALANCE 3.2.4 QUADRANT POWER TILT The analytical methods used to detennine'the core operating limits addressed by the individual Technical Specifications shall be those previously reviewed and approved by the NRC, specifically:

1)

BAW-10122A Rev. 1

" Normal Operating Controls," May 1984

.2)

BAW-10116A, " Assembly Calculations and Fitted Nuclear Data " May 1977 3)

BAW-10117P-A, " Babcock & Wilcox Version of PDQ User's Manual,"

January 1977 4)

BAW-10118A, " Core Calculational Techniques and Procedures,"

December 1979 5)

BAW-10124A, " FLAME 3 - A Three-Dimensional Nodal Code for Calculating Core Reactivity and Power Distributions," August 1976 6)

BAW-10125A, " Verification 'of Three-Dimensional FLAME Code " August 1976 7)

BAW-10152A, "N0ODLE - A Multi-Dimensional Two-Group Reactor Simulator " June 1985 DAVIS-BESSE, UNIT 1 6-16 Amendment No. 8,J2,93,J04,J35,144

f s. :.

', *.Jf M

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT' (Continued) 8)-

BAW-10119, " Power Peaking Nuclear Reliability Factors," June 1977 The methodology for Rod Program received NRC approval in the Safety Evaluation dated January.11,1990.

The core! 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.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revision or supple-ments 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.-

i DAVIS-BESSE, UNIT 1 6-17 Amendment No.144

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ADMINISTRATIVE' CONTROLS-

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ANNUAL" RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT' u;

6.9.1.10: Routine Radiological Environmental 10perating Reports covering ~ the -

operation of the unit during the-previous calendar year shall' be submitted prior to.May.1 of each year. The initial report shall be submitted prior to May 1 of:the year following initial criticality.

' ~

The Annual: Radiological Environmental Operating Reports shall include summaries,~ interpretations, and an analysis of trends of the-results of the radiological environmental surveillance. activities for the report period, including'a comparison ~with the preoperational studies with operational

'q controls, as appropriate-and with previous environmental surveillance reports ^ and an assessment of the observed impacts of the plant' operation on the, environment.- The reports shall also include the results of land s

use censuses required-by Specification 3.12.2.

Thb Annual Radiological Environmental: Operating Reports shall include the

~

results of, analysis of all radiological environmental samples and of all-environmental ~ radiation measurements taken during the period pursuant to the locations specified in the Table and Figures in the ODCM, as well as.summar-ized:and= tabulated.results'of.these analyses and measurements.

In the event that some individual results are not available for inclusion with the report,

~

b the report shall be submitted noting and explaining the reasons for-the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

s The reports shall also include the following: a summary description of the radiological environmental monitoring program; et least two legible maps covering all sampling locations keyed to a table giving distances and directions'from the centerline of one reactor; the results of licensee participation'in the Interlaboratory Comparison Program, required by Specification 3.12.3; and discussion of all analyses in which the LLD required'by Table 4._12-1 was not achievable.

4 DAVIS-BESSE, UNIT 1 6-17a Amendment No. 25,93 l

ML20005G042

Text

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