Text

Amend 218 to License NPF-3,extending SR Intervals from 18 to 24 Months Based on Results of Plant Instrument Drift Study
	ML20203B185
Person / Time
Site:	Davis Besse
Issue date:	12/02/1997
From:	Hansen A; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20203B189	List: ML20203B185; ML20203B214;
References
	NUDOCS 9712120364
	Download: ML20203B185 (37)
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k UNITED STATES y*

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30e46-0001

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IQLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY hNQ THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS BESSE NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATiyG_ LICENSE Amendment No. 218 License No. NPF 3 1.

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

A.

The applications for amendment by the Toledo Edison Company, Centerior Service Company, and The Cleveland Electric illuminating Company (the licensees) dated December 11,1996 (as supplemented by letter dated January 6,1997), January 30,1997 (as supplemented by letter dated September 15,1B j7), and April 18,1997, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the

(

Act, and the rules and regulations of the Cornmission; l

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 satis'ied.

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 ao follows:

-9712120364 971202 1.

PDR ADOCK 05000346

- -. -. - - - - = _ - -

2-(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amersdment No. 218. 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 120 days after issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Allen G. Hansen, Project Manager Project Directorate lll 3 Division of fiaactor Projects Ill/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specificationc Date of issuance: December 2, 1997

ATTACHMENT TO LICENSE AMENDMENT NO. 218 FACillTY OPERATING llCENSE NO. NPF-3 DOCKET NO. 50-346 Peplace 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 Insert TS 1-8 TS 1-8 TS 2-4 through TS 2-4 through 2-6 (3 pages) 2-6 (3 pages)

TS Bases 2-4 through TS Bases 2-4 through 2-7 (4 pages) 2-7 (4 pages)

TS 3/4 3-1 TS 3/4 3-1 15 3/4 3-7 through TS 3/4 3-7 through 3/4 3-9 (3 pages) 3/4 3-9 (3 pages)

TS 3/4 3-12 TS 3/4 3-12 TS 3/4 3-12a TS 3/4 3-12a TS 3/4 3-13 TS 3/4 3-13 TS 3/4 3-21 through 1S 3/4 3-21 through 3/4 3-23 (3 pages) 3/4 3-23 (3 pages)

TS 3/4 3-28 TS 3/4 3-28 TS 3/4 3-30 TS 3/4 3-30 TS 3/4 3-34 TS 3/4 3-34 TS 3/4 3-43 TS 3/4 3-43 TS 3/4 3-49 iS 3/4 3-49 TS 3/4 3-50 TS 3/4 3-50 TS 3/4 4-4 TS 3/4 4-4 TS 3/4 4-14 TS 3/4 4-14 TS 3/4 5-2 TS 3/4 5-2 TS 3/4 5-4 TS 3/4 5-4 TS 3/4 7-5 TS 3/4 7-5 TS 3/4 7-Sa TS 3/4 7-Sa TS Bast-s 3/4 3-1 TS Bases 3/4 3-1 TS Bases 3/4 3-la

'S Bases 3/4 3-la TS Baser. 3/4 5-2a TS Bases 3/4 5-2a TS Bases 3/4 5-2S TS Bases 3/4 5-2b

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TABLE 1.2 fRE0VENCY NOTATION EDTATION FRE0VENCY At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 6 months.*

E At least once per 18 months.*

l R

At least once per 24 months.*

l S /')

Prior to each reactor startup.

N/A Not applicable.

In these Technical Specifications, 6 months is defined to be 184 days, 18 months is defined to be 550 days, and 24 menths is defined to be 730 days.

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l l-DAVIS-BESSE, UNIT 1 1-8 Amendment No. 37,05,170, 218 l

I SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMIT _ING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM SETPOINTS 2.2.1 The Reactor Protection System instrumentation setpoints shall be set consistent with the Allowable Value:, shown in Table 2.2-1.

l APPLICABILITY: As shown for each channel in Table 3.3-1.

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With a Reactor Protection System instrumentation setpoint less conservative than the value shown in the Allowable Values column of Tablo 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the cht.;nel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Allowable Value.

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- DAVIS-BESSE, UNIT 1 2-4 Amend:nent No. 218

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<y Table 2.2-1 Reactor Protection System Instrumentation Trio Setooints ao 0;

Functional unit Allowable values ll M

1.

Manual reactor trip Not applicable.

l' E

i a

2.

High flux 5105.1% of RATED THERMAL POWER with t

four pumps operating

580.6% of RATED THERMAL POWER with three pumps operating

l 3.

RC high temperat.ure 5618'F*

l 7

4.

Flux --Aflux/ flow"'

Pump allowable values not to exceed the limit lines shown in the CORE OPERATING LIMITS REPORT for four and three pump operation.*

5.

RC low pressure"'

21900.0 psig*

lI 6.

RC high pressure 52355.0 psig*

l; 7.

RC pressure-temperature"'

2(16.00 T.ut*F - 7957.5) psig*

lI i

ga 8.

High flux /' number of RC

$55.1% of RATED THERMAL POWER with

., 3 pumps on' one pump operating in each foop*

rao[

50.0% of RATED THERMAL POWER with two pumps operating in one loop and wo P

no pumps operating in the other icop*

l 2';

P 50.,0% of RATED THERMAL POWER with no l'l*,

pumps operating or only one pump j

6 operating

52 "

~~<

9.

Containment pressure high

$4 psig*

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E Table 2.2-1.

(Cont'd)

Tir p my be manually bypassed when RCS pressure 51820 psig by actuating shutdown bypass provided that:

a.

The high flux trip setpoint f s $5% of RATED THERMAL POWER.

i b.

The shutdown bypass high pressure trip setpcint of $1820 psig is imposed.

c.

The shutdown bypass is removed when RCS pressure >1820 psig.

f

Allowable value for CHANNEL FtMCTIONAL TEST.

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l 2.2 LIMil.ING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The reactor protection system instrumentation Allowable Values specified in Table 2.2-1 lave been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.

The shutdown bypass provides for bypassing certain functions of the reactor protection system in order to permit control rod drive tests, zero power PHYSICS TESTS and certain startup and shutdown procedures. The purpose of the shutdown bypass high pressure trip is to prevent normal operation with shutdown bypass activated.

This high pressure setpoint is lower than the normal low pressure setpoint so that the reactor must be tripped before the bypass is initiated.

The high flux setpoint of 55.0% prevents any significant reactor power from being produced.

Sufficient natural circulation would be available to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumps were operating.

Manual Reactor Trio The manual reactor trip is a redundant channel to the automatic reactor protection system instrumentation channels and provides manual reactor trip capability, p

Hiah Flux A high flux trip at high power level (neutron flux) provides reactor core prote: tion against reactivity excursions which tre too rapid to be protected by temperature and pressure protective circuitry.

During normal station operation, reactor trip is initiated when the reactor power level reaches the Allowable Value < 105.1% of rated power.

Ddeto l

transient overshoot, heat balance, and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which was used in the safety analysis.

DAVIS-BESSE, UNIT 1 B 2-4 Amendment No. 45r417218

LIMJTING SAFETY SYSTEM SETTINGS

)

BASES

)

RC Hioh Temoerature The RC high temperature trip 5618'F prevents the reactor outlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients.

Flux -- AFlux/ Flow The power 1svel Allowable Value produced by the reactor coolant system flow is l

based on a flux-to-flow ratio which has been established to accommodate flow decreasing transients from high power where protection is not provided by the high flux / number of reactor coolant pumps on trips.

The power level Allowable Value produced by the power-to-flow ratio provides l

both high power level and low flow proter. tion in the event the reactor power level increases or the reactor coolant flow rate dec*.c 's.

The power level setpoint produced by the power-to-flow ratio provides c.erpower DNB protection for all modes of pump operation. For overy flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate.

For safety calculations the instrumentation errors for the power level were used.

Full flow rate is defined as the flow calculated by the heat balance at 100% power. At the time of the calibration the RCS flow will be greater than or equal to the value in Table 3.2-2.

DAVIS-BESSE, UNIi 1 B 2-5 Amendment No. 15,33,45,51,00,123,100, 218 l

1 l

. s LIMITING SAFET( SYSTEM SETTINGS BASES The AXfAL POWER IMBALANCE boundaries are established in order to prevent reactor thermal limits from being exceeded.

These tharmal limits are either power peaking kW/ft limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the ?ower level trip produced by a flux-to-flow ratio such that the beundaries of t.1e figure in the CORE OPERATING LIMITS REPORT are produced.

RC Pressure - Low. Hich. and Pressure Temperature The high and low trips are provided to limit the pressure range in which reacto operation is permitted.

During a slow reactivity insertion startup accident from low power or a slow reactivity insertion from high power, the RC high pressure set)oint is reached before the high flux setpoint.

The Allowable Value for RC hig1 pressure, 2255 l

psig, has been established to maintain the system pressure below the safety limit, 2750 psig, for any design transient.

The RC high pressure trip is backed up by the pressurizer code safety valves for RCS over pressure protection, and is therefore set lower than the set pressure for these valves, s 2525 psig.

The RC high pressure trip also backs up the high flux trip.

The RC low pressure, 1900.0 psig, and RC pressure-temperature (16.00 T 7957.5)psig,AllowableValueshavebeenestablishedtomaintaintheDR$ ratio l

greater than or equal to the minimum allowable DNB ratio for those design accidents that result in a pressure reduction, it also prevents reactor operation at pressures below the valid range of DNB correlation limits, protecting against DNB.

Hiah Flux / Number of Reactor Coolant Pumos Or1 In conjunction with the flux - aflux/ flow trip the high flux / number of reactor coolant pumps on trip prevents the minimum core DNBR from decreasing below the minimum allowable DNB ratio by tripping the reactor due to the loss of reactor coolant pump (s).

The pump monitors also restrict the power level for the number of pumps in operation.

DAVIS-BESSE, UNIT 1 B 2-6 Amendment No. 23,45,50,S1,149,189, 218

4 LIMITING SAFETY SYSTEM SETTINGS BASES (ontainment Hiah Pressure The Containment High Pressure Allowa' ale Value s 4 psig, provides positive l

assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of coolant accident, even in the absence of a RC Low Pressure trip.

DAVIS-BESSE, UNIT 1 B 2-7 Amendment No. 218 l

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r 3/4.3 INSTRUMENTAT10N 3/4.3.1 REACT 0". PROTECTION SYSTEM INSTRUMENTATION l

LIMITING CANDITION F6R OPERATION 3.3.1.1 As a minimue, the Reactor Protection System instrumentation channels 1

and bypasses of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

$PPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

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}URVEILLANCEREQUIREMENTS 4.3.1.1.1 Each Reactor Protection System instrumentation channel shall be I

demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CAllBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the l

frequencies shown in Table 4.3-1.

i 4.3.1.1.2 The total bypass function shall be demonstrated OPERABLE it least once per REFUELING INTERVAL during CHANNEL CAllBRATION testing of each channel l

.affected by bypass operation.

i 4.3.1.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per REFUELING INTERVAL.

Each test shall include at least one channel per function i

such that all channels.are tested at least once every N times the REFUELING INTERVAL where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

i 1

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-DAVIS-BESSE. UNIT 1 3/4 3-1 Amendment No. 218

3 TABLE 4.3-1 O

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REACTOR PROTECTION SYSTEM Ii4STRUNENTATION SURVEILLANCE REOUIREMENTS CHANNEL MODES IN WHICH R

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

_y FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED E

1.

Manual Reactor Trip N.A.

M.A.

5/U(1)

N.A.

2.

High Flux 5

0(2), and Q(6,9)

N.A.

1, 2 3.

RC High Temperature S

R SA(9) 1, 2 4.

Flux - AFlux - Flow S(4)

M(3) and Q(6,7,9)

N.A.

1, 2.

i 5.

RC Low Pressure S

R SA(9)

I, 2

{

6.

RC High Pressure S

R SA(9)

I, 2 7.

RC Pressure-Temperature S

R SA(9) 1, 2 8.

High Flux / Number of Reactor Coolant Pumps On S

Q(6,9)

N.A.

I, 2 9.

Containment High Pressure S

E SA(9)

I, 2 l

a i

2

10. Intermediate Range, Neutron 1

i E

Flux and Rate S

E(6)

N.A.(5) 1, 2 and

I z

P

11. Source Range, Neutron Flux y

and Rate S

E(6)

N.A.(5) 2, 3, 4 and 5 l

4 u

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12. Control Rod Drive N.A.

N.A.

M(8,9) and S/U(1)(8) I, 2 and *

. 9 Trip Breakers t

j$.

13. Reactor Trip Module Logic N.A.

N.A.

M(9) 1, 2 and

l jk

14. Shutdown Bypass High Pressure S R

SA(9) 2**, 3**,

4**, 5**

i

15. SCP. Relays M.A.

N.A.

R 1, 2 and *

=

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=

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I TABLE 4.3-1 (Continued)

NOTATION If not performed in previous 7 days.

(1)

Heat balance only, above 15% of RATED THERMAL POWER.

(2)

When THERMAL POWER (TP) is above 50% of RATED THERMAL POWER [RTP)

(3) and at a steady state, compare out-of-core measured AXIAL POWER as follows[, API.) to incore measured AXIAL POWER IMBALANCE [ API )

IMBALANCE e

RIE (API, - APl ) = Offset Error 4 -

i TP Recalibrate if the absolute value of the Offset Error is 12.5%.

AXIAL POWER IMBALANCE and loop flow indications only.

l (4)

CHANNEL FUNCTIONAL TEST is not applicable.

Verify at least one (5) decade overlap prior to each reactor startup if not verified in previous 7 days.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(6)

Flow rate measurement sensors may be excluded f rom CHANNEL (7)

CALIBRA110N.

However, each flow measurement sensor shall be calibrated at least once each REFUELING INTERVAL.

l The CHANNEL FUNCTIONAL TEST shall independently verify the (8)

OPrRABILITY of both the undervoltage and shunt trip devices of the Reactor Trip Breakers.

Perforn,ed on a STAGGERED TEST BASIS.

(9)

With any control rod drive trip breaker closed.

When Shutdown Bypass is actuated.

J 2

i DAVIS-BESSE, UNIT 1 3/4 3-8 Amendment No. 43,100,123.135.185r 218 4-

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i INSTRUMENTATION E d l FETY SYSTEM INSTRUMENTATION SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Safety Features Actuation System (SFAS) functional units shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip c tpoint column of Table 3.3-4. with the e

exception of Instrument Strings Functional Units d and e and Inter'ock Channels Functional Unit a which shall be set consistent with the Allowable Value column of Table 3.3-4, and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY:

As shown in Table 3.3-3.

ACILQB:

a.

With a SFAS functional unit trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the functional unit inoperable and apply the applicable ACTION requirement of Table 3.3-3, until the functional unit is restored to OPERABLE status with the trip setpoint adjusted censistent with Table 3.3-4.

l b.

With a 5FAS fu..ctional unit iraperable, take the action shown in Table 3.3-3.

SURVEILLANCE RE0VIREMENTS 4.3.2.1.1 Each SFAS functional unit shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CAllBRATION and CHANNEL FUNCTIONAL TEST during the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the bypas:es shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST o functional units affected by bypass operation.

The total bypass function shall be demonstrated OPERABLE at least once per REFUELING INTERVAL during CHANNEL CALIBRATION testing of each l

functional unit affected by bypass operation.

4.3.2.1.3 The SAFETY FEATURES RESPONSE TIME of each SFAS function shall be demonstrated to be within the limit at least once per REFUELING INTERVAL.

l Each test shall include at least one functional unit per function such that all functional units are tested at least once every N times the REFUELING INTERVAL where N is the total number of redundant functional units in a specific SFAS function as shown in the " Total No of Units" Column of Table 3.3-3.

DAVIS-8 ESSE, UNIT 1 3/4 3-9 Amendment No. 218

1 6

TABLE 3.3-3 (Continued)

TABLE NOTATION Trip function may be bypassed in this MODE with RCS pressure below 1800 psig.

Bypass shall be automatically removed when RCS pressure exceeds 1800 psig.

Trip function may be bypassed in this MODE with RCS pressure below 660 psig.

Bypass shall be automatically removed when RCS pressure exceeds 660 psig.

DELETED This instrumentation, or the containment purge and exhaust system noble gas monitor (with the containment purge and exhaust system in operation), must be OPERABLE during CORE ALTERATIONS or movement of irradiateo fuel within containment to meet the requirements of Technical Specification 3.9.4.

When using the containment purge and exhaust system noble gas monitor, SFAS is not required to be OPERABLE in MODE 6.

All functional units may be bypassed for up to one minute when starting each Reactor Coolant Pump or Circulating Water Pump.

When either Decay Heat Isolation Valve is open.

The provisions of Specification 3.0.4 are not applicable.

ACTION STATEMENTS ACTION 10 - With the number of OPERABLE functional units one less than the Total Number of Units, STARTUP and/or POWER OPERATION may proceed provided both of the following conditions are satisfied:

a.

The inoperable functional unit is placed in the tripped I

condition within one hour.

b.

The Minimum Units OPERAM E requirement is met; however, one additional functional unit may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specificaticn 4.3.2.1.1.

l ACTION 11 - With any component in the htpui logic inoperable, trip the associated components within oas hour or be in at least HOT STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months sud in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l l

DAVIS-BESSE, UNIT 1 3/4 3-12 Amendment No. 28,37,52,102, 218 135,150,ISS,?!!,

d I

a a

1 HBLE 3.3-3 (Continued)

'A;T[0N STATEMENTS ACTION 12 - With the number of OPERABLE Units one less than the Total Number of Units, restore the inoperable functional unit to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the l

next C hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 13 - a.

With less than the Minimum Units OPERABLE and indicated reactor coolant pressure ;t 328 psig, both Decay Heat Isolation Valves (DH11 and DH12) shall be verified closed.

b.

With less than the Minimum Units OPERABLE and indicated reactor coolant pressure < 328 psig operation may continue; however, the functional unit shall be OPERABLE prior to increasing indicated reactor coolant pressure above 328 l

psig.

ACTION 14 - With less than the Minimum Units OPERABLE and indicated reactor coolant pressure < 328 psig, ooeration may continue; however, the functional unit shall be OPEFAbi.E prior to increasing indicated reactor coolant pressure above 328 psig, or the inoperable functional unit shall be placed in the tripped state.

ACTION 15 - a.

With the number of OPERABLE units one less then the Minimum Units Operable per Bus, place the inoperable unit in the tripped condition within one hour.

For functional unit 4.a the sequencer shall be placed in the tripped condition by physical removal of the sequencer module.

The inoperable functional unit may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.1.

b.

With the number of OPERABLE units two less than the Minimum Units Operable per Bus, declare inoperable the Emergency Diesel Generator associated with the functional units not meeting the reqrired minimum units OPERABLE and take the ACTION required of Specification 3.8.1.1.

Amendment No. 28 f&l02,135.21+r 218 DAVIS-BESfE, UNIT 1 3/4 3-12a t

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co TABLE 3.3-4 0

y SAFETY FEATURES ACiUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUES i

O r

INSTRUMENT STRINGS l

a.

Containment Radiation

< 4 x Background at RATED

<4x Background at THERMAL POWE".

RATED THERMAL POWERe b.

Containment Pressure - High s 18.4 psia s 18.52 psias w

c.

Containment Pressure - High-High s 38.4 psia 5 38.52 psiae i

d.,RCS Pressure - Low N.A.

2 1576.2 psigee l

w e.

RCS Pressure - Low-Low N.A.

2 441.42 psiger l

f.

BWST Level 2 89.5 and s 100.5 in H2O 2 88.3 and s 101.7 in Hz0e SEQUENCE LOGIC CHANNELS I'

E a.

Essential Bus Feeder Breaker Trip (90%)

2 3744 volts for 2 3558 volts 5

s 7.8 sec s 7.8 sec I

?+

z

b. Diesel Generator Sta.-t, Load Shed on 2 2071 and s 2450 volts 2 2071 and s 2450 volts i

P Essential Bus (E9%)

for 0.5 i 0.1 sec for 0.510.1 sece i

g INTERLOCK CHANNELS

{

b Decay Heat Isolation Valve M.A.

< 328 psiger l

a.

g and Pressurizer Heater

~

h

Allowable Value for CHANhEL FUNCTIONAL TEST and CHANNEL CALIBRATION m

5

Referenced to the RCS Pressure instrumeatation tap.

l

1. Allowable Value for CHANNEL FUNCTIONAL TEST i

l m

l l

L 5l

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5 TABLE 4.3-2 M

SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE0UIREMENTS "m

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE l

c i

5 FUNCTIONAL UNIT CHECK CALIBRATION TEST

~ REQUIRED l

1.

INSTRUMENT STRINGS a.

Containment Radiation - High S

E M

1, 2, 3, 4, 6 #

~

l b.

Containment Pressure - High 5

E M(2)

I,2,3 c.

Containment Pressure - High-High S

E M(2)

I, 2, 3 d.

RCS Pressure - Low S

R M

I, 2, 3 e.

RCS Pressure - Low-Low S

R M

I, 2, 3 f.

BWST Level - Low-Low S

E M

I, 2, 3 l

2.

OUTPUT LOGIC w

a.

Incident Level #1: Containment 1

Isolation S

E M

I, 2, 3, 4, 6 #

l h.

Incident level #2: High Pressure w4 Injection and Starting Diesel Generators S

E M

I, 2, 3, 4 l

~

c.

Incident Level #3: Low Pressure Injection S

E M

I, 2, 3, 4 l

d.

Incident Level #4: Containment p

Spray S

E M

I, 2, 3, 4 l

e.

Incident Level #S: Containment E.

Sump 84.irct ' 2 *on Permissive 5

E M

I,2,3,4 l

2

?.

3.

MANUAL ACTUAT ON l

a.

SFAS (Except Conta.. mnt Spray NA NA M(I)

I, 2, 3, 4, 6 #

z

?

and En gency Sump Rccirculation) e b.

Containment Spray NA NA M(I)

I, 2, 3 3

4.

SEQUENrE LOGIC CliANNELS S

NA M

I, 2, 3, 4 i

-T

3 m

o l

o

~

E G

TABLE 4.3-2 (Continued)

E

.{

G SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS

.M i

E CHANNEL MODES IN WHICH Z

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE l

j FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED j

5.

INTERLOCK CHANNELS i -.

I a.

Decay Heat Isolation Valve S

R 1, 2, 3 i

b.

Pressurizer Heater 5

R 3 ##

w i

1 w

l g

- See Specification 4.5.2.d.I a

I TABLE NOTATION t-

- t

]

(1) Nanual aduation switches shall be tested at least once per REFUELING INTERVAL. All other l

l k

circuitry associated with manual safeguards actuation shall receive a CHANNEL FUNCTIOCAL TEST f

i 1

E at least once per 31 days.

2

?.

(2) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either vacuum or pressure to the appropriate side of the transmitter.

z P

f j

f These surveillance requirements in conjunction with those of Section 4.9.4 apply durine CORE m

c P

ALTERATIONS or movement of irradiated fuel within the containment only if using the SFAS area M

radiation monitors listed in Table 3.3-3, Items Ia, Za, and 3a, in lieu of the containment l

L purge and exhaust system noble gas monitor.

-4 When either Decay Heat Isolation Valve is open.

f 4y U

v

?

l

~

5 I

= -

a

+

~

wn w

m e

w-w

INSTRUMENTATION SJEAM AND FEEDWATER RUPTURE CONTROL SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.2 The Steam and Feedwater Rupture Control System (SFRCS) instrumentation channels shown in Table 3.3-11 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-12, with the exception of the Steam Generator Level-Low Functional Unit which shall be set consistent with the Allowable Value column i

of Table 3.3-12, and with RESPONSE TIMES as shown in Table 3.3-13.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

a.

With a SFRCS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-12, declare the channel inoperable and cpply the applicable ACTION requirement of Table 3.3-11, until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with Table 3.3-12.

l b.

With a SFRCS instrumentation channel inoperable, take the action shown in Table 3.3-11.

SURVElttANCE RE0VIREMENTS 4.3.2.2.1 Each SFRCS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST during the MODES and at the frequencies shown in Table 4.3-11.

4.3.2.2.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation.

The total bypass function shall be demonstrated OPERABLE at least once per REFUELING INTERVAL during CHANNEL CALIBRATION testing of each channel affected l

by bypass operation.

4.3.2.2.3 The STEAM AND FEEDWATER RUPTURE CONTROL SYSTEM RESPONSE TIME of each SFRCS function shall be demonstrated to be within the limit at least once per REFUELING INTERVAL.

Each test shall include at least one channel per function such that all channels are tested at least once every N times the REFUELING INTERVAL where N is the total number of redundant channels in a specific SFRCS function as shown in the " Total No. of Channels" Column of Table 3.3-11.

DAVIS-BESSE, UNIT 1 3/4 3-23 Amendment No. 218

i l

5?

G6 C?

TABLE 3.3-12

m STEAN AND FEEDWATER RUPTURE CONTROL SYSTEM

~

Ei INSTRUMENTATION TRIP SETPOINTS

-4 FUNCTIONAL UNITS TRIP SETPOINTS ALLOWABLE VAlbio-1.

Steam Line Pressure - Low 2 591.6 psig 2 591.6 psig*

2 586.6 psig**

I 2.

Steam Generator Level - Low"'

N.A.

2 16.9"*

3.

Steam Generator Feedwater Of Differential Pressure - High(2' s 197.6 paid s 197.6 psid*

$ 199.6 psid**

Y E$

4.

Reactor Coolant Pumps - Loss of High s 1384.6 amps s 1384.6 ampsi Low 2 106.5 amps 2 106.5 ampst "3 Actaal water level above the lower steam generator tubesheet.

c23 Where differential pressure is steau generator minus feedwater pressure.

EL

Allowable Value for CHANNEL FUNCTIONAL TEST 2

- Allowable Value for CHANNEL CALIBRATION

?.

Allowable Value for CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION E

J Me

~

.' <. fgg

.e a

3

5 Y'

TABLE 4.3-11 M

J' STEAM AND FEEDWATER RUPTURE CONTROL SYSTE71 p%$t i

INSTRUMENTATION SURVEILLANCE RL0tilREMFNTS g:p g

A$fh U

CHANNEL

~

CHANNEL C3Ai4NEL FUNCTIONAL FUNCTIONAL UNIT CHECK CALIBRATION TEST 1.

Instrunant Channel a.

Steam Line Pressure - Low S

E M

l l

w b.

Steam Generator level - Low S

R M

w c.

Steam Getierator - Feedwater S

E N

l j

Differential Pressure - High d.

Reactor Coolant Fumps - Loss of S

E M

l 2.

Mar.ual Actuation NA NA R

a n

.E W

G 5

n 2

m

l TABLE 4.3-3 g

1 I

G RADIATION M0:11TORING INSTRUMEN'lATION SURVEILLANCE REQUIREMENTS

/

6

'w" M

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E

INSTRUMENT CHECK CALIBRATION TEST REOUIRED M

1.

AREA MONITORS a.

Fuel Storage Pool Area Emergency Ventilation l

System Actuation S

E M

2.

PROCESS MONITORS w

a.

Containment

'a 1.

Gaseous Activity RCS Leakage Detection S

E M

1, 2, 3 & 4 j

f w

w

11. Particulate Activity RCS Leakage Detection S

E M

1, 2, 3 & 4 l

1 E

E

- With fuel in the storage pool or building O

I L

l

C INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTA110N LIMITING CONDITION FOR OPERATION 3.3.3.5.1 The remote shutdown monitoring instrumentation channels shown in Table 3.3-3 shall be OPERABLE with readouts displayed external to the control room.

3.3.3.5.2 The control circuits and transfer switches required for a serious control room or cable spreading room fire shall be OPERABLE.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

a.

With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, either restore the inoperable channel to OPERABLE status within 30 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

With one or more controi circuits or transfer switches required for a serious control room or cable spreading room fire inoperable, restore the inoperab'e circuit (s) or switch (es) to OPERABLE status within 30 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restering the circuit (s) or switch (es) to OPERABLE status, c.

The provisions of Specification 3.0.4 are not applicable.

S SURV.ElllANCE RE0VIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel shall be demonttreted OPERABLE by performance af the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.

4.3.3.5.2 At least once per REFUELING INTERVAL, verify each control circuit l

and transfer switch required for a serious control room or cable spreading room fire is capable of performing the intended function.

DAVIS-BESSE, UNIT 1 3/4 3-43 Amendment No. 1877 218

a,-

W e

TABLE 4.3-10 8

5 POST-ACCIDENT MONITORING INSTktMENTATION SURVEllLANCE REOUIREMENTS

i CRANNEL CHANNE:

El INSTRUMENT CHE(K CALIBRAT_JN 1.

SG Outlet Steam Pressure M

R

,g 2.

RC Loop Outlet Temperature M

R 3.

RC Loop Pressure M

R 4.

Pressurizer Level N

R 5.

SG Startup Range Level M

R w1 6.

Containment Vessel Post-Accident Radiation wi a.) Containment High Range Radiatior, M

8'.

b.) Containment Wide Range Noble Gas M

E 7.

High Pressure Injection Flow M

E l

r$

8.

Low Pressure Injection (DHR) Flow M

E l

P E

5 9.

Auxiliary Feedwater Fluw Rate M

E l

1, 10.

RC System Subcooling Margin Monitor M

R

?

11.

PORV Position Indicator M

R 53 12.

PORV Block Valve Position Indicator M

R N

13.

Pressurizer Safety Valve Positien Indicator M

R is I

14.

BWST Levei S

E l

14 I

15.

Containment Normal Sump Level M

R 13 L

16.

Containment Wide Range Water Level M

R 5

+

.\\

l E'

l

s Y'5 TABLE 4.3-10 (Continued)

w

_POSI-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

CHANNEL CHANNEL CHECK CALIBRATION

~

INSTRUMENT 17.

Containment Wide Range Pressure M

R M

E l

18.

Incore Thermocouples 19.

Reactor Coolant Hot Leg Level (Wide Range)

M R

M E**

l w

20.

Neutron Flux (Wide Range) 1 M

E**

l w

21.

Neutron Flux (Source Range) b

- Neutron detectors may be excluded from CHANNEL CALIBRATION.

II

!l[L i:-

if L

0 L

ij L

5; I

l t

(

t D

i ftEACTOR COOLANT SYSTEM SAFETY VALVES AND PILOT OPERATED RELIEF VALVE - OPERATING LIMITING CONDITION FOR OPERATION 3.4.1 All pressurizer code safety valves shall be OPEPABLE with a lift setting of s 2525 psig.* When not isolated, the pressurizer pilot operated relief valve shall have a trip setpoint of 2 2435 psig and an allowable value of 22435 psig.**

APPLICABILITv: MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVElLLANCE RE0VIREMENTS 4.4.3 For the pressurizer code safety valves, there are no additional Surveillance Requirements other than those required by Specification 4.0.5.

For the pressurizer pilot operated relief valvo a CHANNEL CALIBRATION check 4

shall be performed each REFUELING INTERVAL.

l The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

Allowable value for CHANNEL CALIBRATION check.

DAVIS-BESSE, UNIT 1 3/4 4-4 Amendment No. 33,50,128,135, 218

REACTOR COOLANT' SYSTEM SURVEILLANCE RE0VIAEMENTS (Continued) b.

Containment sump level and flow monitoring system-performance of CHANNEL CAllBRATION at least once each REFUELING INTERVAL.

l c.

Containment atmosphere gaseous monitoring system-performance of CHANNEL CHECK, CHANNEL CAllBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in lable 4.3-3.

DAVIS-BESSE, UNIT 1 3/4 4-14 Amendment No. 218

O EMERGENCY CORE COOLING SYSTEMS SURVElllANCE RE0VIREMENTS (Continued) l b.

At least one.e per 31 days, and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of each solution volume increase of it 80 gallons that is not the result of addition from the borated water storage tank (BWST), by verifying the boron concentration of the CFT solution, c.

At least once per 31 days by verifying that power to the i

isolation valve operator is disconnected by locking the breakers l

in the open position.

d.

At least once per REFUELING INTERVAL by verifying that each core l

flooding tank isolation valve opens automatically and is interlocked against closing whenevor the Reactor Coolant System pressure exceeds 800 psig.

DAVIS-? ESSE, UNIT 1 3/4 5-2 Amendment No. % 218

Revised by NRC Letter Dated June 6, 1995 SURVEILLANCE REQUIREMENTS (Continued) b.

At least once each REFUELING INTERVAL, or prior to operation after ECCS piping has been drained by verifying that the ECCS piping is full of water by venting the ECCS pump casings and discharge piping high points.

c.

By a visual inspection which verifies that no loose debris (rags, trath, clothing, etc.) is present in the containment which could be transported to the containment emergency sump and cause restriction of the pump suction during LOCA conditions. This visual inspect % shall be performed:

1.

For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2.

For all areas of containment affected by an entry, at least once daily while work is ongoing and again during the final exit after completion of work (containment closeout) when CONTAINMENT INTEGRITY is established, 3

d.

At least once each REFUELING INTERVAL by:

1.

Verifying that t;.e interlocks:

a)

Close DH-ll and DH-12 and deenergize the pressurizer heaters, if either DH-11 or DH-12 is open and a simulated reactor coolant system pressure which is greater than the Allowable Value (<328 psig) is l

applied.

The interlock to close DH-11 and/or DH-12 is not required if the valve is closed and 480 V AC power is disconnected from its motor operators.

b)

Frevent the opening of DH-11 an<i DH-12 when a simulated or actual reactor coolant system pressure which is greater than the Allowable Value (<328 psig) l is applied.

2.

a)

A visual inspection of the containment emergency sump which verifies that the subsystem suction inlets are not restricted by debris and that tha sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

b)

Verifying that on a Borated Water Storage Tank (BWST)

Low-Low Level interlock trip, with the motor operators for the BWST outlet isolation valves and the containment emergency sump recirculation valves energized, the BWST Outlet Valve HV-DH7A (HV-DH78) automatically close in 575 seconds after the operator manually pushes the control switch to open the Contr.inment Emergency Sump Valve HV-DH9A (HV-DH9B) which should be verified to open in $75 seconds.

3.

Deleted DAVIS-BESSE, UNIT 1 3/4 5-4 Amendment No. h W,2 b '0,77,135r 132,105,105,200,21',,215,218

t l

i PLANT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) b.

At least once per 31 days on a STAGGERED TEST BASIS by:

1.

Verifying that each valve (power operated or automatic) in the flow path is in its correct position.

2.

Verifying that all manual valves in the auxiliary feedwater pump suction and discharge lines that affect the system's capacity to deliver water to the steam generator are locked in their proper position.

3.

Verifying that valves CW 196, CW 197, FW 32, FW 91 and FW 106 are closed, c.

At least once each REFUELING INTERVAL by:

1.

Verifying that each automatic valve in the flow path actuates to its correct position on a Steam and Feedwater Rupture Control System actuation test signal.

2.

Verifying that each pump starts automatically upon receipt of a Steam and Feedwater Rupture Control System actuation test signal.

The provisions of Specification 4.0.4 are not applicable for entry in MODE 3.

3.

Verifying that there is a flow path from each auxiliary feedwater pump to both steam generators by pumping water from the Condensate Storage Tank with each pump to both steam generators.

The flow paths shall be verified by either stea,a generator level change or Auxiliary Feedwater Safety Grade Flow Indication.

Verification of the Auxiliary Feedwater System's flow capacity is not required.

d.

The Auxiliary Feed Pump Turbine Steam Generator Level Control System shall be demonstrated OPERABLE by performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a CHANNEL FilNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once eac'.i REFUELING INTERVAL.

l e.

The Auxiliary Feed Pump Suction Pressure Interlocks shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATICN at least once each REFUELING INTERVAL.

l CAVIS-BESSE, UNIT 1 3/4 7-5 Amendment No. 42,53,05,122,131,19F,41+r218

)

I

11 ANT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) f.

After any modification or repair to the Auxiliary Feedwater System tnat could affect the system's capability to deliver water to the steam generator, the affected flow path shall be demonstrated available as follows:

1.

If the modification or repair is downstream of the test flow line, each auxiliary feed pump (s) associated with the affected flow path shall pump water from the Condensate Storage Tank to the steam generator (s) associated with the affected flow path; and the flow path availability will be verified by steam generator level change or Auxiliary Feedwater Safety Grade Flow Indication.

2.

If the modification or repair is upstream of the test flow line, the auxiliary feed pump shall pump water through t.u Auxiliary Feedwater System to the test flow line; and the flow path availability will be verified by flow indication in the test flow line.*

This Surveillance Testing shall be performed prior to entering MODE 3 if the modification is made in MODES 4, 5 or 6.

Verification of the Auxiliary Feedwater System's flow capacity is not required.

g.

Following each extended cold shutdown (> 30 days in MODE 5), by:

1.

Verifying that there is a flow path from each auxiliary feedwater pump to both steam generators by pumping Condensate Storage Tank water with each pump to both steam generators. The flow paths shall be verified by either steam generator level change or Auxiliary Feedwater Safety Grade Flow Indication.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

Verification of the Auxiliary Feedwater System's flow capacity is not required.

4.7.1.2.2 The Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks shall be demonstrated OPERABLE when the steam line pressure is greater than 275 psig, by performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once each REFUELING INTERVAL. The CHANNEL FUNCTIONAL TEST shall l

be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 275 psig during each plant startup, if the test has not been performed within the last 31 days.

When conducting u sts of an auxiliary feedwater train in MODES 1, 2, and 3 which require local manual realignment of valves that make the train inoperable, the Motor Driven Feedwater Pump and its essociated flow paths shall be OPERABLE per Specification 3.7.1.7 during the performance of this surveillance.

If the Motor Drivan Feedvater Pump or an associated flow path is inoperable, a dedicated individual shall be stationed at the realigned auxiliary fcedwater train's valves (in communication with the control room) able tc restore the valves tc normal sy:: tem OPERABLE status.

DAVIS-BESSE, UNIT 1 3/4 7-Sa Amendment No. 05,131,193, 218

3/4.3 INSTPUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM INSTRUMENTATION The OPERABILITY of the RPS, SFAS and SFRCS instrumentation systems ensure that 1) tLe associated action and/or trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be ot.t of service for testing or maintenance, and 4) sufficient systen functional capability is available for RPS, SFAS and SFRCS purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptiens used in the accident analyses.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

For the RPS, SFAS Table 3.3-4 Functional Unit Instrument String. 4 and e and Interlock Channel a, ana SFRCS Table 3.3-12 Functional lhit 2:

Only the Allowable Value is specified for each Function. Nominal trip setpoints are specified in the setpoint analysis. The nominal trip setpoints are selected to ensure the setpoints measured by CHANNEL FUNCTIONAL TESTS do not exceed the Allowable Value if the bist ble is performing a:, required. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable provided that operation and testing are consistent with the assumptions of the specific setpcint calculations.

Each Allowable Value specified is more conservative than the analytical limit assumed in the safety analysis to account for in;trument uncertainties approp'iate to the trip parameter.

These uncertainties are defined in the specific setpoint analysis.

A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.

Setpoints must be found within the specified Allowable Value Any setpoint adjustment shall be consistent with the assumptions of the current specific setpoint analysis.

A CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The test verifies that the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift to ensure that the instrument channel remains operational between successive tests. CHANNEL CALIBRATION shall find that measurement errors and bistable setpoint errors are witFin the assumptions of the setpoint analysis. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint analysis.

The frequency is justified by the assumption of an 18 or 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.

DAVIS-BESSE, UNIT 1 B 3/4 3-1 Amendment No. 73,125,128,211,218 (Next page is B3/4 3-la)

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM

~

INSTRUMENTATION (Continued)

The measurement of response time at the specified frequencies provides assurance that the RPS, SFAS, and SFRCS action function associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response tirr.e as defined.

Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified respanse times.

The actuation logic for Functional Units 4.a., 4.b., and 4.c. of Table 3.3-3, Safety Features Actuation System Instrurentation, is designed to provide protection and actuation of a single train of safety features equipment, ossential bus or emergency diesel generator.

Collectively, Functional Units 4.a., 4.b., and 4.c. function to detect a degraded voltage condition on either of the two 4160 volt essential buses, shed connected loads, disconnect the affected bus (es) from the offsite power source and start the associated emergency diesel generator.

In addition, if an SFAS cctuation signal is present under these conditions, the sequencer channels for the two SFAS channels which actuate the train of safety features equipment powered by the affected bu.i will automatically sequence these loads onto the bus to prevent overloading of the emergency diesel genert.br. Functional Unit 4.a. has a total of four units, one associated with each SFAS channel (i.e., two for each essential bus).

Functional Units 4.b. and 4.c. each have a total of four units, (two associated with each essential cus); each unit consisting of two undervoltage relays and an auxilie relay.

An SFRCS channel consists of 1) the sensing device (s), 2) associated logic and output rela,., (including Isolation of Main Feedwater Non Essential Valves and Turbine Trip), and 3) power sources.

The SFRCS response time for the turbine stop valve closure is based on the combined response times of main steam line low pressure sensors. logic cabinet delay for main steam line low pressure signals and closure timt ei the turbine stop valves. This SFRCS response time ensures that the auxiliary feedwater to the unaffected steam generator will act be isolated due to a SFRCS low pressure trip during a main steam line break accident.

Safety-grade anticipatory reactor 1. rip is initiated by a turbine trip (above 45 percent of RATED THERMAL POWER) or trip of both main feedwater pump turbines. This anticipatory trip will operate in advance of the reactor coolant system high pressure reactor trip to reduce the peak reactor coolant system pressure and thus reduce challenges to the pilot operated relief valve.

This anticipatory reactor trip system was installed to satisfy Item II.K.2.10 of NUREG-0737.

The justification for the ARTS turbine trip arming level of 45% is Given in BAW-1893, October, 1985.

DAVIS-BESSE, UNIT 1 B 3/4 3-la Amendment No. 73,125,128,135,211,218

EMERGENCY CORE COOllHG SYSTEMS BASES (Continued)

Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration.

(2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

Containment Emergency Sump Recirculation Valves DH-9A and DH-9B are de-energized during MODES 1. 2, 3 and 4 to preclude postulated inadvertent 2

opening of the valves in the event of a Control Room fire, which could result in draining the Borated Water Storage Tank to the Containment Emergency Sump and the loss of this water source for normal plant shutdown.

Re-energization of DH-9A and DH-9B is permitted on an intermittent basis during MODES 1, 2, 3 and 4 under administrative controls.

Station procedures identify the precautions which must be taken when re-energizing these valves under such Controls.

Borated Water Storage Tank (BWST) cutlet isolation valves DH-7A and DH-78 are de-energized during MODES 1, 2, 3, and 4 to preclude postulated inadvertent closure of the valves in the event of a fire, which could result in a loss of the availability of the BWST.

Re-energization of valves DH-7A and DH-7B is permitted on an intermittent basis during MODES 1, 2, 3, and 4 under administrative controls. Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.

The Decay Heat Isolation Valve and Pressurizer Heater Interlock setpoint is based on preventing over-pressurization of the Decay Heat Removal System normal suction line piping. Tia value stated is the RCS pressure at the sensing instrument's tap.

It has been adjusted to reflect the elevation difference between tue sensor's location and the pipe of concern.

3/4.5.4 BORATED WATER STORAGE TANK The OPERABILITY of the borated water storage tank (BWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA.

The limits en the BWST minimum volume and boron concentration ensure that:

sufficient water is availabl2 within containment to permit recirculation cooling flow to the core following manual switchover to the recirculation mode, and DAVIS-BESSE, UNIT 1 B 3/4 5-2a Amendment No. 191,207,215, 218

EMERGENCY CORE COOLING SYSTEMS BASES (Continued) remain at least 1% ak/k subcritical in the cold The reactor wil)F, xenon free, while only crediting 50% of the 2) condition at 70 control rods' worth following mixing of the BWST and the RCS water volumes.

These assumptions ensure that the reactor remains suberitical in the cold condition following mixing of the BWST and the RCS water volumes.

With either the BWST boron concentration or BWST borated water temperature not within limits, the condition must be corrected in eight hours.

The eight hour limit to restore the temperature or boron concentration to within limits was developed considering the time required to change boron concentration or temperature and assuming that the contents of the BWST are still available for injection.

The bottom 4 inches of the BWST are not available, and the instrumentation is calibrated to reflect the available volume. The limits on water volume, and boron concentration ensure a pH value of between 7.0 and 11.0 of the solution sprayed within the containment after a design basis accident.

The pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion cracking on mechanical systems and componentt.

DAVIS-BESSE, UNIT 1 B 3/4 5-2b Amendment No. 401,215, 218 l

_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ __________

ML20203B185

Text

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