ML20065P694

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Application to Amend License NPF-3,revising Tech Specs to Reflect Changes to Setpoints for Pressurizer Electromatic Relief & Code Safety Valves,Auxiliary Feedwater Sys & Trip Setpoints for Borated Water Storage Tank Level
ML20065P694
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 10/14/1982
From: Murray T
TOLEDO EDISON CO.
To: Stolz J
Office of Nuclear Reactor Regulation
References
862, NUDOCS 8210260283
Download: ML20065P694 (18)
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TOLEDO

%sm EDISON RCHAAS CA&JSE Docket No. 50-346 License No. NPF-3 UU" Serial No. 862 msi ass. sari October 14, 1982 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz Operating Reactor Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Stolz:

Under separate cover, we are transmitting three (3) original and forty (40) conformed copies of an application for Amendment to Facility Operating License No. NPF-3 for the Davis-Besse Nuclear Power Station Unit No. 1.

This application requests that the Davis-Besse Nuclear Power Station Unit 1 Technical Specifications, Appendix A, be revised to reflect the changes attached. The proposed changes include:

1. Changes in Section 3.4.3 and Bases ,
2. Changes in Bases B3/4.7.1.2
3. Changes in Table 3.3-4 The three attachments identify the proposed changes and their safety evaluations. Item 1 concerns the setpoints for pressurizer electromatic relief and code safety valves. Item 2 concerns a revision to the bases on Auxiliary Feedwater System. Item 3 concerns a change in trip setpoints for BWST level and Essential Bus Feeder Breaker Trip to account for instrument error.

This amendment requests two changes of Class III type. Item 2 changes in the bases are for update proposes only since the bases are not part of the technical specifications pursuant to 10CFR50.36(a). Therefore, it is determined to be a Class III amendment request and enclosed is a check for 8,000 as requested by 10CFR170.22.

Very truly yours, f P%/y Attachment 90 pp'g[

cc. ...,...s1d_zns;.ctor THE TOLEDO ED! SON COMPANY EOISON PLAZA 300 MADISON AVENUE TOLEDO OHIO 43652 bu/R

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8210260283 821014 PDR ADOCK 05000346 P PDR

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APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1 Enclosed are forty-three (43) copies of the requested changes to the Davis-Besse Nuclear Power Station Unit No. 1 Facility Operating License No. NPF-3, together with the Safety Evalution for the requested change.

The proposed changes include:

1. Changes in Section 3.4.3 and Bases
2. Changes in Bases B3/4.7.1.2
3. Changes in Table 3.3-4 By /M b Station Sdperintendent p For R. P. Crouse Vice President, Nuclear Sworn and subscribed before me this 14th day of October, 1982. i

% w 0 Notary Publi/

NORA LYNN FL000 Notary PuWic. State of Ohio My Comensson Empires Sept.1.1987 l

Docket No. 50-346 License No. NPF-3 Serial No. 862 October 14, 1982 Attachment 1 I. Changes to Davis-Besae Nuclear Power Station Unit 1, Appendix A Technical Specifications Section 3.4.3 and Bases.

A. Time required to Implement This change is to be effective upon NRC approval and implementation by Toledo Edison.

B. Reason for Change (Facility Change Request 79-348 Rev. A)

To provide increased margin between the PORV and Code Safety valve setpoints. This increased margin is facilitated by the relocation of pressurizer code safety valves.

C. Safety Evaluation See Attached

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SAFETY EVALUATION This amendment request proposes changes to the PORV and the Pressurizer Code Safety Valve setpoints in the Technical Specifications in light of the relocation of code safety valves to the top of the pressurizer.

The safety function of the pressurizer code safety valves is to provide RCS over pressure protection in accordance with the ASME code. PORV is only used to prevent the opening of the code safety valves.

The accident analysis assumes that the Pressurizer code safety valves open at 2500 psig + 3%. In the original configuration there was 35 feet of piping between the pressurizer and the safety valve. The full flow pressure drop was 65 psig in this piping. This piping and the pressure drop are being eliminated by the relocation of code safety valves, therefore, the Technical Specification code safety valve setpoint can be raised by 65 psi (from 2435 to 2500 psig). With a 1% tolerance on the setpoint, the Technical Specification requirement should be <2525 psig.

This setpoint can accomodate more margins to avoid PORV from challenging the safety valves as shown in the following table.

Prior to Moving After Moving Proposed New Pressurizer Pressurizer Technical Code Safety Code Safety Specification Valves Valves Trip Setpoints Code Safety Valve Setpoint (psig) 2435 + 1% 2435 + 1% 2500 + 1%

. Reactor Coolant System pressure with full flow through the code safety valves (psig) 250v + 1% 2435 + 1% 2500 + 1%

PORV open setpoint (psig) 2390 2390 2390 Margin between code safety valve and PORV open setpoints (psig) 45 45 110 Based on the above, it is concluded that the above change provides increased margin between the PORV and Code Safety Valve Setpoints. At the same time, the safety function of the code safety valves is not degraded.

Pursuant to the above, this is not an unreviewed safety question.

REACTOR COOLANT SYSTEM

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.'.' .M REACTOR COOLANT SYSTEM SAFETY VALVES AND ELECTROMhTIC RELIEF VALVE - OPERATING LIMITING CONDITION FOR OPERATION 4257GP'f 3' .

3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of N When not isolated, the pressurizer electromatic relief valve shall have a trip setpoint of 3, 2390 PSIG and an allowable value of 2, 23S5.5 PSIG.**

239o APPLICABILITY: 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEIt!.ANCE REQUIREMENTS 4.4.3 Tor the pressurizer code safety valves, there are no additional Surveillance Requirements other than those required by Specification

. 4.0.5. For the pressurizer electromatic relief valve a channel cali-bration check shall be performed every 18 months. -

  • The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
    • Allowable value for channel calibration check. .

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

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LIMITING SAFETY SYSTEM SETTINGS .,

5 BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal .

limits are either power peaking kw/ft limits or DNBR limits. The AXIAL -

POWER IMBALANCE reduces the power level trip produced by a flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.

RC Pressure - Low, High and Pressure Temperature The High and Low trips are provided to limit the pressure range in which reactor 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 -

setpoint is reached before the High Flux Trip Setpoint. The trip set-point for RC High Pressure, 2300 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,-2435-psig. The RC High Pressure trip also backs up the High Flux trip. \g

!, The RC Low Pressure,1985 psig, and RC Pressure-Temperature (12~.60 -

T *F-5660) psig, Trip Setpoints have been established to maintain the Dggtratio greater than or equal to 1.30 for those design accidents that

' result in a pressure reduction. It also prevents reactor operation at j pressures below the valid range of DNB correlation limits, protecting against DNB.

High Flux / Number of Reactor Coolant Pumps On In conjunction with the Flux - a Flux-Flow trip the High Flux / Number of Reactor Coolant Pumps On trip prevents the minimum core DNBR from decreasing below 1.30 by tripping the reactor due to the loss of reactor ,

coolantpump(s). The pump monitors also restrict the power level. for l

the number of pumps in operation.

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,JAVIS-BESSE, UNIT 1 B 2-6 Amendment No. 3 3 l . - - - . - . - - . - . _ - . .. ._ _

bewsas cp rag p/*CE FRi kl005L(

PROPOSED Bf LETTER REACTOR COCLANT SYSTEM Serial No. 6 (,9 Datej]/26/@

SAFETY VALVES - SHUTDOWN seE REVS 5ED 3 4 2 ATTAc44EDSEc.Tibc,ETQ LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2435 PSIG + 1%.*

APPLICABILITY: MODES 4 and 5.

ACTION:

i With no pressurizer code safety valve OPERABLE, immediately suspend all-operations involving positive reactivity changes and place an OPERABLE DHR loop into operation in the shutdown cooling mode.

i SURVEILLANCE REQUIREMENTS l

l 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

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

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DAVIS-BESSE, UNIT 1 3/4 4-3 o

. REACTOR COOUJT SYSTEM

.'. CrAMES.NW0i;5tt 'I SATETT VALVES - SHLTJOWN IEfus d L(U[R LIMITING CONDITION FOR OPERATION b'N' ** WI _ Date 12ldGlEO 3.4.2 Decay Heat Removal System relief valve DH-4849 shall be OPERABLE with a lift setting of f 330 PSIG* and isolation valves DH-11 and DH-12 open and control power to their valve operators removed.

APPLICABILITY: MODES 4 and 5.

ACTION:

A. With DH-4849 not OPERABLE: .

1. Make the valve OPERABLE within eight hours; or
2. a. Within next one bour, disable the capability of both high pressure injectio'n (HPI) pumps to inject water into the reactor coolant system; and
b. Vithin next eight bours:
1. Disable the automatic transfer of makeup pump suction to the borated water storage tank on low makeup tank level; and
2. Reducemakeuptankleveltof73inchesandreduce reactor coolant system pressure and pressurizer level within the acceptable region on Figstre 3.4.2-a (in MODE 4)and 3.4.2-b (in MODE 5).

B. With DE-11 or DH-12 closed, open DH-21 and DH-23 within one hour.'

C. With the control power not removed from DE-11 and DH-12, remove the power to the valve operators at the Motor Control Centers within ene hour. ,

SURVEILLANCE REQUIREMENTS 4.4.2. Decay Heat Removal System relief valve DH-4849 shall be deter-mined OPERABLE:

a. per the surveillance requirements of Specification 4.0.5.
b. at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying either:
1. isolation valves DH-11 and DH-12 open with control power removed from their valve operators; or
2. valves DH-21 and DE-23 open.

The lift setting pressure shall correspond to a=bient conditions of the valve at nominal operating te=perature and pressure.

DAVIS-BESSE, UNIT 1

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. Sena e GG9 dated tz/zs/go 3/4.4 REACTOR COOLANT SYSTEM

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BASES

The plant is designed to operate with both reactor coolant loops in operation, and maintain DNBR above 1.30 during all no'rmal operations and anticipated transients. With one'.eactor coolant pump not in operation in one loop, THERMAL POWER is restricted by the Nuclear Overpower Based on RCS Flow and AIIAL POWER IMBALANCE, ensuring that the DNBR will be maintained above 1.30 at the maximum possible THERMAL POWER for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR equal to 22%, whichever is more restrictive.

In MODES 3, 4 and 5, a single reactor coolant loop or DER loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requires two DER loops to be OPERABLE. ,

Natural circulation flow or the operation of one DHR pump provides adequate flow to ensure mixing, prevent stratification and produce -

gradual reacchity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction vill, therefore, be within the capacity of operator recognition and control. ,

3/4.4.2 and 3/4.4.3 SAFETT VALVES .

The pressurizer code safety valves operate to prevent the RCS from being 4 pressurized above its Safety Limit of 2750 psig. Each safety valve is designed to relieve 336,000 lbs. per hour of saturated steam at the valve's setpoint.

During operation, all pressuriaer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient. .

The relief capacity of the decay heat removal system relief valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that this relief valve is not OPERABLE, reactor coolant system pressure, pressurizer level and cake up water inventory is limited and the capability of the high pressure injection system to '

inject water into the reactor coolant system is disabled to ensure operation within reactor coolant system pressure - temperature limits.

Demonstration of the safety valves' lif t settings vill occur only during shutdown and will be performed in accordance with the prov'isions of Section X1 of the ASME Boiler and Pressure Code.

Ch.J:GES M i..:0'JiLY U

DAVIS-SESSE UNIT 1 B 3/4 4-1 Se M tb W Ca:e. @h< o-m - _ . _ _ .- - .- - ..

REACTOR COOLANT SYSTEM BASES For a RP5 high Me trip setpoint of 2300 psig, the maxim ove Soot of the Reactor Coolant System pressure for a loss of f d-Water 0FW) event would be 2350 psig. Also, the LOFW is the ximun 3.

over-pre ure anticipated transient. The string inaccuraci and drift for e RPS high pressure trip are 15.29 psi, or 16 si con-s ervatively. The maximum pressure peak for an anticip ed transient i s then 2366 p ' .

The inaccuraci and drift for the string t t controls the i

e lectromatic relief va e for the pressurizer inaccuracy e 16.75 psi, or 17 ps-conservatively. Include in this value is of 4 psi and a drift of 7.5 psi for the ansmitter. T e 4 psi and 7.5 psi were contined by taking the square oot of t sum of the squares, giving 8 ,5 psi. Subtracting 4 psi fro .5 i gives a value of 4.5 psi that i s attributable to only the drift. 1e 8.5 psi was then added to 1 1 accuracy and drift values for o e omponents in the string to ob-

! t tin a total of 16.75 psi. c The allosable value of,_2385.5 psig is tained by subtracting 4 .5 psi due to the drift om the trip setpoin of >2390 psig. The m inimum lift pressure f r the pressurizer elect Tic relief valve ip then (2400 the )maximum psig = 2373 psig. Consequ tly, the resul-

. tant margin betwee pressure peak of 2366 ig and mini-dum lift pressur of: 2373 psig for the pressurizer elect ~ matic relie f salve followin an anticipated transient is 7 psi.

Thus a 2300 psig RPS high pressure trip setpoint and the ove alues r the pressurizer electromatic relief valve will avoid a ua -

io f the pressurizer electromatic relief valve during anticipate s -

nsients.

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THL5 PAGE To - B E RE PL M(:D BY ATTAEHED GASES.

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DAVIS-BESSE, UNIT 1 B 3/4 4-la Amendment No. 3 3

BASES The pressurizer code safety valves must be set such that the peak Reactor Coolant System pressure does not exceed 110% of design system pressure (2500 psig) or, 2750 psig. The control rod group withdrawal accident will result in the most limiting high pressure in the RCS. The analysis assumes RPS high pressure trip at 2300 psig and the code safety valves open at 2500 psig. The tolerance on the RPS instrument accuracy is 30 psi and, it is +3% for the code safety valve settings. The pressurizer electromatic relief valve was assumed not to open for this transient. The resulting system peak pressure was calculated to be 2716 psig. Therefore, the code safety valve setpoint is conservatively set at 5 2525 psig which is the maximno pressure of 2500 psig + 1% for tolerance.

The pressurizer electromatic relief valve should be set such that it will open before the code safety valves are opened. However, it should not open on any anticipated transients. Loss of Feedwater (LOFW) was identified as the limiting anticipated transient for RCS pressure. The analysis assumes RPS high pressure trip at 2300 psig; with 30 psi for instrument errors, the resulting peak RCS pressure is calculated to be 2380 psig. This includes a 50 psig pressure overshoot on a LOFV _

transient.

. , o Docket No. 50-346 License No. NPF-3 .

Serial No. 862 October 14, 1982 Attachment 2 I. Changes to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications, Bases B3/4.7.1.2.

A. Time required to Implement This change is to be effective upon NRC approval.

B. Reason for Change (Facility Change Request 82-138)

To correct the discrepancy between the FSAR and the Technical Specification.

C. Safety Evaluation

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See Attached ,

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SAFETY EVALUATION This revision to the Technical Specification Bases Section 3/4.7.1.2, page B3/4.7-2 relates to the pumping capacity of Auxiliary Feedwater Pumps.

The following safety evaluation constitutes the review of the change pursuant to the 10 CFR 50.59.

The safety function of the Auxiliary Feedwater System is to provide feedwater to the steam generators for the removal of reacto* decay heat in the absence of main feedwater and to promote natural circulation of the Reactor Coolant System in the event of a loss of all four reactor coolant pumps.

Per the existing Technical Specification Bases Section B3/4.7.1.2, the j required capability of the auxiliary feedwater pumps to deliver feedwater flow to remove decay heat and reduce the reactor coolant system temper-ature to less than 280*F is given to be 850 gpm at 1035 psig. As per FSAR Vol. 9, Section 15, Page 15.1.7-1 and per B&W analysis submitted to the NRC in Toledo Edison Serial No. 717 dated May.22, 1981, a flow rate of 800 gpm is determined to be acceptable to meet the required safety function of the Auxiliary Feedwater System. Therefore, the numbers in the bases need to be revised accordingly. Since a capacity of 800 gpm at 1050 psig has been demonstrated to be acceptable in the FSAR and Toledo Edison submittal of May 22, 1981 this change does not degrade the safety function of the Auxiliary Feedwater System. Therefore this is not an unreviewed safety question.

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PLANT SYSTEMS 2

BASES 3/4.7.1.2 AUXILI ARY FEE 0'a'ATER SYSTEMS The OPEPABILITY of the auxiliary feedwater kystems ensures that the Reactor Coolant System can be cooleo down to less than 280*F from normal operating conditions in the event of a total loss of offsite 600 1050 Each steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow o a gpm at a pressure of %34 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce

, the Reactor Coolant System temperature to less than 280*F where the Decay Heat Removal System may be placed into operation.

3/4.7.1.3 CONDENSATE STORAGE FACILITIES

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The OPEPABILITY of the condensate storage tank with the and to minimum water volume ensures that sufficient water is available .to maintain the RCS at HOT STANDBY conditions for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> with steam discharge to atmosphere and to cooldown the Reactor Coolant System to less than 280'F in the event of a total loss of offsite power or of the main feedwater system. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

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,, 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that l the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures

, that no more than one steam generator will blowdown in the event of a b steam line rupture. This restriction is required to 1) minimize the 1

a DAVIS-BESSE, UNIT 1 B 3/4 7-2

Docket No. 50-346 License No. NPF-3 Serial No. 862 October 14, 1982 Attachment 3 I. Changes to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications Table 3.3-4.

A. Time required to Implement This change is to be effective upon NRC approval.

B. Reason for Change (Facility Change Request 82-116)

To account for instrument error in the trip setpoint for the BWST level and Essential Bus Feeder Breaker Trip.

C. Safety Evaluation

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See Attached 9

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SAFETY EVALUATION This amendment request proposes that the lower end of the Technical Specification Borated Water Storage Tank (BWST) low level interlock trip setpoint be changed from 91.5 inches to 89.5 inches with a corresponding change in the Technical Specification allowable value. The request also proposes a change to the Essential Bus Feeder Breaker C1 (D1) trip set- .

point from 711.5 seconds to <7.8 seconds in the Technical Specifications.

The safety function of~the BWST low level interlock is to provide a

permissive output to enable the manual transfer of decay heat (DH) and containment spray (CS) pump suction from the BWST to the containment emergency sump on low level in the BWST. This trip is also used to prevent' a premature manual transfer and to protect these pumps from cavitation for lack of proper net positive suction head. The safety function of the C1 (D1) essential bus feeder breaker trip setpoint is to ,

provide a time delay setpoint of the undervoltage relays (set at 90%) on j the 4.16 KV safety related buses which actuate the diesel generators in the event of loss of offsite power. It also ensures acceptable volt.'ges

on the safety related 480V buses for continuous and emergency operations.

Change In BWST Low Level Setpoint At present, the suctions to the DH and CS pumps are manually transferred

- from the BWST to the Containment Emergency Sump on a Safety Features Actuation System (SFAS) trip of incident level 5 (BWST low level). The lower end of the Technical Specifications for the BWST low level interlock trip setpoint is 91.5 inches. However, the existing setpoint range for the BWST low level interlock allowed by the Technical Specifications does

. not permit adequate margin for errors inherent in the instrumentation.

The following table shows the adequacy of the BWST levels required for this suction transfer to be successfully performed as a manual transfer in light of the proposed setpoint change.

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0 Table 1 BWST Level BWST Volume Description (Inches) (Gallons)

Develop Minimum Level to Transfer Suction to Containment Emergency Sump

1. Accident Analysis minimum level to start 36 the transfer per the original analysis
2. Instrument string inaccuracy and drift 13.5
3. Lowest safe indicated reading to start 49.5 control room operator action to transfer Develop Minimum Contained Volume
1. Highect (lowest) indicated reading to 100.5 (89.5)

Interlock trip can occur (Technical Specification trip limits)

2. Instrument string inaccuracy and drift 13.5 .
3. Highest (lowest) a tual level that 114 (76)

Interlock trip can occur (this volume may not be available for the decay heat or containment spray pump)

4. Instrument String drift 1.2
5. Highest (lowest) allowable 101.7 (88.3) interlock trip
5. 360,000 gallons required to be added 334.3 360,000 for Emergency Core Cooling System (ECCS) analysis
7. Lowest safe indicated level for ECCS 448.3 482,778 Analysis in Modes 1, 2, 3, & 4 (minimum volume presently required by Tech-nical Spec-ification)

.c As shown in the above table, a control room operator will manually perform this transfer when the safety grade level indicators in the control room indicate between 49.5 and 100(89.5) inches of BWST level. This will give a control room operator about 4 minutes to make the transfer safely. At maximum flow from BWST this will occur at approximately 23 minutes.

During this time, the control room operator is procedurally instructed to monitor the B%ST level indicators to initiate timely manual transfer. As observed from the above calculation table, the minimum level requirements of BWST will be met if the transfer is initiated within about 4 minutes after the indicated BWST level drops to the interlock setpoint. This assumes the worst case condition of a BWST level draw-down of 11.3 inches per minute.

A control room operator will manually perform this transfer about 23 minutes after the initial SFAS trip that started all high pressure injec-tion, low pressure injection and containment spray pumps assuming their maximum flow. The accident analysis requires 360,000 gallons to be added for ECCS analysis when in Modes 1, 2, 3, & 4.

Chante In Undervoltage Relay Time Delay Setpoint Each safety related 4.16 KV bus C1 (D1) contains four relays which sense-bus voltage and initiate tripping of the bus (C1 or D1) feeder breakers on

, undervoltage. The 90% setting is the minimum permissible operating voltage of the safety related 4.16 KV buses that ensures acceptable voltages on the safety related 480V buses for continuous and emergency operation. .The maximum allowable operating delay for there "90% voltage"

. relays is 9 seconds as assumed in.the accident analysis. In determining the Technical Specification trip setpoint for the time delay, the maximum error and drift inherent in the relay must be subtracted from the 9

. seconds to er sure that the accident analysis value is not exceeded. The tolerance in the time delay is 110% (of setpoint) with an additional 15%

I (of setpoint) for drift. Therefore, the maximum error including drift would be 15% of setpoint. Therefore, the setpoint should be <7.8 seconds in the Technical Specification as shown in Table 2 below. By making these changes the accident analysis value of 9 seconds will not be exceeded and l

the safety function of the protective relay will be enhanced.

Table 2 l Sescription Trip Setpoint

1. Accident Analysis Value 9 sec.
2. Technical Specification Value <7.8 sec.
3. Maximum Error including drift (15% of setting) 0.15x7.8 = 1.17 These changes in the Technical Specifications will provide the same safety function as is provided by the present setpoint as discussed above. No adverse environment will be created by the change and t5e safety functions noted earlier will not be affected. Pursuant to the above it is concluded that no unreviewed safety question is involved.

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l ". SAFETY TEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 2

P fullCTIONAL UtilT -

C TRIP SETP0lHT ALLOWABLE VALUES lil5TRUMEllT STRINGS

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a. Containment Radia tion .

<. 2 x Background a t RATED IIIERMAL POWER

< 2 x Background at, RATED THERMAt. PowrR

b. ~ontainment Pressure - liigh < 18.4 psia < 18.52 psla#
c. Containment Pressure - liigh-liigh < 3fl.4 psia <.38.52 psla#
d. RCS Pressure - Low > 1620.75 psig > 1615.75 psig#

5 c. RCS Pressure ' Low-Low

.., > 420.75 psig > 415.75 pstg' w

f. DWST Level 29 3

> -9h5- a n d~< 100. 5 i n . H 2O

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> 40r-3. a nd 39,5 < 101.7 in, li p 0 C SEQUEtiCE LOGIC CilANNELS

a. Essential Bus Feeder Breaker Trip (90%)

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> 3744 vol ts for > 3558 volts h __7 i ' .5 nc _ for -7 i '.5 3rc j

R. b. Olesel Generator Start, load Shed on < -). g .s f c.

2 Essential Bus (597,)

< 7 7 sEC.

% > 2071 and < 2450 volts > 2071 and < 2450

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for R.510.1 sec volts for

. 0.5 1 0.1 see, ItiTERLOCK C11ANNELS g a. Occay llent isolation Valve

< 438 psig

" and Pressurizer llea ter < 443 psig1* -

cn

  • Allowable Yalue for CllANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION.

Referenced to the centerline of DH11 and 01112

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