Text

Amend 60 to License DPR-34,revising Certain Setpoints for Protective Sys to Allow for Instrumentation Error
	ML20151A992
Person / Time
Site:	Fort Saint Vrain
Issue date:	06/20/1988
From:	Calvo J; Office of Nuclear Reactor Regulation
To:	;
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ML20151A991	List: ML20151A988; ML20151A992; ML20151A996; ML20151B003;
References
	NUDOCS 8807200191
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'o UNITED STATES

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.j NUCLEAR REGULATORY COMMISSION WASHINGTON, D, C 20555

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PMLIC SERVICE COMPANY OF COLORAC0 DOCKET NO. 50-267 FORT ST. VRAIN NUCLEAR GENEPATING STATION AME':0 MENT TO FACILITY OPERATING LICENSE Amendment No. 60 License No OPR-3a

1. The Nuclear 7erulatory Commission (the Commission) has found that:

A. The application for amendment by Public Service Company of Colorade (the licensee) dated February 8, 1988, complies with the standarcs and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in ccnfermity with the application, as amended, the provisions of the Act, and the rules and regulaticos of the Commission; C. There is reasonable assurance: (i) that the activities authori:ed by this amendnent can be conducted without endangering the health and safety of the public, and (ii) that such at.tivities will be ccnducted in cor.pliance with the Comission's regulations; D. The issuance of this license amendment will not be inimical to the corron 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 Comission's regulations and all applicable requirements have been satisfied.

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i 8807200191 880620 1 PDR ADOCK 05000267 P PDC ,

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2. Accordingly, the license is amended by changes to the Technical l Specifications as indicated in the attachment to this license amendment, !

and paragraph 2.D.(2) of Facility Operating License No. OPR-34 is hereby !

amended to read as follows: l

. J (2) Technical Specifications The Technical Specifications qqntained in Appendices A and B, as revised through Amendment No.ou , are hereby incorporated in the license. The licensee 4 hall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective 90 days after its date of issuance.

FOR THE NOCLEAR REGULATORY COMMISSION CI . (QAw Jose A. Calvo, Director Project Directorate - IV Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical

Specifications Date of Issuance: June 20,1988 l

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ATTACHMENT TO LICENSE AMEN 0 MENT NO. 60 TO FACI,LITY OPERATING LICENSE NO. OPR-34 DOCKET NO 50-267 Replace the following pages of the Appendix A Technical Specifications with the attached pages as indicated. The revised pages are identified by amendr.ent number and contain vertical lines indicating the areas of change.

Remove Insert 3.3-1 . 3.3-1 3.3-2 3.3-2a 3.3-2b 3.3-2c 3.3-3 3.3-3a 3.3-3b 3.3 d 3.3-4 3.3-5 3.3-5 3.3-6 3.3-6 3.3-7 3.3-7 3.3-8 3.3-8 4.4-1 4.4-1 4.4-2 4.4-2 4.4-3 4.4-3a 4.4-1b 4.4-3c 4.4-4 4.4-4a 4.4-5 4.4-4b 4.4-4c 4.4-4d 4.4-6 4.4-Sa 4.4-5b ,

4.4-5c 4.4-7 4.4-7a 4.4-7b 4.4-8 4.4-8 4.4-10 4.4-10 4.4-11 4.4-10a 4.4-12 4.4-10b 4.4-10e 4.4-12a 4.4-11 4.4-11a 4.4-12b 4.4-12 4.4-12a 4.4-12b 4.4-12c 4.4-13 4.4-13

Fort St. Vrain si l Technical Sp l Amendment # g fications Page 3.3-1 3.3 LIMITING SAFETY SYSTEM SETTINGS i Applicability Applies to the trio settings for instr 6 ents and covtces I which provids for monitoring of reactor power, not reheat temperature, reactor internal pressure, and moisturd content of the helium coolant.

Objective To provide for autoestic protective action such that the .

principal process varf,4bles do not exceed a safety limit as a result of transients.

Specifica' ton LS$$ 3.3 - Limitina Safety System Settings The Limiting Safety System Settings for trio shall be as specified in Table 3.3.1. The following definitions are ustd in the table: l Trio Setooint - The trip setpoint is the least conservative "as lef t" value for a channel to be considered Operable.

1 Allowable Value -

The allowable value is the least '

conservative Operable.

"as found" value for a channel to be considered l

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Fort 55.Vraindi

' Technical Sp knendment e gifit:ations Page 3.3-2a

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$pectrication Lss$ 3.3 Table 3.3-1 '

t.!MITING SAFETY $Y$ TEM SETTINGS TRIP ALLOWABLE PARAMETra FUNCT t0N SETPOINT VAWE '

1. Reactor Core Limiting Safety System Settings

4) Linear Scram Varies as a Varies as a Channel-High Function of Fwnction of (Neutron Indicated Indicated Flus) Thermal Thermal Power per Pcwor per Figure 3.3-1 Figure 3.3-1 b) Reheat Scram < 1055 < 1067 Steam Begree F segree F Temperature-High c) Primary $ cram 1 68.6 pst 1 72.7 psi Coolant below normal, below normal, Pressure- prog rasped programmed Programmed with Circu*

Low with Circu-14 tor Inlet lator Inlet Temperature. Temperature Upper TRIP per Figure SETPOINT of 3.3 2. Upcor 631.1 psia. limit to produce trip at 627 psia.

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Fort St. Vrain el Technical Sp cifications Amendment , O Page 3.3-2b Specification LS$$ 3.3 Table 3.3-1 (Continued)

LIMIT!NG SAFETY SYSTEM SETTINGS TRIP ALLCVABLE PARAMETER -FUNCTICN SETootNT %ALUE

2. Reactor Vessel '

Pressure Limiting Safety System Settings

4) Primary Scram and i 46 pst i 52.7 pst Coolant Preselected above normal, above normal, Pressure- Loop Shutdown programmed programed Programed and Steam / with Circu- with Circu-Hign water Oump lator Inlet lator Inlet Temperature. Temperature Upper TRIP per Figure SETPOINT of 3.3-2. Upper

< 746.3 psia. limit to

{cwor TRIP ' produce trip SETPOINT of at < 753 i 538.3 psia, esti. Lower limit to produce trip at i 545 psia b) Primary Scram, Loop Coo' ant i 60.5 1 62.2 Shutdown, degree F degree F Moisture- and Steam / dewpoint dewpoint High Vater Cump toeperature toeperature c) PCRV Pressure Pressure: Relief Rupture Ofsc 812 esig plus 820 psig (Low Set or minus 4 Safety Valve) psi 1

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Forg 5%, Vrain 01 Technical Specifications Amendmen2 e60 e

Page 3.3-2c Specification LS$$ 3.3 Table 3,3-1 (Continued)

LIMITING $AFETY SYSTE.4 SETTINGS TRIP ALLOWABLE DaRMETER NNCTICN SETp0!NT VAUJE Low Set Safety Valve 796 estg plus 804 psig '

or minus 8 pst '

Rupture Otse (High Set Safety 832 estg plus 840 pstg or minus 8 pst Valve)

High $tt Safety Valve 812 p519 plus 820 pstg or minus 8 est d) Helium Pressure Circulator Reitef penetration Interspace Pressure:

Ruoture Disc (2 Per 825 esig plus 842 psig or minus 17 Penetration) pst Safety Valve (2 Per 805 esig plus 829 pstg or minus 24 Penetration) psi e) Steam Pressure Generator Relief Penetration Interspace Pressure:

Rupture Disc (2 For Each 825 psig plus 842 estg or minus 17 Steam Generator) est safety Valve (2 For Each 475 estg plus 489 pstg or minus 14 Steam Generator) pst l

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Greater than 70 to 80 1 95 $ 96 Greater than 80 to 90 1105 $ 106 i

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HIGH NEUTRON FLUX SCRAM DETECTOR DECALIBRATION i CIIRVFR FOR CYCL F 4 ,

Fort Stc Vrain #1 Technical Specifications .

Amendment # 60 l

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700 soo CIRCULATOR INLET TEMPERATURE ('F)

FIGURE 3.3 2 PRIMARY COOLANT PRESSURE vs. CIRCUt# TOR INLET TEMPERATURE ALLOWABLE OPERATION l

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, Fort St. Vrain #1 Technical . Specifications ,

Amendment # 60 !

.' Page 3.3-4 Basis for $cecification LS$$ 3.3 Safety Limits have been established in Specification $L 3.1 and SL 3.2 to safeguard the fuel particle integrity and the l reactor pristry coolant system barriers, protective devices have been provided in the plant design to ensure that autoestic corrective action is taken when required to prevent the Safety Limits from being etceeded during normal operation or during operational transients resulting from possible operator e rro rs , or as a result of equipment malfunction. This specification establishes the Trip

$etpoints and Allowable Values for these automatic protective devices.

Operation with se,tpoints less conservative than the Trip i Setpoint but within the Allowable Value is acceptable since '

an allowance has been sade in the safety analysis to ,

accommodate this error, as described below, i General Methodolecy The Analysis Value is the value of a parameter for which a Trip and initiation of automatic protective action is

?.ssumed to occur in FSV accident analyses (F$AR Chapter 14). ,

Provided that the trip cccurs at a value equal to or more i conservative than the Analysis Value, analyses demonstrate 1 that consequences of the accident or transient are !

acceptable.

!$A Standard, $67.04-1982 has bean applied to these Analysis values to arrive at Allowable Values and Trip Setpoints for each PPS parameter, l

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, Fort St. Vrain Al Technical Specifications Amendment J 60

,- Page 3,3-5 Basis for Specification L$$13.3 (Contiaued)

Linear Channel - High (Neutron Flus)

The neutron flux Trip 5etpoints are established to protect the fuel particle integrity during rapid overpower transients. The power range nuclear channels respond to ,

changes in neutron fluz. During normal power operation, the channels are calibrated using a plant hee.t balance so that the neutron flux that is sensed is indicated as percent of Rated Thermal Powe r. For slow maneuvers, those where core thermal power, surface heat flux, and the heat transferred to the helium follow the neutron flux, the power range nuclear channels.will indicate reactor Thermal Power. For fast transients, the neutron flux change will lead the change in heat transferred from the core to the helium due ,

to the effect of the fuel, moderator and reflector thermal time constants. Therefore, when the neutron flux increases to the scram Trip $stpoint rapidly, the percent increase in heat flux and heat transferred to the helium will be less than the percent increase in neutron flux. Trip Setpoints that ensure a reactor scram at no greater than 140% Rated The rmal Power are sufficient for the plant because the negative temperature coefficient of reactivity and large heat capacity of the reactor limit the transient increases in fuel and helium troperatures to acceptable values.

Control rod shim bank movement can result in decalibration of the external-core neutron flux detectors. To ecteunt for this potential decalibration and other instrumentation errors, the actual Trto Setpoint is administratively set less shan 140% Rated Thermal Power based upon indicated powe r. These administratively set flux Trip $etpoints ensure the scram will occur at or less than 140% Rated Thersal Power for those postulated reactivity accidents evaluated in FSAR Section 14.2. Additional discussion on detector decalibration is given in updated FSAR 5ection j 7.3.1.2.1. -

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Fort St. Vrain #1 Technical Specifications Amendment # 60

,- Page 3,3-6 Basis for Specification 1.555 3.3 (Continued)

Reheat staas Teeperature - Hich High reheat steam tesperature indicates either an increase in Thermal Power generation without an appropriate increase in helium cooling flow rate or a decrease in steam flow rate. (Reheat steam temperature in lieu of reactor core outlet helium temperature is used because of the difficulty in meast'ing gross helium temperature for protective system ,

purposes.) The design of the steam generator is such that changes in hot helium temperature due to a power increase '

first affect the reheat steam temperature, thus allowing the latter to serve as an index of the heltum temperature. A reheat steam teope rature scram is provided to prevent excessive Power-to-Flow-Ratio due to a power increase or steam flow imbalance. (FSAR Section 14.2) ,

primary Ceolant pressure - proa amed low The low primary coolan; pressure Trip Setpoint has been estabitshed to maintain the fuel particle coating integrity due to loss of primary coolant as a result of a coolant leak.

Primary Coolant Pressure - progra wed Mich The sajor potential source of primary coolant pressure increase above the nomal operating range is due to water and/or steam inleakage by means of a defective evaporator-economizer-superheater subheader or tube. For a double-ended offset tube rupture, the rate of water and steam inleakage will not exceed 35 lbs/sec initially. esulting in a maximum rate of primary coolant pressure increase of approximately 1 pst oer second. The normal PPS action upon detection of moisture is reactor scram, loop shutdown, and steam / water dump (F5AR Section 7.1.2.5), occurring after approximately 12 seconds, assuming rated power and flow conditions, in this situation, the peak PCRV pressure at 100% reactor power does not exceed 705 psia. The Trip Setpoint of less than or equal to 46 psi above the normal operating pressure between 25% and 1005 rated power is selected: (1) to prevent falso scrans due to normal plant i transients, and (2) to allow adequate time for the normal l protective action (high moisture) to terminate the accident l while limiting the resulting peak PCRV pressure in the l unlikely event that the nomal protective action was inoperative. In this case, Reactor Pressure would continue to rise to the high pressure Trip Setpoint. The resulting peak PCRV pressure would be less than the PCRV Reference Pressure. The high pressure Trip Setpoint is programmed as a function of load, using helium circulater inlet temperature as the measured variable indicative of load, as shown in Figure 3.3-2. The PCRV safety valves provide the i ultimate protection against primary coolant systes pressure i exceeding the PCRV Reference Pressure of 845 psig.

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Technical Sp stfica% ions Amendnent#gU i

Page 3.3-7 ,

Basis for Soecification L55$ 3.3 (Continued)

Primary Coolant Moisture - High The high moisture Trip Setpoint corresponding to 60.5 degrees F dewpoint was established, considering the moisture monitor ciaracteristics and the necessity to minim 12e water inleakage to the primary coolant system. A Trip would be reached after several hours of full power operation with a minimum water / steam inleakage rate in excess of about 20 lbs/hr. Below that inleakage rate, the Trip Setpoint would never be reached, but the indicating instruments would show an abnormal condition. For maxis.um design leakage rates, the system behavior is as discussed in the preceding section on Primary Coolant Pres sure-Progransned High. Backup protective action is provided by the high primary coolant pressure scram, loop shutdown, and dump of a pre selected loop and remaining loop steam depressurization. (FSAR Sections 7.1.2.3' and 7.1.2.4.)

pCRV Pressure The PCRV safety valvas provide the ultimate protection against primary coolant system pressure exceeding the PCRV Reference Pressure of 845 psig. This engineered safeguard system consists of the isolation valves, the rupture discs.

the relief valves, and the containment tank. Two safety valves are provided, either of which is adequate to prevent exceeding the PCRV Reference Pressure in the event of a steam generator subheader rupture, which is the only credible means of substantially increasing the prima ry coolant pressure. If the pressure in the PCRV were to rise significantly above the Nonnel Working Pressure, the low-set rupture disc would rupture within the range of 804 psig

(-15), to820psig(+1%). The low set safety valve, set at 796 psig plus or minus 15, would be wide open and relieving at full capacity at or above 820 psig (31 accumulation). If the pressure still continued to rise, the high-set rupture disc would rupture between 824 psig and 840 psig. The high-set safety valve, set at 812 psig plus or minus 15, would be relieving at full capacity above 836 psig (3! ;ccumulation).

As the pressure decreased, the high-set safety valve would close at a pressure of approximately 690 psig and the low-set safety valve at approximately 677 psign the corresponding primary system pressure would be approximately 737 psig when the low set safety valve closed. The minimum ,

permissible trip setpoint of each PCRV overpressure re;lef j train rupture disc and relief valve is specified to provide '

assurance that primary coolant helium will not be vented to l atmosphere during primary roclant pressure surges, resulting ;

from transients or accitents, in which pressurts do not I approech the Allowable Value and thereby do not challenge '

the integrity of the PCRY. (FSAA Section 6.8.3) ;

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Techn9 Cal'$pecifications )

Amendment'#60 l Page 3.3-8 !

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Basis for Specification LSS$ 3.3 (Continued)

Helium Circulator Penetration Interspace Pressure !

The penetration intarspaces are protected against pressures I exceeding PCRV Reference Pressure (845 psig). The safety valves are set at 805 psig and rupture discs are set at 825 psig (nominal). A redundant safety valve and rupture disc I are provided. The rupture discs would burst in the pressure j range of 808 psig (-21) to 842 psig (+21). The safety !

valves would open in the range of 781 psig (-31) to 829 psig

(+3%) and would relieve at full capacity at 886 psig (los accumulation). The safety valves would reseat at about 725 '

psig. The safety valve and rupture disc relieving pressures were specified "so as to comply with the ASME Boiler and Pressure Vessel Code. Section !!!. Class B. Nuclear Vessels, for overpressure protection. The minimum pennissible trip '

setpoint of each rupture disc and associated relief valve is specified to provide assurance that PCRV penetration interspace helium, which could potentially be radioactive, will not be vented to atmosphere during interspace pressure surges in which pressures do not approach the Allowable Value and thereby do not challenge the integrity of the PCRV penetration. (FSAR Section 5.8.2)

Steam Generator Penetration interspace Pressure The six steam generator penetration interspaces in each loop are provided with cossen upstream rupture discs and safety valves to protect against pressures exceeding PCRV Reference Pressure (845 psig). A redundant safety valve and rupture disc are provided. The rupture discs would burst in the pressure range of 808 psig (-21) to 842 psig (+25). with a nominal setting of 825 psig. The safety valves are each set at 475 psig which allows for a pressure drop in the inlet lines of 370 pst when relieving at valve' capacity. The minimum permissible trip setpoint of each rupture disc and associated relief valve is specified to provide assurance that PCRV penetration interspace helium, which could potentially be radioactive, will not be vented to atmosphere during interspace pressure surges in which pressures do not approach the Allowable Value and thereby Jo not challenge the integrity of the PCRV penetration. (FSAR Section 5.8.2) l l

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For% St. Vrain 01 Technical Specifications

/Vnendment e .43,60 Page 4.4-1 4.4

!N$TRUMENTATICN ICR Of 6RATICN AND CCNTRCL SYSTEWS - LIMITING CCN0!T!CNS Applicabitity Applies to the plant protective system and other critical instrumentatio3 and controls.

Objective To assure and other criticalthernstrumentation operability of the plant orotective system by cefining the minimum operable instrument channels and trip settings. '

specification LC0 4.4.1 - plant protective system Instrumentation, Limittna Conditions for Operation The limiting conditions for the plant protective system instrumentation are shown on Tables 4.4=1 through 4.4*4 These tables utiltre the following definitions:

Georee of Redundancy - Dif ference between the nuncer of operable channels and tne minimum number of operable channels which when tripped will cause an automatic systes trip.

Operable Channel - A channel is operable if it is capable of fulftlling its cesign functions.

Inoperable Channel - Opposite of operable channel.

Trip setpoint -

The trip setacint is the least conservative Operable. "as left" value for a channel to be considered Allowable Value -

The allowable value is the least conservative Operable. "as founc* value for a channel to be consicered manner:TablesIf the 4.4-1through4.4-4arekobereadinthefollowing minimum operable channels or the minimum degree of redundancy for each functional unit of a table cannot be met or cannot be bypassed under the stated permi ssible bypass conditions, the following action shall be tateh:

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Fort St. Vrain el Technical Soecifications tenendment . eA3,60

. Page 4.4-2 For Table 4.4*1, the reactor shall be shut down within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except that to facilitate maintenance on see plant Protective System (PPS) moisture sonitors, sne solsture sonttor input trip functions to the Plant Protective Systee vnich cause scram, loop shutdown, circulator tris, and steam water dumo say be disabled far up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . During the time that tne Plant Protective Systee moisture sontter trips are disabled, an observer in direct communication with the reactor operator shall be positioned in the control room in the location o' pertinent instrumentation. The observer shall continuously montAor the primary coolant moisture levels indicated by at least two soisture monitors and the primary coolant pressure indications, and shall alert the ,

reactor operator to any indicated moisture or pressure change. During the time in which the trip functions are disaDied the reautrements of LCO's 4.2.10 and 4.2.11 shall be met and primary coolant shall not exceed a noisture concentration of 100 ppev.

For Table 4.4-2, the affected loop shall be shut down within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

For Table 4.4-3, perform one of the following within 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

a. The reactor shall be shutdown, or b.

the affected helium circulator shall be shutdown.

For Table 4.4-4, the reactor small be shut down within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If, within the indicated time limit, the minimum numcer of operable channels and the minimum degree of reduncancy can be reestablished, the system is considered normal and no further action needs to be taken.

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Fort St. Vrain #1 -

Technical Specifications Amendment i13,60

,- Page 4.4-3a specification LCO 4.A.1 ,

Table 4.4 1 (Part 1)

!NSTRLt'ENT OPERAT!NG REQUIRENENTS FOR PLANT PROTECT!vt SYSTEM SCRAM TRIP ALLOWA8LE NO. FUNCT!0NAL UNIT SETPO!NT VALUE ,_

14. Manual Scram Not Applicable Not Applicable (Control Room)

Ib. Manual Scram Not Applicable Not Applicable *

(Outside Control Room)

2. Startup Channel-High $8.3E+04 cps 19.3E+04 cps Count Rate 3a. Linear Channel High -----------See Table 3.3 ---

Channels 3,4,5 (Neutron Flux) 3b. Linear Channel-High -----------See Table 3.3-1-------

Channels 6,7,8 (Neutron Flux) 4 Primary Coolant Moisture High Level Monitor < 60.5 degree F < 62.2 degree F Iewpoint Iowpoint Loop Monitor < 20.4 degree F ( 22.1 degree F 3ewpoint Bewpoint

5. Reheat Steam Temperature $ 1055 degree F i1067degreeF

-High Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4 8 and 4.4 9 9

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Fort St. Vrain #1 Technical Specifications Amendment # 60

. Page 4,4-3b l Specification LCO 4.4.1 Table 4.4-1 (Part 1)

INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEQCRAM ,

TRIP ALLOWABLE ,

NO. FUNCTIONAL UNIT $ETPOINT VALUE +

6. Primary Coolant Pressure -------- $ee Table 3.3 2 - - ---

Progra med Low -

7 C+b ary Coolant Pressure - ------ See Table 3.3 2- -------

yrassned High

8. N Aeheat Header Pressure 144psig 143psig

.uw

9. Main Steam Pressure Low 11529psig 11517psig

10. Plant Electrical System-Loss > 278V > 266V 7 31.5 Seconds 335 Seconds ,

11. Two Loop Trouble t 'pplicable Not Applicable

12. High Reactor Building < 161 degree F

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< 166 degree F

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Temperature (Pipe Cavity)

Notes for bles 4.4 1 through 4.4-4 are on Pages 4.4 8 and 4.4-9 l

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. Fort St. Vrain #1 Technical Specifications Amendment # 60 i

? Page 4,4-3c l SPECIFICATION LCO 4.4.1 l TABLE 4.4-1 (Part 2) l INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECT!vE SYSTEM. SCRAM MINIMUM MINIMUM PERM!$51BLE OPERABLE DEGREE OF BYPASS NO. FUNCTIONAL UNIT CHANNELS R EDUf:DANCY CON 0!TIONS la. Manual (Control Room) 1 0 None Ib. Manual (Outside Control 2 (f) 1 None ,

Room) ,

2. Startup Channel-High 2 1 Reactor Mode '

Count Rate Switch in "RUN" 3a. Linear Channel-High, 2 (f) 1 N:ne Channels 3, 4, 5 3b. Linear Channel High, '

2 (f) 1 None Channels 6, 7, 8 4 Primary Coolant Moisture High level Monitor 1(f,t) 1(c) (h2)

Loop Monitor 2/ Loop (f,t) 1/ Loop (hl)

5. Reheat 3 ten 2(b.f) 1 None Temperature - High

6. Primary Coolant 2 (f k) 1 Less Than 30%

Pressure - -Rated Power Progransned Low

7. Primary Coolant 2 (f.k) 1 None Pressure -

Progransned High

8. Hot Reheat Header 2(f) 1 Less Than 30t Pressure - Low Rated Power

9. Main Smu 2(f) 1 Less Than 30%

Pressure - Low Rated Power

10. Plant Electrical 2(e,f) 1 None System - Loss

11. Two Loop Troubtri 2 1 Reactor Mode ,

Switch in :

"Fuel loading" !

12. High Reactor Building 2 (f) 1 None I Temperature (Pipe Cavity)

Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 l

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Fort 5%. Vrain #1 Technical Sp Amendment # gifications Page 4.4-4a Specification LCO 4.a.1 Table 4.4-2 (Part 1)

INSTRUMENT OPERATING REQUIREMENTS FOR THE PLANT PROTECTIVE SYSTEM. LOOP SHUTDCVN 4

TRIP ALLCWABLE MO. FUNCT!cNAL UNIT S ETPOINT VALUE la. Deleted Ib. Deleted Ic. Deleted Id. Celeted le. Deleted If. Deleted 2a. Deleted 2b. Deleted 1

2c. Deleted l 2d. Deleted l Ja. Loop 1 Shutdown Logic Not Applicable Not Applicable 3b. Loop 2 Shutdown Logic Not Applicable Not Applicable Ja. Circulator 1A and 19 Not Applicable Not Applicable Shutdown - Loop Shutdown Logic Notel for Iable$ 4.4*1 through 4.4-4 are on Pages 4.4-8 and 4.4-9

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Fort St. Vrain #1 Technice.1 Specifications Amendment # 60

.- Page 4.4-4b i

. l l

Specification LCO 4.4.1 Table 4.4-2 (Part 1)

INSTRUMENT OPERATING REQUIREMENTS FOR THE PLANT PROTECTIVE SYSTEM. LOOP SHUTDCW TRIP ALLOWA8LE NO. FUNCTIONAL UNIT SETPOINT VALUE ,

4b. Circulator 1C and 10 Not Applicable Not Applicable Shutdown - Loop Shutdown Logic Sa. Steam Generator 1 796 osig 1 801 psig Penetration Overpressure, Lcop 1 Sb. Steam 'tnerator 1 796 psig i 801 psig Penetrasion Overpressure, Loop 2

64. High Reheat Header < 3.2 mrem /hr < 3.5 mres/hr Activity, Loop 1 Ibove Ibove 8ackground Background 6b. High Reheat Header < 3.2 aree/hr < 3.5 arem/hr Activity, Loop 2 Ibove Ibove Background Background

74. Low Superheat Header 1 798 degree F t 794 degree F Temperature, Loop 1 (p) 7b. Low Superheat Header 1 798 degree F 3 794 degree F Temperature Loop 2 (p) 7c. High Offferential 1 44.8 degree F 1 46.7' degree F Temperature Between Loop 1 and Loop 2 (D) .

i l

Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 1

4 1

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Fort'St. Vrain #1 Technical Specifications Amendment # 60 i

, Page 4.4c4c l

1 SPECIFICATICN LCO 4.4.1 TABLE 4.4-2 (Part 2)

INSTRUMENT CPERATING RE0V!REWENTS FOR PLANT PROTECTIVE SYSTEM, LCCP SHUTCC%N MINIMUM MINIMUM PERMISSISLE OPERABLE CEGREE OF BYPASS NO. FUNCTIONAL UNIT CHANNELS REQUNDANCY CON 0!TICNS la. Deleted .

Ib. Deleted '

Ic. De eted 1d. Deltted le. Deleted If. Deleted 2a. Deleted 2b. Deleted 2c. Deleted 2d. Deleted 3a. Loop 1 Shutdown 2 1 None Logic 3b. Loop 2 Shutdown 2 1 None Logic Notes fJr Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4 9 l

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Fort 52. Vrain *1 ;

Technical Speci fications Amendment 0 60 Page 4.4-4d l

SPECIFICAftCN t.C0 4.4.1

_T.A8LS 4. 4-2 (pa rt 2)

INSTRUMENT OPERATING REQUIRENENTS FOR PLANT PROTECTIVE SYST

@ P SHUTOOWN MINIMUM MINIMUM F OPERA 8LE OEGREE OF L>

NO. FUNCTIONAL. UN!Y CHANNELS REQUNOANCY _ C4 4a. Circulator IA and 18 2 1 None Shutdown - Loop -

Shutdown Logic 4b. Circulator 1C and 10 2 1 Hone Shutdown - Loop Shutdown Logic Sa. Steam Generator )

Penetratton 2 (f) 1 None Overpressure. Loop 1 Sb. Steam Generator 2 (f) 1 Penetration None Overpressure Loop 2

64. High Reheat Header Activity, Loop 1 2 (f) 1 None 6b. High Reheat Header Activity, Loop 2 2 (f) 1 None

74. Low Superheat Header Temperature, Loop 1 (p) 2 (f) 1 Less Than 30%

Rated Power 7b. Low Suoerheat Header Temperature, Loop 2 (p) 2 (f) 1 Less Than 30% I Rated Power '

7c. High Offferential 2 (f) 1 Temperature Between Less Than 30%

Loop 1 and Loop 2 (p) Rated Power Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9

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Fort St. Vrain el Technical Specifications Amendment f 52, 60 Page 4.4-Sa l

Soecification L 0 4.4.1 Table 4.4 3 (Dart 1)

INSTRUMENT OPERATING RECUIREMENTS FOR THE PLANT PROTECTIVE SY CIRCut> TOR TRIP t TRIP ALLCWABLE NO. EL'NCTICNAL QNIT $$7o0!NT VALUE -

1. Circulator Speed 5 1850 com Below 1 2035 rpm Below '

- 1,cw Normal As Normal As Programmed by Programmed by Feecwater Flow Feecwater Flow 2a. t.ooo 1, Fixed > 230,500 lb/hr > 230,500 lb/hr Feecwater -

{20%ofnormal [20% of normal Flow - Low (Both Full Load) Full L. cad)

Circu14 tors) 2b. Loco 2, Fixed > 230,500 lb/hr > 230,500 lb/hr Feeowater [20% of normal [20%ofnormal Flow - Low (Both Full t.cac) Full t. cad)

Circulators)

3. t.oss of Circulator 3 459 psid t 454 psid Searing Water 4 Circulator 1 796 psig i 801 psig Penetration Treuele

5. Circulator Orain 1 8.5 osid 3 8.0 psid Malfunction

6. Cte:ulator Speed - 1 11,495 rpm High Steam 1 11.684 rpm

7. Manuel Not Not Applicable Applicable Notes for Tables 4,4-1 thr: ugh 4.4-4 are on Pages 4.4 8 and 4.4-9 4

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fora St.'Vrain el

' Technical Specifications Amendment # 60 Page 4.4-5b Specification t.C0 4.4.1 Table 4 4-3 (part 1)

INSTRUMENT CPERATING REQUIREMENTS FOR THE OLANT PROTECTIVE SYSTEM, CIRCULATCR TRIP TRIP AL'.0WABLE NO. FUNCTIONAL UNIT SETPOINT VALUE

8. Circulator seal 3 -5.2" H20, 3 -6.1" H20, '

Malfunction 3 +74.8" H2O 3 +76.1" H2O

9. Cf reulator Soeed - 5 8,589 rpm 1 8,786 rpe High Water Notes for Table 5 4.4*1 through 4.4-4 art on Pages 4.4*$ and 4.4*9

Fort St. Vrain el Technical Specifications Amendment f60 Page 4,4-5c SPECIFICATION LCO 4.4.1 TABLE 4.4-3 (Dart 2)

INSTRUMENT OPERATING REQUIREMENTS FOR PLANT PROTECTIVE SYSTEM.

CIRCU uTOR TRIP MINIMUM MINIMUM PERMISSIBLE OPERABLE OEGREE OF BYPASS NO. FUNCTIONAL UNIT CHANNELS REOUNOANCY CON 0!TIONS

1. Circulator Speed 2 (f) 1 Less Than 30%

- Low (r) Rated Power Za. l. cop 1, Fixed Feed- 2 (f) 1 Less Than 304 water Flow - Low Rated Power (Both Circulators) '

2b. Loop 2 Fixed Feed- 2 (f) 1 Less Than 30%

water Flow - Low Rated Power (Both Ctreulators)

3. Loss of Circulator 2 (f) 1 None Bearing Water (r) 4 Circulator 2 (f) 1 None Penetration Trouble (r)

5. Circulator Orain (

2 (f) 1 None Malfunction (r)

6. Circulator Speed - 2 (f) 1 None High Steam (r)

7. Manual 1 0 None B. Ctreulator Seal 2 (f) 1 Opposite loop Malfunction (r) shutdown or 1

I circulator seal l

malfunction trfo !

of otner circulator in same loop

9. Circulator Speed - 2 (f) 1 None High Water Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 I

Fort' St. Vrain #1 Technical Specifications Amendment # 60

. Page 4.4-7a 1 l

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Specification LCO 4.4.1 Table 4.4 4 (Part 1)

INSTRUMENT OPERATING RE0V!REMENTS FOR THE PLANT PROTECTIVE I I

SYSTEM, RCD WITNDRAWAL PR0i.~lBIT (RWP)

TRIP ALLOWABLE !

NO. FUNCTIONAL UNIT SETPOINT VALUE .

1. Startup Channel-Low 14.2cos 13.2 cps ;

Count Rate 2a. Linear Channel-Low > $5 >5%

Power RWP (Channels 3 Indicated Indicated 4 and 5) 1hermal Themal l Power (m) Power 2b. Linear Channel-Low > 5% >5% i Power RWP (Channels 6, Tndicated Indicated I 7 and 8) Thennal itarmal I Power (m) Power 3a. Linear CNnnel-High < 30% < 30%

Power RWi- (Channels 3, Tndicated Indicated .

4 and 5) Thermal Thermal )

Power (n) Power l 1

3b. Linear Channel-High < 30% < 30% '

Power RWP (Char.nels 6, Indicated Indicated I 7 and 8) Themal Themal Power (n) Power l

l l

Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4 9

.,. .v. .

Fort St. Vrain #1 Technical Specifications Amendmene 60

,- Page 4.4-7b ECIFICATIONLCO4.4.1 TABLE 4.4 4 (Part 2)

INSTRUMENT OPERATING REQUIREMENTS FOR REACTOR PROTECTIVE SY5 TEM, ROD WITHDRAWAL PROHIBIT (RWP)

MINIMUM MIN! MUM PERM!SSIBLE OPERABLE DEGREE OF BYPASS NO. FUNCTIONAL UN!T CHANNELS REDUNDANCY CONDIT!0NS

1. Startup Channel - Low' 2 1 Above 1.0s-035 Count Rate Rated Power 2a. Linear Channel - Low 2 1 (g)

Power RWP (Channels 3, 4, and 5) 2b. Linear Channel - Low 2 1 (g)

Power RWP (Channels 6, 7, and 8)

34. Linear Channel - High 2(f) 1 None PowerRWP(Channels 3, 4, and 5) 3b. Linear Channel - High 2 (f) 1 None Power RWP (Channels 6, 7,and8)

Notes for Tables 4.4 1 through 4.4-4 are on Pages 4.4-8 and 4.4-9 l

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Fort St, Vrain #1 Technical Specifications Amendment #2 ,60 Page 4.4-8 SPECIFICATION LCO 4.4.1 NOTES rus I Asto 4.4-1 m uusH 4.4-4 la) Deleted, b) Two themocouples from each loop, total of four, constitute one channel. For each channel, two thermocouples must be operable in at least one operating loop for that channel to be considered operable.

1 c) With one primary coolant high level moisture monitor tripped. I trips of either 1000 primary coolant moisture monitors will I cause full scram. Hence, number of operable channels (1) minus l mir 5um number required to cause scru (0) equals one, the i minimum degree of redundancy. l d) Deleted. l e) One channel consists of tid ee undervoltage relays each monitoring a single phase of a 400 VAC essential bus. A channel :

trip will occur when two of the three undervoltage relays comprising that channel .' opera te after a preset time delay indicating loss of bus voltage. Initiation of a scram requires ,

two of the three undervoltage relays on two of the three 480 VAC l essential buses to operate. I f) The inoperable channel must be in the tripped condition, unless the trip of the channel will cause the protective action to occur. Failure to trip the inoperable channel regi:f res taking the appropriate corrective action as listed on Pages 4.4-1 and 4.4 2 within the specified time limit.

g) RWP bypass pemitted if the bypass also causes associated single channel scram, hl) For loop monitors only, pemissible bypass conditions include:

1. Any circulator buffer seal malfunction.

!!. Loop hot reheat header high activity.

!!!. As stated in LCO 4.9.2.

h2) For high level monitors only, pemissible bypass conditions include:

1. As stated in LCO 4.9.2.

j) Deleted.

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k) One operable helium circulator inlet themocouple in an operable I loop is required for the channel to be considered operable. I m) Low Power RWP bistable resets at 4% after reactor power initially exceeds 5%.

l n) Power range RVP bistables automatically reset at 10% after reactor uwer is decreased from greater than 30%. The RW? may be manually reset between 10% and 305 power.

p) Item 74, must be accompanied by item 7c. for Loop 1 shutdown.

Item 7b. must be accompanied by item 7c. for Loop 2 shutdown.

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t-Fort St. Vrain '#1 Technical Specifications Amendment #43,60

Page 4.4-10 Basis for Specification LCO 4.4.1 The plant protect 1N system automatically initiates protective functions te prevent established limits from being exceeded. In addition, other protective instrumentation is provided to initiate action which-mitigates the consequences of accidents. Some protective actions are necessary only during startup and/or Low Power and require bypass at power; others are required during power operation and need to be bypassed at startup and/or Low Power. A simple method, based on a minimum of .

administrattu control, has been devised to sequence and bypass protective actions. The equipment consists of two selector switches (Reactor Mode and Interlock Sequence) on ,

tha reactor control board. This specification provides the limiting conditions for operation necessary to preserve the effectiveness of these instrument systems.

If the minimum operable channels or the minimum degrees of redundancy for each func*'onal unit of a table cannot be eet or cannot be bynessed under the stated permissibl's bypass coa 11tions, the following action shall be taken-For Table 4.4-1, the reactor shall be shut down within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

For Table 4.4-2, the affected loop shall be shut down within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . <

For Table 4.4-3, perform one of the following within 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

l 1) The reactor shall be shutdown, or

2) the affected helium circulator shall be shutdown.

For Table 4.4-4, the reactor shall be shut down within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If, within the indicated time limit, the minimum number of operable channels and the minimum degree of redundancy can be reestablished, the system is considered normal and no further action needs to be taken.

The trip level settings a o included in this section of the specification. The bases for these settings are briefly discussed below. Additional discussions pertaining to the scram, loop shutdown and circulator trip inputs may" be found in Sections 7.1.2.3, 7.1.2.4 and

7.1.2.6, respectively, of the FSAR. High moisture instrumentation is discussed in Section 7.3.2 of the FSAR.

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Fort S2. Vrain 01 Technical Specifications i Amendment # g ,60 Page 4.4-10a Basis for Specification LCO 4.4.1 (Contin'.ed)

To accommadate the instrument drift assumec , occur between operational tests and the accuracy to wni,h Trio Setpoints can be measured and calibrated. Allowabit Values and Trip Setpoints have been specified in Part 1 of ltoles A.4-1 through 4.4-4 The methodology used for calculat.ing tr.: AI'owable Values and Te p Setpoints is discussed in Technical Specification LSS! 3.3.

4. Scram le:ots The simultaneous insertion af the control rods will be initiated by the following conditions: ,

Manual Scram A manual scram is provided to give the operator means for emergency shutdown of the reactor independent of the automatic reactor protective system. The Reactor Mode Switch (RMS) in the "off" position also causes a manual scram.

Start up Channel - Mich Count Rate High start up count rate is provided as a scram for use du'ing fuel loading, preoperational testing, or other low power operations.

Lirear Channel - Nich (Neutron Flux)

See Technical Specification LS$$ 3.3.

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Fort St. Vrain el Technical Specifica tions Amendment o gg,60 Page 4.4-10b Basis for Soecification LC0 4.4.1 (Cont'nued) peimary Coolant Moisture - High See Technical Specification LS$$ 3.3. I Reheat Steam Temeerature - High See Technical Specification LS$$ 3.3.

primary Coolant pressure - programmed low See Technical Specification LS$$ 3.3.

primary Coolant pressure - programmed High '

s l

See Technical Specification LSSS 3.3.

Hot Reheat Header Pressure - L3 Low -oheat steam prfslura is an indication of either a cold -eheat steam i e ve a hot reheat steam line rupture in i eertion of line common to both loops.

Loss of the cold reheat steam line results in loss of the steam supply to the circulators which necessitates plant shutdown.

The direct scram in this case precedes a scram resulting from the two-loop trouble.

The loss of either steam line results in loss of plant generation output, and a reactor scram is appropriate in this sit.ation. The Trip Setpoint is selected to I l

be below normal operating and transient levels, which vary over a wide range.

Main Steam pressure - Low Low main steam pressure is an indication of main steam line rupture or loss of feedwater flow. Irnmedi a te shutcown of the reactor is appropriate in this case.

In addition, the superheater outlet stop check valves !

are automatically closed to reroute main steam to the flash tank (through the individual loop bypass valves and desuperheaters). This is required for the contir:yed operation of the helius circulators on steam. The Trip Setpoint is selected to be below i l

normal operating levels and systes transients.

i Plant Electrical System - t.oss Loss of plant electrical system power requires 'a scram to prevent any Powe r-to-Flow mismatches from l

occurring. A preset time delay is provided following a power loss before the scram is initiated to allow an emergency diesel generator to start. If it does start, the scram is avoided.

Fort St. Vrain #1 Technical Specifications Amendment #bo

Page 4.4-10c I

1 i

l Basis for Specification LCO 4.4.1 (Continued)

Two-Loor Trouble Scram logic l l

Operation on one loop at a maximum of about 50% power may continue following the shutdown of the other loop (unless by scram as in the case of high .

moisture) preceded Onset of trouble in the remaining loop (two-loop tr6uble) results in a scram. Trouble is defined as a signal which normally initiates a loop '

shutdown. Similarly, simultaneous shutdown signals to !

both loops result in shutdown of one of the two loops only, and a reactor scram. However, actuation of both i Steam Line Rupture Detection / Isolation System (SLRDIS) '

l loops, effectively snuts down both loops because it sends an actuation logic signal to all four circulator ,

trip logic channels. The consequences of a two-loco l shutdown and subsequent loss of forced circulation )

have been analyzed and found to be acceptable. The j consequences are bounded by an interruption of forced -

circulation cooling accident described in FSAR Section 14.4.2.2, Safe Shutdown Cooling.

High Reactor Building Temperature (Pipe Cavity)

High temperature in the pipe cavity would indicate the presence of a steam leak. A staam leak or pipe rupture under the PCRV within the support ring would also be detectable in the pipe cavity, therefore only one set of sensors and logic is required to monitor both areas. The setpoint tas been set above the SLRDIS pre-trip temperature s'arw..

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- ~ - , - - e , - . - , , .- , , , . , , _ , , _ . , . . . _ . , , _ , , ,

~ 90r4 M. Urdin #1 1

. Technical Specifications Annndment # E0,60 Page 4.4-11 Basis for Specification LCO 4.4.1 (Continued)

b. Loep Shutdown Inouts The following loop shutdown inputs are provided primsrily for equipment protection and are not relied upon to protect Safety Limits. Malfunction of these items could prevent a scram due to loss of the two loop trouble scram input. ,

Shutdown of Both Circulators (Loop Shutdown L'o gic)

Shutdown of both circulators in one loop is a loop shutdown input so that secondary coolant flow is automatically isolated to the affected loop's steam generator upon loss of primary coolant flow in that -

loop. This loop shutdown ensures proper reactor protection system action (scram) through the two-loop trouble scram in the event of the loss of all four circulators. Low feedwater flow to both loops can result in automatic trip of all four circulators, which would activiate the two loop trouble scram.

Steam Generator Penetration Overpressure (Loop 1/ Loop 2)

Steam generator penetration overpressure is indicative of a pipe rupture within the penetration. A loop shutdown is appropriate for such an accident, and the helium pressurizing line to the penetration is closed to prevent moisture backflow to the purified helium system. The penetration overpressure is handled by relief valves; however, to minimize the amount of steam / water released, the steam generator contents are also dumped.

The steam generator interspace rupture discs are set at 825 psig (nominal). The burst pressure range (plus or minus 21) is 808 psig to 842 psig (Technical Specification LS$$ 3.3, Table 3.3-1). The relief valve is sized to allow a 370 psi pressure drop in a safety valve inlet line when the valve is relieving at nameplate capacity of 126.000 lb/hr superheated steam at 1000 degree F. This prevents the penetration pressure from exceeding the reference pressure of 845 psig.

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' Fort St. Vrain el l l

Technical Specifications '

Aknendment # 60 l Page 4.4-11a l

r I

I Basis for Specification LCO 4.4.1 (Continued)

Hfoh Reheat Header Activity - (Loop 1/ Loop 2) l I

High reheat header activity is an indication of a reheater tube rupture resulting in leakage of reactor {

hellus into the steam systee. The Trip Setpoint ensures detection of major reheat tube ruptures and an {.

on-scale reading, with up to design value circulating activity for post accident sonttoring. Detection of smaller size leaks or leaks with low ctreulating coolant activity can be detected and aiarmed by the backup re h'es t condensate sonttors and/or the air I ejector monitor. '

l Low Superheat Header Temperature (Loop 1/ Loop 2) and Hton Differential Temperature Between Loop 1 and Loop 2 Low superheat header temperature in a loop is I indicative either of a feedwater valve or controller failure yielding an excessive loop feedwater flow rate or a deficiency of helium flow rate,. and a loop shutdown is appropriate. The required coincident high differential temperature between loops functions to prevent the loop Trip from occurring during normal operation at low main steam temperatures such as in a normal plant shutdown.

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Fort St, Vrain #1 l Technical Specifications i Amendment # 90,60

,- Page 4,4-12 Basis for Specification LCO 4.4.1 (Continued)

c. Circulater Shutdown inputs All circulator shutdown inputs are equipment protecticn items. With the exception of Circulator Speed High on I water turbine drive, all circulator shutdown inputs are I connected to the two-loop trouble scram logic through I the loop shutdown system. These items are included in l Table 4,4-3 because a mal'Jnction could prevent a scram due to loss of the two-loop trouble scram input.

Circulator Speed High on water turbine drive is 4 included to afford protection to the water turbine I assembly against the effects of overspeed during l continued core cooling upon loss of steam drive I capability. - l l

Circulator Speed - Low '

Too low a circulator speed causes a mismatch between thermal power input and heat removal (feedwater flow) in a steam generator, which may result in flooding the superheater section. The circulator Trip causes an l automatic adjustment, as required, in the turbine l governor setting, feedwater f!ow rate, and remaining circulator speed to maintain stable steam pressure and temperature conditions.

Looo 1/ Loop 2 Fixed Feedwater Flow - low Thi Fixed Feedwater Flow - t.ow is an equipment I protection feature designed to protect the steam generator from overheating for complete loss of feedwater flow.

Loss of Circulator Bearing Water In order to prevent circulator damage upon loss of normal and backup bearing water supplies, a gas pressurized water accumulator is fired when water pressure falls below the Trio Setpoint value. The Trip 5etpoint value is selected so that adequate water I pressure is available during circulator coastdown, I which lasts for about 30 seconds, to maintain clearances within the circulator bearings of at least 0.001 in. Tests and analyses have shown that a Trip at 450 psid provides substantial clearance margin ;

above 0.001 in, when the circulators are operating at '

normal speeds.

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Fort St. Vrain H Technical Specifications Amendment # 60

,- Page 4,4-82a Basis for Specifiestion LCO 4.4.1 (Continued)

Circulator penetration Trouble Circulator penetration overpressure is indicative of a pipe rupture within the penetration. A circulator Trip is appropriate for such an accident and the helium pressurizing line to the penetration is closed to prevent moisture backflow to the purified helium system. The overpressure is handled by the penetration relief valves. The penetration interspace rupture discs are set at 825 psig (nominal). The burst pressure range (plus or minus 2%) is 808 psig to 842 psig (Technical Specification LSSS 3.3, Table -

3.3-1). The relief valve is sized to allow a 40 psi pressure dr6p in the safety valve inlet line when the valve is relieving at nameplate capacity (170 gpm).

Circulator Orain Malfunction This Trip is provided to prevent steam from entering the bearing of an operating circulator. A differential pressure controller is utilized to maintain the bearing water main drain pressure above the steam turbine exhaust pressure. When the pressure differential drops, the steam water drain control valves are opened to prevent steam from entering the bearings. If the above controls do not work, three PPS differential pressure switches for each circulator, set at greater than er equal to 8.5 psid, will initiate an automatic shutdown of the circulator.

Circulator Speed - High Steam The speed sensing system response and Trip setting are ;

chosen so that under the maximum overspeed situation 1 possible (loss of restraining torque) the circulator I will remain within design criteria. l l

Circulator Trio - Manual (Steam / Water) j A manual Trip of each circulator for both steam and water turbine drives is available so that in an emergency an operator can trip a circulator when j required.

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Fort $2. Vrain 01 !

Technical Specifications l Amendment # 60 l

,. Page 4.4-12b Basis for Speciff eation LCO 4.4.1 (Continued)

Circulator Seal Malfunction (Low /Hich)

A high reverse differential of -6.1" H2O would be reasonable evidence that bearing water is leaking into the primary coolant system. An increasing differential pressure of +76.1" H2O would be reasonable evidence that primary coolant is leaking into the bearing water and thus into the closed circulator service system. In both cases a circulator trip with brake and seals set is appropriate.

Circulator Smeed - Hioh Water The Trip Setpoint has been established above normal -

operating speeo. Equipment testing ensures that this Trip Setpoint will prevent failure due to fatigue cracking.

Steam Leak Detection in the Reactor Sutidino steam Leak Detection in the Reactor %11 ding is required for equipment qualification of Safe Shutdown Cooling Systems. The ALLOWA8LE VALUE is set at 152.8 degrees F per minute rate of rise in order to prevent exceeding the harsh environment temperature profile to which the safe shutdown electrical equionen t is qualified, per the requirements of 10CFR50.49. A setpoint calculation analysts performed per ISA Standard 567.04 and RG1.105 results in the stated ALL0wA8LE VALUE and TRIP SETPOINT as specified in the LCO and this basis. The TRIP SETPOINT has been establishoo with suf ficient margin between the technical specification limit for the process variable and the nominal TRIP SETPOINT to allow for

1) inaccuracy of the instruments; 2) uncertainties in the calibration; 3) instrument drift that could occur during the interval between calibrations; and 4) inaccuracies due to ambient temperature changes, vibration and other environmental conditions. The TRIP SETPOINT is set at 152.3 degrees F per minute rate of rise until such time as the drift characteristics of the detection systee are better understood from actual plant operating esperience and the assumptions used in the setpoint analysis are verified.

SLA0!$ destgh incorporates two panels, each with its own set of sensors for the Reactor and Turbine Buildings and dual logic trains in each panel. The 1 SLACIS design preserves the single failure concept. A l single failure will neither cause nor prevent SLRDIS actuation in the event of a high energy line break. l The probability of an inadvertent actuation is I extremely small due to the matrin logic eetleyed for circulater trip and valve actuation. The SLR0!$

panels are referred to as "loops"; however, due to the way the outputs of the panels are cometned to provide protective action and satisfy the single failure concept, the SLR0!$ loops do not correspond to primary or secondary loops. 5

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Fort St. Vrain 4 Technical Specifications Amendment , u0 e Page 4.4-12e Basis for Speci f f eat'en LC0 a.4.1 (Continued)

For each SLROIS loop, the OPERABILITY requirements and ;

their respective ACTIONS represent good operating practices and judgment for a four enannel detection system with a 2 of a coincidence trip logic. The fourth channel may be placed in bypass for test and/or maintenance purposes, subject to the ACTION statement !

restrictions, while preserving a 2 of 3 coincidence logic OPERAaLE. The Steam Line Rupture Detection / Isolation System as designed and installed has spare channels available for input. Any of the available channels may be selected for input signal processing pro'vided the surveillances are current on the channels used. The SLROIS is required to be -

OPERABLE only at power (above 2% rated thermal power).

Analyses with rated reactor power at 2% demonstrate that automatic actuation of SLR0!$ is not likely to occur during a high energy line break lasting until it is manually terminated one hour initiation. The. temperatures as analyzed infollowie.g at both tne reactor and turbine buildings stay well below the temperature for which the equipment is qualified.

The ACTION statements for inoperable SLNIS detection and information processing equipment allow one channel

' in each butiding to be inoperable for up to 7 days; a second inoperable channel in either butiding require..

that power be reduced to below 2% within 12 hou The 7 day ACTION time for a single detector channe'.'s. is acceptable based on preservation of a 2 out of 3 coireidence detection system still in operation.

ACT'.ON 3 is applicable to other functions within the SLRJIS instrumentation panel such as loss of power from instrument buses, or other failures in the logic trains and Essociated electronics. A 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time period in ACTION 3 for inoperability of those ;

associated SLR0!$ functions minimized the time that SLR015 may operate with limited functional capaellity.

An inoperable valve or associated equipment is al? owed !

for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . High energy line break analysis for i environmental qualification assumes the worst-case 1 single active failure. Thus, a single valve inoperable for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is within the bounds of analysis. When two or more valves and/or associated equipment is inoperable, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed to restore the inoperable equipment. Repairs may be performed while the plant is at power, thus, minietzing thermal cycling of plant and insta11ec equipment.

Steam Leak Oetection in the Turbine But1dino is required for equipment qualification of Safe Snutdown Cooling Systees. Thus, the Itaits and basis are the '

same as discussed in the bests for steam leak detection in the reactor building. ,

I

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/ Page 4.4-13 Basis for Specification LCO 4.4.1 (Continued)

d. Rod Withdrawal Prohibit Inputs The temination of control rod withdrawal to prevent further reactivity addition will occur with the following conditions:

Startup Channel - Low Count Rate Start-up Channel - Low Count Rate is provided to prevent control rod pair withdrawal and reactor startup without adequate neutron flux indication. The .

trip level is selected to be above the background noise level."

Linear Channel - Low Power RVP Linear Channel (55 Power) directs the reactor operator's attention to either a downscale failure of a power range channel or improper positioning of the Interlock Sequence Switch. (FSAR Sections 7.1.2.2 and 7.1.2.8)

Linear Channel - High Power RWP Linear Channel (30% Power) is provided to prevent contml rod pair withdrawal if reactor power exceeds the Interlock Sequence Switch limit for the "Low Power" position. (FSAR Sections 7.1.2.2 and 7.1.2.8) l

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