ML20210C392

Amend 108 to License DPR-40,incorporating Revised Limiting Conditions for Operation & Surveillance Requirements for Steam Generator Isolation Signal

ML20210C392
Person / Time
Site:	Fort Calhoun
Issue date:	04/28/1987
From:	Calvo J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20210C385	List: ML20210C382 ML20210C392 ML20210C397
References
TAC-61975, NUDOCS 8705060176
Download: ML20210C392 (10)

Text

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/ 'o UNITED STATES g

! n NUCLEAR REGULATORY COMMISSION g E WASHINGTON. D. C. 20555

\....+/

OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1

. AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.108 License No. DPR-40

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

A. The application for amendment by the Omaha Public Power District (the licensee) dated July 17, 1986 complies 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 Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

8705060176 870428 PDR ADOCK 05000285 P PDR

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2. Accordingly, Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No.

DPR-40 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 108, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION d G. 0h Jose A. Calvo, Director ProjectDirectorate-IV DivisionofReactorProjects-III, IV, V and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: April 28, 1987

ATTACHMENT TO LICENSE AMEN 0 MENT NO.108 FACILITY OPERATING LICENSE NO. OPR-40 DOCKET N0. 50-285 Revise Appendix "A" Technical Specifications as indicated below. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages 2-61 2-61 2-62 2-62 2-63 2-63 2-64 2-64 i

2-65 2-65 2-69 2-69 2-69a 2-69a

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D

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. 2.0 LIMITING CONDITIONS FOR OPERATION 2.14 Engineered Safety Features System Initiation Instrumentation Settings Applicability Applies to the engineered safety features system initiation instrumentation settings.

Objective To provide for automatic initiation of the engineered safety features in the event that principal process variable limits are exceeded.

Specifications The engineered safety features system initiation instrumentation setting limits shall be as stated in Table 2-1.

Basis (1) High Containment Pressure The basis for the 5 psig set point for the high pressure signal is to establish a setting which would be exceeded quickly in the event of a DBA, cover a spectrum of break sizes, and yet be far enough above normal operation maximum internal pressure to prevent spurious initiation.

High containment pressure initiates the steam generator isolation signal which will close the main steam isolation and bypass v&lves and the main feedwater isolation and bypass valves.

(2) Pressurizer Low Pressure The pressurizer low pressure safety injection signal is a diverse signal to the high containment pressure safety injection signal.

The 1600 psia setting includes an uncertgipty of + 22 psia and is the setting used in the safety analysis.t li (3) Containment High Radiation (Air Monitoring)

The containment air monitoring system comprises a moving paper filter particle monitor (channel RM-050) and a sample chamber gas monitor (channel RM-051) installed in a comon housing.(2)

Optionally, the sampling point for channels RM-050 and RM-051 can be switched from the containment to the ventilation discharge duct.

The ventilation discharge monitoring system consists of a moving l paper filter particle monitor (RM-061) and a sample chamber gas '

monitor (RM-062) installed in a comon housing. An iodine monitor for I-131 (RM-060) also monitors these releases.

2-61 Amendment No. 108

2.0 LIMITING CONDITIONS FOR OPERATIONS 2.14 Enaineered Safety Features System Initiation Instrumentation Settings (Continued)

(3) ContainmentHighRadiation(AirMonitoring) (Continued)

The setpoints for the isolation function will be calculated in accordance with the ODCM.

Each channel is supplied from a separate instrument A.C. bus <

and each auxiliary relay requires power to operate. On failure of a single A.C. supply, the A and B matrices will assume a one-out-of-two logic.

(4) Low Steam Generator Pressure A signal is provided upon sensing a low pressure in a steam generator to close the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition ,

of reactivity. The setting of 500 psia includes a +22 psi i

uncertainty and was the setting used in the safety analysis.(3)

Closure of the MSIVs (and the bypass valves, along with main feedwater isolation and bypass valves) is accomplished by the steam generator isolation signal which is alggicalycombleation of low steam generator pressure or high containment pressure.

As part of the AFW actuation logic, a separate signal is provided to terminate flow to a steam generator upon sensing a low pressure in that steam generator if the other steam generatcr pressure is greater than the pressure setting. This is done to minimize the temperature reduction in the reactor coolant system in the event of a main steamline break. The setting of 466.7 psia includes a +31.7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.

(5) SIRW Tank Low Level Level switches are provided on the SIRW tank to actuate the valves in the safety injection pump suction lines in such a manner so as to switch the water supply from the SIRW tank to the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal. The switchover point of 16 inches above tank bottom is set to prevent the pumps from running dry during the 10 seconds required to stroke the valves and to hold in reserve approximately 28,000 The FSAR lossgallons of ataccident of coolant least 1800 ppm (4) analysis borated water.

assumed the recirculation started when the minimum usable volume of 283,000 gallons had been pumped from the tank.

2-62 Amendment No. J,)/,J),$),$p,)p),108

2.0 LIMITING CONDITIONS FOR OPERATIONS 2.14 Engineered Safety Features System Initiation Instrumentation Settings (Continued)

(6) Low Steam Generator Water Level As part of the AFW actuation logic, a signal is provided to initiate AFW flow to one or two steam generators upon sensing a low water level in the steam generator (s) if the absolute steam generator pressure criteria are satisfied. This function ensures adequate steam generator water level is maintained in the event of a failure to deliver main feedwater to either steam generator. The setting of 28.2% of wide range tap span includes a +13.2% uncertainty; therefore, a setting of 15% of wide range tap span was used in the safety analysis.

(7) High Steam Generator Delta Pressure As part of the AFW logic, a high steam generator differential pressure signal is generated to provide AFW to the higher pressure steam generator with a concurrent low level signal if both steam generator pressures are less than 466.7 psia.

If the differential pressure between steam generators is less than the setting, neither steam generator is supplied with AFW in the presence of a low level signal. The setting of 119.7 psid includes a -15.3 psi uncertainty; therefore, a setting of ,

135 psid was used in the AFW safety analysis.

References (1) USAR, Section 14.1.3 (2) USAR, Section 11.2.3.2 <

(3) USAR, Section 14.12 (4) USAR, Section 14.15 (5) USAR, Section 7.4.6 ,

(6) USAR, Section 7.5.2.5 (7) USAR, Section 14.4.1 l

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2-63 Amendment No. 65,108

TABLE 2-1 g

m

5. Engineered Safety Features System Initiation Instrument Setting Limits g

k Setting Limit Functional Unit Channel g:

1. High Containment Pressure a. Safety Injection 1 5 psig 3

. b. Containment Spray (3) a c. Containment Isolation

? d. Containment Air Cooler a DBA Mode

? e. Steam Generator Isolation 5 '

2. Pressurizer Low / Low Pressure a. Safety In,jection 1 1600 psia (1)

b. Containment Spray (3)

c. Containment Isolation

d. Containment Air Cooler m DBA Mode a In accordance with the Offsite

• 3. Containment High Radiation Containment Ventilation Isolation Dose Calculational Manual

4. Low Steam Generator Pressure a. Steam Line Isolation > 500 psia (2)

b. Auxiliary Feedwater Actuation _s 466.7 psia

5. SIRW Low level Switches Recirculation Actuation 16 inches +0, -2 in. above -

tank bottom ,

6. 4.16 KV Emergency Bus low a. Loss of Voltage (2995.2 + 104) volts ,

Voltage > Trip

. 1 5.9(4)20.8 seconds

b. Degraded Voltage > 3825.52 volts

• T"'~P i) Bus 1A3 Side T4.8 .5) seconds ,

2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems Applicability Applies to plant instrumentation systems.

Objective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specifications The operability, pemissible bypass, and Test Maintenance and Inoper-able bypass specifications of the plant instrument and control systems shall be in accordance with Tables 2-2 through 2-5.

(1) In the event the number of channels of a particular system in service falls one below the total number of installed channels, the inoperable channel shall be placed in either the bypassed or tripped condition within one hour if the channel is equipped with a key operated bypass switch, and eight hours if jumpers or blocks must be installed in the control circuitry. The inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of discovering loss of operability; however, if the inoperability is determined to be the result of malfunctioning RTDs or nuclear detectors supplying signals to the high power level, themal margin / low pressurizer pressure, and axial power distribution channels, these channels may be bypassed for up to 7 days from time of discovering loss of operability. If the inoperable channel is not restored to operable status after the allowable time for bypass, it shall be placed in the tripped position or, in the case of malfunc-tioning RTDs or linear power nuclear detectors, the reactor shall be placed in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If active maintenance and/or surveillance testing is being perfomed to return a channel to active service or to establish operability, the channel may be bypassed during the period of active maintenance and/or surveillance testing.

This specification channel applies when the plant is attoorthe high10-above rage trip-wide

% power andrange log is operating below 15% of rated power.

(2) In the event the number of channels of a particular system in service falls to the limits given in the column entitled " Minimum Operable Channels", one of the inoperable channels must be placed in the tripped position or low level actuation pemissive position for the auxiliary feedwater system within one hour, if the channel is equipped with a bypass switch, and within eight hours if jumpers or blocks are required. If the channel has not been restored to operable status after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of discovering loss of operability, the reactor shall be placed in a hot shutdown condition within the follow-ing 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, operation can continue without containment 2-65 Amendment No. 8, 29, 54, 65, M,108

TABLE 2-4 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS Test Maintenance Minimum Minimum Pemissible and Operable Degree of Bypass Inoperable Channels Redundancy Condition 8voass

& Functional Unit 1 Containment Isolation 1 None None N/A A Manual B Containment High During Leak (f) 1 Pressure A B

2((a)(e) 2 a)(e) 1 Test Pressurizer tow / Low A 1 Reactor Coolant (f)

B 2((a)(e) 2 a)(e) 1 PressuregsThan 1700 psia l

2 Steam Generator Isolation 1 None None N/A A Manual i

1 None None N/A B Steam Generator Isolation (i) Steam Generator Steam Generator (f)

Low Pressure A 1/ Steam 2/Ste'a9 Gen l s Gen Pressure 550 psia c)(lessThan B 2/St9ag 1/ Steam Gen 181 Gen (ii) Containment High During Leak (f)

A 1 Pressure B 2((a)(e) 2 a)(e) 1 Test 3 Ventilation Isolation 1 None None N/A A Manual B Containment High (f)

A None If Containment Radiation 2((d) Ventilation B 2 d) None Isolation Valves Are Closed a A and B circuits each have 4 channels, b Auto removal of bypass above 1700 psia.

c Auto removal of bypass above 550 psia.

2-69 Amendment No. EE, SA, 108

TABLE 2-4 (Continued) d A and B circuits are both actuated by any one of the five VIAS initiating channels; RM-050, RM-051, RM-060, RM-061, or RM-062; however, only RM-050 and RM-051 are required for containment ventilation isolation.

e If minimum operable channel conditions are reached, one inoperable channel must be placed in the tripped condition within eight hours from the time of discovery of loss of operability. The remaining inoperable channel may be bypassed for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of discovery of loss of operability and, if an inoperable channel is not returned to operable status within this time frame, a unit shutdown must be initiated (see Specification 2.15(2)). l f If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from the time of discovery of loss of operability. If bypassed and that channel is not returned to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of discovery of loss of operability, that channel must be placed in the tripped condition within the following eight hours. (See Specification 2.15(1) and exception l associatedwithmaintenance.)

2-69a Amendment No. SS,108

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